I DO-24042

SUMMARY REPORT OF DESIGN CRITERIA FOR A THERMAL FLUX LIQUID METAL PACKAGE LOOP IN THE ADVANCED TEST REACTOR

March lOr.'i

Babcock and Wilcox Comp; Lynchburg, Virginia

Ebasco Services Inc. New York, New York

#

DISCLAIMER Portions of this document may be illegible in electronic image products. Images are produced from the best available original document.

L E G A L N O T I C E This report was prepared as an account of Government sponsored work. Neither ttw United States, nor Che Commission, nor any person acting on behalf of the Commission":

A. Makes any warranty or represenUUon, expressed or Implied, with respect to the accu­racy, completeness, or usefutness of the Information contained In this report, or that the use of any Information, apparatus, method, or process disclosed In this report may not infringe privately owned rights; or

B. Assumes any ItabtlUies with respect to the use of, or for damages resulting from the use of any Information, apparatus, method, or process disclosed In this report.

As used In the above, "person acting on behalf of the Commission" includes any em­ployee or contractor of the Commission, or employee of such contractor, to the eirtent that such employee or contractor of the Commission, or employee of such contractor prepares, dlsaemlnates, or provides sccesa to, any information pursuant to his employment or contract with the Commission, or his employment with such contractor.

This report has been reproduced directly from the best available copy.

Printed in USA. Price $2.75. Available from the Office of Technical Services, Department of Commerce, Washington 25, D. C,

IDO-24042 REACTOR TECHNOLOGY

SUMMARY REPORT OF DESIGN

CRITERIA FOR A THERMAL FLUX LIQUID METAL

PACKAGE LOOP IN THE ADVANCED TEST REACTOR

P r e p a r e d For

U . S . ATOMIC ENERGY COMMISSION IDAHO OPERATIONS OFFICE ADVANCED TEST REACTOR

CONTRACT NO. AT(I0-1)-1075

MARCH 1963

Approved By:

:/ i-i U; A. H. Lazar

r-^C/^

R. H. Gordon

BABCOCK & WILCOX COMPANY ATOMIC ENERGY DIVISION

1201 KEMPER STREET LYNCHBURG, VIRGINL\

EBASCO SERVICES INCORPORATED TWO RECTOR STREET

NEW YORK 6, NEW YORK

PREFACE

This Design Cr i t e r i a for the Liquid Metal Loop Faci l i t ies associated

with the Advanced Test Reactor was p repared by Ebasco Services Incor­

porated and its nuclear subcontractor the Babcock h. Wilcox Company,

under Contract No. AT( 10-1) - 1075, adminis tered by the Idaho Operations

Office of the United States Atomic Energy Commission.

The Design Cr i t e r i a for a par t of the planned Liquid Metal Loop

Faci l i t ies was issued in July 1962 in IDO-24041 and the complete resu l t s

a re now presented in IDO-24041, Supplement 1, IDO-24042, and IDO-

24043. To achieve a cohesive presenta t ion of these interdependent doc­

uments the following History, Abst rac t , and Commentary have been p r e ­

pared for both Fas t and Thermal Flux Loops and the associated Hot

Cell Faci l i ty . Accordingly, and for r eade r continuity and clari ty, these

sections a re repeated in each docunnent.

HISTORY

A need for facil i t ies in which prototype liquid metal cooled fast and

the rmal r eac to r a s sembl ie s could be i r rad ia ted under conditions of high,

fast and the rma l neutron fluxes has existed for some t ime. Heretofore,

however, none of the available facili t ies possessed the requis i te c h a r a c ­

t e r i s t i c s to satisfy these r equ i r emen t s .

The Atomic Energy Connmission recognized that the Advanced Test

Reactor (ATR) possessed the cha rac te r i s t i c s for these types of exper i ­

ments and reques ted Atomic Power Development Associa tes , Inc. (APDA)

to review the ATR design, the exper imenta l requ i rements of the liquid

meta l cooled r eac to r p r o g r a m s , and the design features and r equ i r e ­

ments of in-pile package loops for test ing these assembl ies in the ATR.

APDA repor ted , APDA-144, that the ATR could be used for the

i r rad ia t ion of typical fast r eac to r fuel assembl ies to high burnups in a

fast r eac to r environment and recommended that the P r a t t &c Whitney

Aircraf t PW-19 loop design be used for this purpose with relat ively

minor modifications, namely, lengthening the tes t specimen section and

installing a removable bottom closure which would permi t the removal

and re inse r t ion of tes t spec imens .

The appropr ia te fast neutron spec t rum would be provided by a

boron loaded t he rma l neutron filter surrounding the test section in

conjunction with an extended core to compensate for the large react ivi ty

dec rease caused by the fi l ter. APDA's reasons for recommending the

adaptation of the PW-19 loop design for this tes t facility were that they

found no major design problems and that the cost of the associa ted r e ­

sea rch and development p rogram could be kept to a minimum since the

n e c e s s a r y work on the major loop components had already been done by

P r a t t & Whitney Aircraf t .

APDA also repor ted , APDA- 145, that the ATR could be used as a

tes t facility for sodium cooled the rmal reac to r fuel a s sembl i e s . The

bas i s for this decision was the high the rmal neutron flux available in

the ATR which would a s s u r e adequate heat generat ion with acceptable

flux and power depress ions while providing a tes t section of sufficient

s ize to obtain meaningful data on the hydraulic and the rma l cha rac ­

t e r i s t i c s of a smal l group of fuel e lements , in addition to i r radia t ion

data on the fuel itself.

iii

The conceptual design of the thermal flux loop presented by APDA (

utilized a package loop s imi la r in design to tliat proposed for the Fas t

Flux Loop. Sodium was proposed as the p r imary coolant ra ther than NaK

since it is the coolant that would be used in the reac tor sys tems being

tested and because of its bet ter heat t ransfer capabilit ies. An additional

feature of the APDA design was the control of sodium tempera ture by

varying the effectiveness of the sod ium- to- reac to r water heat exchanger

which, depending on the t empera tu res required utilized mercu ry and/or

helium as the heat t ransfer fluid(s) between the sodium and the reac tor

water .

Subsequent to the issuance of these two documents, the Atomic

Energy Connmission asked Ebasco Services Incorporated and its nuclear

subcontractor, the Babcock & Wilcox Company (B&W) to review these

documents . This review indicated that the infornnation presented was not

in sufficient detail to permit the initiation of Title I design of these loops.

Therefore , in January 1962, the Atomic Energy Commission requested

that Ebasco p repare design c r i t e r i a for these loops based on utilizing the

package concept previously recommended by APDA. However, the use of

m e r c u r y to vary heat exchanger effectiveness in the the rmal flux loop

would not be permit ted, and another heat t ransfer scheme was to be

devised. In the period between January and October 1962 a detailed effort

was nnade towards establishing sat isfactory c r i t e r i a for these loops. Dur­

ing this period, many heretofore unrecognized problems of extreme im­

portance were uncovered which indicated the necess i ty of re-evaluat ing

the package loop concept as applied to the experiment and reac tor r e ­

qui rements .

In October 1962, a meeting was held to d iscuss the various advantages

and disadvantages of the package loop concepts based on the inherent r e ­

s t r ic t ions imposed by the ATR and the requi rements of the established

design objectives. As a resul t of this meeting the Atomic Energy Com­

miss ion directed Ebasco and B&W to stop work on the development of

design c r i t e r i a based on the package loop concept, to p repare a report

summariz ing all work completed during the prepara t ion of these design

c r i t e r i a , and to state the reasons for terminat ing work on the package ^

loop concept. The documents which make up this report a r e :

IV

IDO-24041 - Design Cr i te r ia for a Fas t Flux Liquid Metal

Loop in the Advanced Test Reactor Issued

July 1962.

IDO-2401, Supple- Design Cr i te r ia for a Fast Flux Liquid Metal

Loop in the Advanced Test Reactor ,March 1963.

IDO-24042 - Summary Report of Design Cr i t e r i a for a

T h e r m a l Flux Liquid Metal Package Loop in

the Advanced Test Reactor, March 1963.

IDO-24043 - Summary Report of Design Cr i t e r i a for the

Liquid Metal Package Loops' Hot Cell Facil i ty

and Handling System at the Advanced Test

Reactor , March 1963.

v

ABSTRACT

The following is a suinnriary of the resu l t s of the c r i t e r i a work pe r ­

formed pr ior to the termination of work on the package concept, for the

liquid metal loop facilities for the Advanced Test Reactor. The purpose

of this work was to establish design c r i t e r i a for the loops and their a s ­

sociated facili t ies which would satisfy the design objectives specified

by the Atomic Energy Commission.

The following briefly summar i ze s , at the t ime when work was

terminated, the conceptual designs and design c r i t e r i a which had been

de termined:

1) nuclear c r i t e r i a for the thermal neutron fil ter and extended core combination in conjunction with the reference tes t specimen for the Fas t Flux Loop.

2) an appropr ia te secondary coolant, which in con­junction with the PW-19 loop, would provide the necessa ry heat exchanger performance capability to satisfy the design objectives for the Fas t Flux Loop.

3) pump performance required to satisfy the Fas t Flux Loop design objectives when utilizing the PW-19 loop.

4) physical modifications required to permi t the use of the PW-19 loop as the Fas t Flux Loop in the ATR and the effects of these modifications on the loop performance capability.

5) Research and Development Program required to confirm and achieve the design c r i t e r i a es tab­lished in IDO-24041.

6) conceptual design of the thermal loop including the in -core section, sodium-to-hel ium heat exchanger, heliunri-to-reactor water heat exchanger, pump motor and associated bearing design requi rements , pre l iminary shielding requi rements , control system requi rements , and e lec t r ica l hea te r s within the permiss ib le space envelope.

7) nuclear c r i t e r i a for the in -core section in conjunc­tion with the reference thermal flux test specimen which would satisfy the requirennents of the design objectives.

8) design c r i t e r i a for the secondary coolant sys tems including c o m p r e s s o r s , heat exchangers, f i l te rs , snubbers, purification sys tems , helium supply system, vacuum system, piping, valving, and control sys tems which would permi t the sat isfactory opera­tion of the Fas t and Thernnal Flux Loops in confor­mance with the design objectives.

vi

9) shielding c r i t e r i a of the secondary sys tems pip­ing and f i l ters based on the activit ies of impuri t ies in the helium and res idual s ta inless s teel chips blown out of the sys tem during operation.

10) design c r i t e r i a for the utili t ies and s t ruc tu res r e ­quired for sat isfactory operation of the loops.

11) conceptual design of a removable end c losure for the loops which would pe rmi t remote renaoval and inser t ion of fuel e lements .

12) conceptual design of top end connections which were sat isfactory for remote manipulation.

13) conceptual specimen removal and handling p ro ­cedures and the associa ted cooling requi rements during these operat ions .

14) conceptual loop handling procedures for al l s tages of handling operat ions and the cooling requi rements during these operat ions .

15) neutron and gamma heat generation ra tes in the in-core tubes which were used in prepara t ion of the loop s t r e s s ana lys is .

16) sys tem performance and hydraulic cha rac te r i s t i c s in conformance with the design objectives.

17) radiation levels that could be expected from the loops during various handling operations and pre l iminary loop shielding a r r angemen t s .

18) fission gas inventory contained in the loops at the completion of operation for 10,000 hours .

19) conceptual hot cell facility design which would mee t the needs of the liquid meta l loop program.

20) based on the conceptual design of the hot cell facility a rch i tec tu ra l , s t ruc tu ra l , mechanical , e lec t r ica l , nuclear , and viewing c r i t e r i a were establ ished.

21) conceptual design with accompanying c r i t e r i a for the la rge iner t a tmosphere hot cell .

22) conceptual design of a special ized liquid metal handling and purification systenri required to pe rmi t r e - i r r ad i a t i on of fuel elements and r e ­use of the Thermal Flux Loop.

23) space requi red to perform the various fuel e le­ment examinations required by the design ob­jec t ives .

However, as the c r i t e r i a work p rogressed it was determined that

' 'despite the accompl ishments summar ized above not all the p rogram r e ­

qui rements could be satisfied. It was also determined that a comprehensive

V l l

and costly r e s e a r c h and development program would be necessa ry to

establ ish basic design and operational inforrpation for the loops and their

associa ted equipment and faci l i t ies . The need for such a p rogram had

not been previously contemplated and, in addition, there was no a s s u r ­

ance that the performance of this p rogram would permi t the achieve­

ment of the loop performance and operating capabili t ies required by the

design objectives.

While no single i tem can be identified as the major obstacle in the

achievement of a sat isfactory packaged loop design, evaluation of the

many loop design problems - which requ i re solution in such a way as

to satisfy the exper imente r ' s design objectives without compromising

reac tor design or operating philosophy - ra ised considerable doubt as

to the feasibili ty of this concept. These problems a r e summar ized h e r e ­

after. Also, s ince mos t of the design problems a r e common to both the

Fas t and Thermal Flux Loops, they have not been segregated for each

concept.

1) The design and operation of a packaged pump-motor combination to mee t the hydraulic performance r e ­quired within the space l imitations appears m a r ­ginal, par t icular ly for the thermal loop.

2) The objective of a 10,000-hour loop life ra ised many significant ma te r i a l s problems, the mos t notable of which we re :

a) Design of bearings and selection of lubricants to operate in the liquid metal vapor and high radiation environment.

b) Design of motor winding insulation in conjunc­tion with appropr ia te shielding to operate in the high radiation environnment.

c) Selection of an adequate p r e s s u r e tube ma te r i a l w^hich would unquestionably perform as required .

3) The design objective of re -us ing loops throughout the des i red 10,000-hour life ra i sed many quest ions. The PW-19 loop had not been designed to be re -used , and its use as the Fas t Flux Loop precluded meeting this design objective. While the Thermal Flux Loop is a new design, the feasibility of re-uSing it is questionable since (a) the design of sat isfactory loop fill l ines which would permi t remote manipulation, in the l imited space available within the loop envelope.

V l l l

had not been achieved, (b) the effectiveness and operabil i ty of the external sodium purification sys tem had not been demonstra ted and (c) the pract icabil i ty of remotely machining and welding of ma te r i a l s that had been subjected to varying periods and intensi t ies of radiation, elevated t empera tu re s and liquid metal environment had not been established.

4) The design objectives required the adaptation of PW-19 loop lo satisfy the requi rements of the F a s t Flux Loop. Many of the design changes r e ­quired to enable its installat ion in ATR would be of such a nature as to question the pract ical i ty of such adaption. The mos t significant changes would be :

a) Modification of the heat exchanger to permi t refueling of the ATR without moving the loop.

b) Modification of the helium re turn tubes to pe rmi t refueling the ATR without moving the loop.

c) Modification of the liquid metal and helium fill l ines to pe rmi t installation of previously i r r ad ia t ed fuel specimens in a new loop within the hot cell facility

5) Since the ATR core has regions which exhibit posi ­tive void coefficients of reactivity, a loop rupture would resu l t in an inc rease of react ivi ty. There ­fore , it is deemed n e c e s s a r y to ei ther double con­tain the i n - r eac to r portion of the loop high p r e s s u r e (greater than r eac to r operating p res su re ) helium secondary coolant sys tem or reduce the heliuna p r e s s u r e below that of the r eac to r water . It is doubtful, because of the fixed design of the PW-19 loop, that the Fas t Flux Loop could be double con­tained and sti l l at tain the other design objectives. Use of secondary coolant at a reduced p r e s s u r e would dec rea se the maximum achievable specimen fission power below the design objective of 150 kw. F o r the thermal loop, even with the g rea t e r design lat i tude, the feasibili ty of double containment while sti l l meeting al l the design objectives is a lso doubtful.

6) The design objective of r e - i r r ad i a t i on of tes t spec i ­mens is considered marg ina l since (a) the p ro ­cedures required to a s s u r e adequate cooling during specimen removal , examination and r e - inse r t ion a r e complicated and hazardous , (b) the feasibility of s imultaneously heating the loop to 250 F-400 F to fill with sodium while cooling the tes t section to prevent specimen overheating has not been es tab­l ished, (c) the absolute minimum required decay

I X

p e r i o d s of four weeks for a fas t t e s t s p e c i m e n and 27 weeks for a t h e r m a l t e s t s p e c i m e n be fo re they can be r e - i n s e r t e d (even with foirced cool ing) r e ­s u l t s in a long and expens ive p r o g r a m to a c h i e v e 10,000 h o u r s of i r r a d i a t i o n , and (d) the high r a d i a ­t ion l e v e l s f r o m the loop would r e q u i r e r e m o t e handl ing p r o c e d u r e s wi thin the r e a c t o r v e s s e l a s wel l a s the hot c e l l which f u r t h e r c o m p l i c a t e s the v e r t i c a l handl ing of the 16-foot long, 1,000-pound loop .

7) While the f i l t e r and ex tended c o r e a p p e a r e d f e a s ­ib le to sa t i s fy the phys i c s r e q u i r e m e n t s of the F a s t F lux Loop, the high hea t g e n e r a t i o n r a t e s i n ­duced in the f i l t e r by the a t t enua t ion of t h e r m a l n e u t r o n s c a u s e d e x c e s s i v e m e c h a n i c a l and t h e r m a l s t r e s s e s in the concep tua l d e s i g n of the f i l t e r and loop p r e s s u r e tube . In addi t ion , the r e p l a c e m e n t of t h e s e c o m p o n e n t s a t 2 ,000- and 8 0 0 - h o u r i n t e r ­v a l s r e s p e c t i v e l y , a s a r e s u l t of boron and u r a n i u m b u r n u p was u n a c c e p t a b l e to the o p e r a t o r s i n c e the add i t i ona l loop handl ing would be tinme con­s u m i n g and po ten t i a l ly in ju r ious to the s p e c i m e n a n d / o r loop .

8) While t h e r e a p p e a r to be no unso lvab le p r o b l e m s a s s o c i a t e d wi th the hot ce l l fac i l i ty de s ign , s p e c i ­m e n handl ing r e q u i r e s e i t h e r a u x i l i a r y f o r c e d gas cool ing o r d e c a y p e r i o d s f a r in e x c e s s of the m i n i m u m s s t a t e d in 6 (c) above . F o r c e d cool ing c o m p l i c a t e s r e m o t e handl ing and en t a i l s the r i s k of spec innen f a i l u r e should cool ing be l o s t .

9) The d e s i g n ob jec t ive of examin ing n iob ium c lad s p e c i m e n s in the hot ce l l f ac i l i ty does not a p p e a r f e a s i b l e s i n c e i t i s doubtful tha t the highly p u r e i n e r t gas a t m o s p h e r e r e q u i r e d could be m a i n ­ta ined in s u c h a l a r g e hot c e l l .

COMMENTARY

Based on the resul ts of the c r i t e r i a studies it was concluded that

achievement of the design objectives could not be predicted with any

high degree of confidence when utilizing the package loop concept

based on the design philosophy of the PW-19 loop. Therefore , it was

recomnnended to the Atomic Energy Commiss ion that the bottom r e ­

entrant loop concept be investigated since it appeared to be a more

prac t ica l and feasible means of fulfilling the design objectives. How­

ever , the AEC te rmina ted r e - en t r an t loop study work before any sub­

stant ial effort was expended and ul t imately elected to suspend any fur­

the r considerat ion of the liquid meta l loops in the ATR.

If in te res t in the liquid meta l reac tor fuel testing p rogram is r e ­

vived, sa t is factory designs may be achieved by (a) utilization of other

package loop concepts , (b) appropr ia te modification of the design ob­

jec t ives while utilizing the PW-19 loop concept, or (c) utilization of an­

other tes t r eac to r while utilizing the PW-19 loop or other package loop

concepts . Additional work would be required to establish feasibility of

these or other s c h e m e s .

X i

Acknowledgement

Ass is tance rendered by the following individuals and organizations in prepar ing the Design Cr i t e r i a and reviewing this document is a c ­knowledged.

United Nuclear Corporat ion

L. Berkowitz D. M, Rosh

Phil l ips Pe t ro leum Company

L. H. Jones

Atomic Energy Commission

J . C. McKinley, IDO R. Silver, COO E. Sowa, ANL-COO

x i i

C O N T E N T S

Page

INTRODUCTION 1

PURPOSE 1

DESIGN OBJECTIVES

3.1 Qualitative Objectives 1

3.2 Quantitative Objectives 2

CONCEPTUAL DESCRIPTION OF THE FACILITY

4.1 General 4 4.2 In-Reactor System 4 4.3 Out-of-Reactor System 6 4.4 Faci l i ty Per formance Summary 7 SYSTEM AND COMPONENT CRITERIA

5.1 General 9 5.2 In-Reactor System 9

5.2.1 Nuclear Cr i t e r i a 10 5.2.2 Thermal Cr i t e r i a 10 5.2.3 Hydraulic Cr i t e r i a 11 5.2.4 Mechanical Cr i t e r i a 12

5.3 Out-of-Reactor Systems

5.3.1 Shielding 21

5.3.2 Mechanical . 25

5.4 Instrumentat ion and Controls

5.4.1 General 31 5.4.2 In-Reactor Instrunnentation 32 5.4.3 Out-of-Reactor Instrumentat ion . . . . . . 32 5.4.4 Data Logging System 33

5.5 Utili t ies 5.5.1 E lec t r i ca l System 34 5.5.2 H i g h - P r e s s u r e Demineral ized Water System . 34 5.5.3 Ventilation 34

5.6 S t ruc tures 35

OTHER REQUIREMENTS 36

INFORMATION TO BE SUPPLIED BY IDO AT START OF TITLE I DESIGN 36

REFERENCES 37

X I 1 1

TABLES Page

I.. The rmal Flux Loop Facil i ty Design Objortives . . . 2 II.. P r i m a r y and Secondary Coolant System

Per formance Summary 7

I I I . P r i m a r y Coolant Thermal Conditions 11

IV . Per formance Capability Summary - Helium: Cooled P r i m a r y Coolant Heat Exchanger 16

V. Per formance Capability Summary - Secondary Coolant to Reactor Water Heat Exchanger . . . . 17

V i : Bas is for Est imation of Activation of Stainless Steel Chips 22

V I I . Stainless Steel Activation in Heat Exchangers of ATR

Liquid Metal Package Loops - 10,000 Hour Irradiat ion . 23

V I I I . Composition of Commerc ia l Grade Helium 24

IX. Basis for Estimation of Secondary Coolant Activation . 24

X. Helium Impurity Activation in ATR Thermal Flux Loop - 10,000 Hour Irradiat ion 25

XI . Design Basis - Thermal Flux Loop Secondary

Coolant System 26

X I I . Secondary Coolant System Compressor Charac te r i s t i cs . 28

X I I I . Secondary Coolant System Compressor Space Require­ments 29

X I V . Secondary Coolant System Heat Exchanger Charac ­t e r i s t i c s 30

FIGURES

(Listed numerical ly following page 37)

1. Reference Test Specinnen

2. Thermal Flux Loop

3. Hydraulic Charac te r i s t i cs of P r i m a r y Coolant System

4. Location of the Thermal Flux Loop in the Advanced

Tes t Reactor Core

5. Thermal Flux Loop In-Core Sections

6. Thermal Flux Loop Heat Exchangers

7. E lementary Flow and Instrumentation Diagram

8. Thermal Flux Loop Maximum Physical Envelope

9. Containment Tube End Closure

10. Suggested Top End Arrangement for the Thermal Flux Loop

11. Secondary Coolant System F i l t e r 12. Faci l i ty Arrangement Plan 13. Faci l i ty Arrangennent Sections

x i V

APPENDIXES

A. Design Objectives and Fixed P a r a m e t e r s

B. Secondary Coolant Systenn

C. Thermal Analysis of Helium Cooled Heat Exchanger

D. Gamma and Neutron Heat Generation Rates

E. Tempera tu re Distribution In In-Core Tubes

F . Secondary Coolant Impuri t ies Activations

G. Sodium System Control and Safety System Cr i te r ia

H. Maximum Physical Envelope

I . P r e l i m i n a r y Physics Per formance of the Reference Fuel Test Specimen

J . Heat Transfe r Control Philosophy

K. System Per fo rmance Charac te r i s t i c s

L. Systenn Hydraulic Charac te r i s t i c s

M. Functional Requirements for Auxiliary Services

N. The rma l Loop Shielding Analysis

•

X V

1. INTRODUCTION

A need has existed for some t ime for facil i t ies in which prototype sodium cooled fuel a s sembl i e s can be i r r ad ia ted under conditions of high the rmal neutron flux. Heretofore , however, none of the available facili t ies pos se s sed the requis i te cha rac t e r i s t i c of high thermal neutron flux which would provide adequate heat generat ion with acceptable flux and power depress ions .

The Atomic Energy Commiss ion recognized that the Advanced Test Reactor (ATR) p o s s e s s e d these cha rac t e r i s t i c s and asked Atomic Power Development Assoc ia tes , Inc. (APDA) to review the ATR design and the design requ i rement s of an in-pile package loop suitable for testing liquid meta l cooled the rmal reac to r fuel a s s e m b l i e s . APDA repor ted (Refer­ence 1) that the ATR could be used as a t es t facility for liquid metal cooled thermal r eac to r fuel e lements and p resen ted a conceptual in-pile loop design for this purpose .

Subsequently, in January 1962, Ebasco Services Incorporated, and its nuclear subcontractor , the Babcock & Wilcox Company, undertook to evaluate this and a l te rna te concepts and to p r e p a r e c r i t e r i a for the Tit les I and I I design of the Thermal Flux Loop Faci l i ty which will be pa r t of the ATR (Reference 9). In this work, they have drawn upon AEC design objectives (Reference 2) and the APDA conceptual design study (Reference 1).

2. PURPOSE

The purpose of this document is to summar ize all work performed and to p resen t the c r i t e r i a which had been es tabl ished pr ior to the t e r m i ­nation of work on the package loop concept for the the rmal flux loop.

3. DESIGN OBJECTIVES

Qualitative and quantitative design objectives have been establ ished by the Atomic Energy Commiss ion in conjunction with the e x p e r i m e n t e r s ' r equ i r emen t s . These objectives form the bas i s for the design of the Faci l i ty .

3.1 Qualitative Objectives

The qualitative objectives for which the Faci l i ty is to be designed include the following:

a) Design of the in-pile package loop c losure to p e r ­mi t p rompt removal , inspection, re inse r t ion and r e -i r rad ia t ion of a tes t specimen.

b) Design of the in-pile package loop to pe rmi t r e ­fueling of the ATR core without ra is ing the loop from the flux t rap in which it is instal led.

c) Re-use of a loop for t es t specimen i r r ad ia t ions .

d) No adverse effects on the per formance of other exper iments within the ATR.

- 1 -

3.2 Quantitative Objectives

The quantitative AEC objectives (Reference 2) a r e summar ized and compared in Table I with objectives which this c r i t e r i a study indicate might be achieved in the Faci l i ty . However, the design of a completely feasible loop to achieve these objectives depends on the sat isfactory solution of the mechanical problems indicated in section 5.0 of this docu­ment.

The design of the reference specimen w^hich is the bas i s of the objectives is i l lus t ra ted in Figure 1.

Table I

THERMAL FLUX LOOP FACILITY DESIGN OBJECTIVES

a) Loop P a r a m e t e r :

In-Pi le Loop Design and Dimensions

AEC and User Exper iment

Design Objectives (Reference 2)

TBD *

Basic Mater ia l s of Construction TBD *

P r i m a r y Coolant Maximum P r i m a r y Coolant

Tennperature, F Operating P r e s s u r e , psig Secondary Coolant Cover Gas Minimum Loop Design Life,Hr.

b) Reference Test Specinnen:

Type of Fuel F i s s i l e Mater ia l F i s s i l e Mater ial Density

g /cc Fuel Type of Spectrum Radial Power Depression

Initial, m a x / m i n Angular Power Variation

Maximum power ratio, f a s t / t he rma l , (0.625 ev cutoff) 20

Fuel Burnup, 10 f iss ions/cc Fuel

Burnup incrennent, 10 f i s s ions /cc Fuel

20

Sodium TBD *

60-150 TBD * TBD * 10,000

Ceramic (UC) Uranium (0.4 - 1.2) TBD *

Thermal 1.5-2.5

Uniform as Poss ib le

0.30

5-15

1-3

Result of Cr i t e r i a Studies

Self-Contained Pack­age, See F igure 2

Stainless Steel, Type 316

Sodium 1400

60 Helium Helium Not Establ ished

Ceramic (UC) Uranium 0.324

Thermal 1.34

1.07 (surface, max / ave) (1.13 max/min)

0.32

5-15{~4,100 t to 1 2 , 3 0 0 t h r @ 1,500 kw)

l-3(~820t to 2 , 5 0 0 W hr @ 1,500 kw)

- 2 -

b) Reference Test Specimen: (Cont'd)

Specimen Power , Generation (Fiss ion) , kw

Average Pin Power , (Fiss ion) , kw/in.

Maximum Pin Power , (Fiss ion) , kw/in .

Number of Pins Specimen Length, in. Pin , Outside Diamete r , in. Fue l , Outside Diamete r , in. Bond Mater ia l Cladding Mater ia l Cladding Thickness , in. Geometry

Spacer Outside Diamete r , in. Test Section Length, ft. Minimum Test Section

Diamete r , in.

c) Nominal Design Conditions

P r i m a r y Coolant Tempera ture at Specimen Outlet (Max), F

P r i m a r y Coolant Tempera ture at Specimen Inlet (Min), F

P r i m a r y Coolant Maximum Axial Tempera ture Dif­ferent ia l , F

Minimum Heat Exchanger Capacity, kw

Maximum Allowable Test Specimen Tennperature at Zero Power (Post I r r a d i a ­tion), F

Hydraulic Charac te r i s t i c s Maximum Flow, gpm Specimen P r e s s u r e Drop

(Maximum Flow), ft. ATR Flux Trap Location Flux Trap Flow Baffle

Inside Diamete r , in. P r i m a r y Coolant will be

"Double Continued"

150-1,500

5

5-7 48 (max) 0.560 0.500 Sodium Stainless Steel 0.020 Triangular Pi tch,

Spiral Spacers 0.090 TBD * 2.35

1,400

800

350

TBD *

1,000

TBD * 180 100

Eas t 5.25

Yes

1,500

4.5

8.5 (hot spot)

7 48 0.560 0.500 Sodium Stainless Steel 0.020 Triangular Pi tch ,

Spiral Spacers 0.090 4 2.425 (max)

1,400

800

350

0.25

1,000

See Figure 3 180 390

Eas t 5.25

Yes

* To be determined by loop des igner ,

t Fuel burnup t imes a re es t imated .

- 3 -

4. CONCEPTUAL DESCRIPTION OF THE FACILITY

4.1 General

The ATR (Reference 3) is a 250 Mwt tes t r eac to r providing expe r i ­mental loop i r rad ia t ion space for the AEC testing p rogram, and is located in a nor thwest extension of the MTR-ETR complex at the National Reactor Testing Station in Idaho.

The reac tor is water cooled, and is fueled with fully enr iched U30g d ispersed in and clad with an aluminum alloy. Boron carbide is included in the fuel ma te r i a l as a burnable poison. The ATR core region cons is t s of a 4-foot high ver t ica l fuel annulus shaped in a serpentine fashion b e ­tween and around nine flux t rap a r e a s in a 3 x 3 square a r r a y , F igure 4. The core region is surrounded by a metal l ic bery l l ium ref lector .

The four flux t rap positions on the c o r n e r s of the square a r e r e f e r r e d to as inner flux t rap posit ions (lobes) since they a r e enclosed by the fuel annulus. Those on the sides of the square a r e exter ior to the fuel annulus and a r e r e f e r r e d to as outer flux t rap posi t ions. The center position is r e f e r r ed to as the center flux t rap . These flux t raps will accommodate nine major tes t loops including six p r e s s u r i z e d water loops, two liquid metal cooled loops, and one gas cooled loop.

The Thermal Flux Loop Faci l i ty to be instal led in the ATR will con­sist of a sodium cooled in - reac to r package loop section and an out-of-reac tor section which includes the helium secondary coolant sys tem, in­strunnentation, and other subsystems requ i red to support the Faci l i ty . These a r e descr ibed briefly below.

4.2 In-Reactor System

The in - reac to r section of the Thermal Flux Loop Faci l i ty , F igure 2, is located in the 50 Mw east outer flux t rap position of the ATR core as in­dicated in Figure 4. It cons is t s of the following:

a) Vert ical in-pile package loop containing a totally enclosed circulat ing sodium system which removes heat from the four foot long reference tes t specimen. Figure 1, by forced circulat ion and t r ans fe r s the heat to flowing helium in the sodium-to-hel ium p r i m a r y heat exchanger.

b) Hel ium-to-water heat exchanger with heliunn flowing inside and reac tor cooling water flowing outside the tubes.

c) Loop extension tube supporting the package portion of the loop in the ATR vesse l and providing passage for the helium secondary coolant supply and re turn piping and power and instrumentat ion leads from the in - r eac to r section out-of-pile.

The in - reac to r portion of the loop which houses the fuel tes t specimen and the sodium p r imary coolant sys tem is s imi lar in concept to that of ( the fast flux loop (Reference 4) and is i l lus t ra ted in Figure 2. This self-contained sys tem is filled with sodium and cons is t s of a pump and v a r i ­able speed e lect r ic motor dr iver , p r i m a r y heat exchanger, sodium

- 4 -

expansion r e s e r v o i r , inner liquid meta l flow tube capable of accepting a four-foot long test specimen, outer flow tube, e lec t r ic h e a t e r s , con­necting piping, and associa ted instrumentat ion and power l eads .

The pump ci rcula tes the p r i m a r y coolant through the fuel tes t spec i ­men and prinnary heat exchanger at controlled flow ra tes adequate to maintain the t empera tu re r i se a c r o s s the tes t specimen. The bottom portion of the loop, Figure 5, which houses the four-foot long reference tes t specinnen is about five and one-half feet long and has an outside d iameter of four inches . The conceptual design of the s t ruc tura l member s in the tes t section of the loop is based on providing double containment of the liquid meta l and rejecting a maximum portion of the total heat gener ­ated in this section d i rec t ly to the reac tor cooling water flowing downward outside the loop.

Above the test section of the loop is the p r i m a r y heat exchange which t rans fe r s heat from the flowing sodium to the flowing secondary coolant hel ium. The design a r r angemen t of the heat exchanger shown in Figure 6 is based on providing a maximum of heat t ransfer surface within the available space . The cen te rmos t tubes ca r ry sodiunn upward from the tes t section to the suction of the pump and the outermost tubes d i rec t the sodium from the pump discharge downward and back into the test section. Secondary coolant helium enters at the top of the heat exchanger, flows downward on the shell side para l le l to the axis of the tubes in both the inner and outer tube sect ions , and leaves the heat exchanger at the bottom.

Immediately above the heat exchanger a re flow baffles which di rect the sodium to and from the pump. Resis tance type e lec t r ic hea te r s jacketed in s ta inless s teel sheaths a r e located in the pump discharge section. These hea te r s a r e required to maintain the sodium in the molten condition during var ious loop opera t ions .

The pump impel ler is located above the heater region of the loop. Conceptually it is of the mixed flow design and will have a maximum ca­pacity of 180 gpm at 500 feet head when operating at approximately 10,000 rpm. Directly above the pump impel ler is the sodium sump, sized to accommodate the the rmal expansion of the sodium during i ts r i se in tempe ature from the filling t empera tu re to the maximum operating tennperature. The sodium in the sump a lso se rves as a heat sink to prevent overheating the pump drive nnotor and bea r ings . Elect r ic hea te rs may be required in the sump to prevent the sodium from freezing. The pump dr ive , located direct ly above the sump, is a var iab le speed motor controlled by varying the frequency of the applied e lec t r ic power . Shielding mus t be supplied above and around the sump to pe rmi t local , manual , making and breaking of heliunn and power connections atop the loop and to reduce the total dose received by the motor during 10,000 hours of operation to an acceptable level . A discussion of the p re l iminary shielding analysis is given in Appendix N.

The secondary coolant helium passes fronn the out -of - reac tor section into the top of the loop through a 1-3/4 inch inside d iameter inlet line inside the extension tube. The helium line is then routed down around the outside of the loop p r e s s u r e tube into the top of the p r i m a r y heat exchanger . F rom the heat exchanger the helium flows upward outside the loop proper through 28 one-half inch d iameter tubes approxinnately

- 5 -

eight feet long which fo rm the h e l i u m - t o - r e a c t o r w a t e r hea t e x c h a n g e r . T h e s e tubes a r e h e a d e r e d a t the top of the loop in to a s ing le 2 - 5 / 8 inch i n s ide d i a m e t e r h e l i u m out le t p ipe which p a s s e s up t h rough the e x t e n s i o n tube and thence to the o u t - o f - r e a c t o r s y s t e m . A baffle s u r r o u n d i n g the h e l i u m - t o - r e a c t o r - w a t e r h e a t e x c h a n g e r t ubes d i r e c t s the r e a c t o r w a t e r flow downward a r o u n d the ou t s ide of the t h e r m a l flux loop and in to the flux t r a p baf f le . T h i s a r r a n g e m e n t u t i l i z e s the ex i s t ing p r e s s u r e d r o p a c r o s s the r e a c t o r c o r e to p r o v i d e f o r c e d c i r c u l a t i o n of the r e a c t o r w a t e r , t hus m a x i m i z i n g the hea t e x c h a n g e r ' s h e a t t r a n s f e r r a t e .

The r e m o v a l of the t h e r m a l flux loop f rom the A T R m a y be a c c o m ­p l i s h e d a f t e r a d e c a y t i m e of l e s s than one h o u r , a p e r i o d suff ic ient to p e r m i t r e m o v a l of decay hea t by n a t u r a l convec t ion to w a t e r . The r e ­m o v a l p r o c e d u r e c o n s i s t s of (a) r a i s i n g the loop f rom the c o r e , (b) p l ac ing i t in the A T R r e a c t o r v e s s e l d r o p t u b e , and (c) l o w e r i n g i t in to the A T R work ing c a n a l . Dur ing the loop r e m o v a l o p e r a t i o n s , s e c o n d a r y coo lan t h e l i u m flow i s not r e q u i r e d . H o w e v e r , the p u m p m u s t cont inue to o p e r a t e to p r e v e n t s p e c i m e n o v e r h e a t i n g , and t h e r m o c o u p l e and e l e c t r i c p o w e r c o n n e c t i o n s m u s t be nnaintained th roughou t the r e m o v a l of the l o o p . T h e s e o p e r a t i o n s and t h e i r a s s o c i a t e d p r o b l e m s a r e d e s c r i b e d in d e t a i l in IDO-24043 ( R e f e r e n c e 5).

4.3 Out -o f -Reac to r S y s t e m

C o n c e p t u a l l y , the o u t - o f - r e a c t o r s e c t i o n of the Thernna l F l u x Loop F a c i l i t y supp l i e s h e l i u m to the p r i n n a r y h e a t e x c h a n g e r in the i n - r e a c t o r s e c t i o n a t cons t an t t e n n p e r a t u r e and p r e s s u r e , and o v e r a r a n g e of flow r a t e s a d e q u a t e to nnaintain d e s i r e d t e m p e r a t u r e l e v e l s in the i n - r e a c t o r s e c t i o n . It i nc ludes the fol lowing:

a) O u t - o f - r e a c t o r p o r t i o n of the h e l i u m s e c o n d a r y coolan t sys tenn .

b) B y p a s s heliunn d ry ing s y s t e m - u s e d to m a i n t a i n the nno i s tu re con ten t of the s e c o n d a r y coo lan t a t a c c e p t a b l e l e v e l s .

c) Heliunn supply systenn - u s e d to fill the s y s t e m and to p r o v i d e nnake -up .

d) V a c u u m systenn - u s e d to e v a c u a t e the s y s t e m be fo re it is f i l led wi th the s e c o n d a r y coo l an t .

e) I n s t r u m e n t a t i o n .

The o u t - o f - r e a c t o r p o r t i o n of the s e c o n d a r y coolan t systenn c o n s i s t s of f i l t e r s d o w n s t r e a m and u p s t r e a m of the i n - r e a c t o r s e c t i o n , a h e l i u m to h i g h - p r e s s u r e d e m i n e r a l i z e d w a t e r (HDW) systenn hea t e x c h a n g e r , m a i n and a u x i l i a r y c o n n p r e s s o r s , and a s s o c i a t e d v a l v e s and p ip ing . T h i s systenn and the s u p p o r t i n g d r y i n g , heliunn supply , v a c u u m and i n s t r u m e n ­t a t ion s y s t e m s , i l l u s t r a t e d s c h e m a t i c a l l y in F i g u r e 7, wi l l pernni t the e x p e r i m e n t e r to s e l e c t and c o n t r o l o p e r a t i n g cond i t ions ove r a wide r a n g e .

- 6 -

All ou t -of - reac tor section equipment, with the exception of the ^compressors , is located ei ther within or adjacent to Cubicle IB in the f i rs t basement of the ATR building. The c o m p r e s s o r s , because of their s ize, cannot be located within the cubicle and, therefore , have to be located together with those for the fast flux loop (Reference 4), in a c o m p r e s s o r vault outside the south wall of the ATR building. The vacuum sys tem will a lso serve the fast flux loop and will be located within Cubicle lA along the wall adjacent to Cubicle IB. Helium supply, ins t rumentat ion, and control sys tems will be located outside of and adjacent to the cubic les .

4.4 Faci l i ty Per fo rmance Sunnmary

The per formance of the prinnary and secondary coolant sys tems for a Thermal Flux Loop containing the reference tes t specimen is summar i zed in Table I I .

Table I I

PRIMARY AND SECONDARY COOLANT SYSTEM

PERFORMANCE SUMMARY

(Basic• Reference Specimen in 50 Mw ATR Outer Flux Trap Position)

P r i m a r y Coolant System

Coolant Flow Rate , gpm @ 1214 F Pump Total Developed Head, ft. Flow Rate Through Specimen, gpm Leakage Flow Rate (Specimen Bypass) , gpm P r e s s u r e Drop Through Specimen, ft. Tempera tu re , Mixed Mean, at Specimen Inlet, F Tempera tu re , Mixed Mean, at Specimen Outlet, F Tempera tu re , P r i m a r y Heat Exchanger Inlet, F Tempera tu re , P r i m a r y Heat Exchanger Outlet, F P r i m a r y Coolant Axial Tempera tu re Difference, F Pump Inlet Tempera tu re , F

Secondary Coolant System

Coolant System Flow Rate, Ib /h r Tempera tu re , P r i m a r y Heat Exchanger Inlet, F Tempera tu re , P r i m a r y Heat Exchanger Outlet, F Tempera tu re , at Outlet of In-Reactor Section, F P r e s s u r e , at Inlet of In-Reactor Section, psig P r e s s u r e , at Outlet of In-Reactor Section, psig

Sodium 140 296 135 5 245 1,050 1,400 1,400 1,079 350 1,214

Helium 12,000 140 501 350 600 500

- 7 -

Heat Balance

Specimen F i s s ion Power , kw 1,500 Test Section Gamma Heating (Including Specimen,

Sodium and Structure) , kw 208.7 Gamma Heating, to P r i m a r y

Heat Exchanger Outlet to Tes t Specimen Inlet, kw 228.6

Heat Input F r o m Sodiiim Pump, kw 12.0

Total Heat Input, kw 1,949.3

Heat L o s s e s F r o m P r i m a r y Coolant to ATR Water in In-Core Section, kw 376

Heat T rans fe r r ed F r o m P r i m a r y Coolant to Secondary Coolant, kw 1,573.3

Total Heat T rans fe r r ed F r o m P r i m a r y Coolant, kw 1,949.3

Heat Trans fe r red F r o m Secondary Coolant to ATR "Water, kw 659.2

Heat T rans fe r r ed F r o m Secondary Coolant to HDW, kw 914.1

Total Heat Trans fe r red F r o m Secondary Coolant, kw 1,573.3

- 8 -

5. SYSTEM AND COMPONENT CRITERIA

5.1 General

Where possible, c r i t e r i a for the Titles I and I I design of the Thermial Flux Loop a r e presented in this section in sufficient detail to give the r eade r an understanding of how they were established, why par t icu la r c r i t e r i a were selected and what problems and a l te rna t ives were considered and st i l l remain to be resolved and evaluated.

5.2 In-Reactor System

The initial conceptual design of the rmal flux package loop p repared by APDA, Reference 1, utilized a controllable level of liquid m e r c u r y to vary the effectiveness of heat t ransfer surface in t ransfer r ing heat from the p r i m a r y coolant sodium to the reactor cooling water . In this manner the p r i m a r y coolant sodium tempera tu res could be maintained at the levels required by the exper iment .

The possible resu l t s of a leak of m e r c u r y into the reactor coolant water were considered and deenaed to be undesirable . Therefore, the design c r i t e r i a work included the investigation of (a) substituting another fluid for m e r c u r y in the var iable liquid level control schenae and, if a sat isfactory substitute fluid could not be found, (b) prepar ing an a l ternate scheme that would pe rmi t operating the loop over the des i red control range (see Reference 9). The resu l t s of this investigation. Appendix J , indicated that no sat isfactory substitute fluid could be found to be used with the var iable level control scheme and that a flowing secondary cool­ant, helium, could be used to give sat isfactory heat removal control .

Therefore , based on utilizing helium as the secondary coolant, the i n - r e a c t o r portion of the Thermal Flux Loop, F igure 2, was conceptually designed to comply with the qualitative and quantitative design objectives outlined in Sections 3.1 and 3.2. In addition, the design at temped to max i ­mize heat t ransfer d i rec t ly to the reac tor wate r .

The configuration shown in F igure 2 nei ther provides for nor reflects a complete solution to all the design problems associa ted with the package loop concept since work was terminated p r io r to completion of the design. Solutions to the following design problems a r e not reflected in the concep­tual design presen ted :

a) Consideration of double containment for high p r e s s u r e helium coolant which would prevent the inadvertent escape of gas into the reac to r cooling water in the event of a leak.

b) Completely adequate design solution to the problems created by differential thermal expansion.

c) Necessa ry shielding to prevent interference with other reac tor operat ions and to l imit the total in­tegrated dose received by the pump dr ive motor and bear ings .

- 9 -

d) A sat isfactory pump-dr ive-motor and bearing design. e) La te ra l support of loop within the r eac to r .

It should be emphasized that additional study and analysis a r e required in o rde r to es tabl ish the complete feasibil i ty of this conceptual design with­in the l imitat ions imposed by the design objectives and safety considerat ions .

5.2.1 Nuclear Cr i t e r ia

Those c r i t e r i a set forth in Table I, Section 3.2, that apply to the nuclear design of the sodium cooled thermal loop a r e :

Type of Fuel F i s s i l e Mater ia l F i s s i l e Mater ia l Density g/cc-fuel Type of Spectrum Radial Power Depress ion, initial,

Max/Min Angular Power Variation Maximum Power Ratio, f a s t / t he rma l

(0.625 ev cutoff) Fuel burnup, 10^0 f i ss ions /cc- fue l Burnup Increment, 10 f i ss ions /cc- fue l Specimen Power (fission), kw Average Pin Power (fission), kw/in. Maximum Pin Power (fission), kw/in. Number of Pins Specimen Length, in. Pin Outside Diameter , in. Fuel Outside Diameter , in. Bond Mater ia l Cladding Mater ia l Cladding Thickness, in. Geometry

Spacer Outside Diameter , in.

Ceramic (UC) Uranium 0.4-1.2 Thermal 1.5-2.5

Uniform as possible 0.30

5-15 1-3 150-1500 5 9 5-7 48 0.560 0.500 Sodium Stainless Steel 0.020 Tr iangular Pi tch,

Spiral Spacers 0.090

Studies ca r r i ed out for the reference specimen in a 50 Mw single lobe model indicate that the above nuclear c r i t e r i a can be met reasonably well without significantly affecting other exper imenta l a r e a s within the r eac to r . The detai ls of these studies a r e presented in Appendix I.

5.2.2 Thermal Cr i te r ia

Those design objectives set forth in Table I, Section 3.2, that apply to the the rmal design of the loop a r e :

In-Pi le Loop Design

P r i m a r y Coolant Tempera tu re at Specimen Outlet (Max), F

P r i m a r y Coolant Tempera ture at Specimen Inlet (Min), F

P r i m a r y Coolant Maximum Axial Tempera tu re Differential, F

Self-Contained Package

1400

800

350

- 10 -

In addition, the nuclear objectives and the specimen configuration must be considered.

The in -core section has been designed to maximize heat t r ans fe r d i rec t ly to the r eac to r cooling water from the p r i m a r y coolant sodium.. The sodium-to-helium heat exchanger was designed to remove the r e s t of the heat gener ­ated when operating within the design conditions. Table I I I s u m m a r i z e s the sodium thermal conditions in the loop. It can be seen that the the rmal c r i t e r ia have been met . Detailed discussions of these c r i t e r i a a r e included in Appen­dixes C, D and K.

Table I I I

PRIMARY COOLANT THERMAL CONDITIONS

Specimen Power , Kw 1500 Tempera tu re at Specimen Inlet, F 1050 Tempera tu re at Specimen Outlet, F 1400 Tempera tu re at P r i m a r y Heat 1400

Exchange Inlet, F Tempera tu re at P r i m a r y Pump, F 1214 Tempera tu re at P r i m a r y Heat 1079

Exchanger Outlet, F Tempera tu re at Test Section Inlet, F 1079 Tempera tu re at Specimen Inlet, F 1050

5.2.3 Hydraulic Cr i te r ia

Those design objectives set forth in Table I, Section 3.2, that apply to the hydraulic design of the loop a r e :

In-Pi le Loop Design

Minimum Test Section Diameter , in. P r i m a r y Coolant Tempera tu re at

Specimen Outlet (Max), F P r i m a r y Coolant Maximum Axial

Tempera tu re Difference, F Maximum Flow, gpm Specimen P r e s s u r e Drop

(Maximum Flow), ft

In addition, the nuclear objectives and the specimen configuration must be considered.

Under the specimen design power and tempera ture conditions, the sodium flow ra te is 140 gpm and the p r e s s u r e drop a c r o s s the test specimen is 245 feet. If sodium flows at 180 gpm, the specimen p r e s s u r e drop is 390 feet. In both cases the p r e s s u r e drops a r e considerably g rea t e r than that permit ted by the design object ives. A more detailed discuss ion of hydraulic performance is given in Appendix L.

It is apparent that, in order to meet the design objective of 100 foot specimen p r e s s u r e drop at 180 gpm, ei ther the specimen configuration must be changed or some of the other design objectives, such as specimen power or thermal conditions.

Self - Contained Package

2.35 1400

350

180 100

- 11 -

5.2.4 Mechanical Cr i te r ia

The feasibil i ty of the conceptual design for the Thermal Flux Loop baWl on the package concept had not been completely established when work was te rminated . This section p resen t s the background information that had been developed and outlines the problem a r e a s which requi re fur ther study.

5.2.4.1 Space E nvelope

The space envelope. F igure 8, for the en t i re i n - r e a c t o r port ion of the loop was developed as a resul t of the requi rements imposed by the design objectives, and par t i cu la r ly the physical boundary l imitat ions of the ATR and the other exper iments contained therein .

The in -core section is the bot tommost portion of the loop. It extends downward through the flux t rap from the top active fuel line of the r eac to r fuel e lements (elevation 82 feet 0 inches) for a distance of 4 feet 5 inches and upwards from the top of the active fuel line for a distance of 11 inches making the total length 5 feet 4 inches . This section has a 4 inch outside d iameter and contains the four foot long reference fuel test specimen. The d iameter was fixed as a resul t of the minimum water thickness required between the outside of the loop and the flux t rap baffle to attain the specified fast-to-thermal-power ra t io .

The section containing the sodium-to-hel ium heat exchanger, the e l ec ­t r ic hea ter (s ) , the pump and the expansion r e s e r v o i r is the next section above the in -core section. It extends from elevation 82 feet 11 inches to elevation 88 feet 6 inches. The outside d iamete r of this section is 5.125 inches. Both the outside d iameter and the elevation to which this section extends above the core a r e dictated by the requ i rement that refueling of the ATR core should be accomplished without moving the loop. The refueling operation requi res that the outside d iameter of this section, for 180 degrees of per iphery , be slightly smal le r than the inside d iamete r of the flux t rap baffle from which the ATR fuel e lements are supported. Maintaining this d iameter up to elevation 88 feet 6 inches p e r m i t s the fuel e lements to be lifted s traight up for a distance of 6-1/2 feet above the top active fuel line before any la te ra l movement is impar ted to the element . The section from elevation 88 feet 6 inches to 9 1 feet 10 inches contains the e lec t r i c motor , the motor shielding and a portion of the biological shielding. This section has an outside d iameter of 8-1/2 inches which is a reasonable maximum that will not in terfere with other equipment or operat ions in the r eac to r .

The uppermost section (from elevation 91 feet 10 inches to elevation 93 feet 3 inches) is approximately 17 inches long and 5-15/16 inches in d iamete r . The top 10 inches of this section must fit through the six inch penetrat ion holes in the top head c losure plate of the reac tor ves se l . Ten inches is the minimum length required to make or break the connections at the top of the loop p r io r to loop removal or installation in the r eac to r . The remaining seven inches is taken up with the helium t ransi t ion p a s ­sages .

The he l ium- to - r eac to r water heat exchanger is located on the east side of the loop. It is pa ra l l e l to and extends approximately seven i n c h e s ^ ^ from the ver t ica l axis of the loop. It s t a r t s at the bottom of the sod ium-to-helium heat exchanger and te rmina tes at the top of the pump motor sect ion.

- 12 -

Since the heat exchanger is located in a 180 degree f lared a r ea above the beryl l ium reflector , it does not interfere with the refueling operat ions . The maximum radial pro t rus ion of this section from the loop ver t ica l cen te r -line is 7-1/2 inches.

Based on these various l imitat ions the total loop length is 15 feet 11 inches . The maximum length that can be moved horizontally within the reac to r v e s s e l to the drop tube is approximately 16 feet. Fu r the r detailed discussions of the loop space envelope a re presented in Appendix H.

5.2.4.2 Pump, Motor, and Sump

A pre l iminary investigation into the feasibility of developing a pump for the thermal loop was made . Several pump manufacturers were called in to study the problems involved in developing a pump-dr ive motor com­bination that would:

a) Pump 180 gpm of sodium to a head of 500 feet at a pumping t empera tu re of 1215 F ,

b) Turn the flow 180 degrees ,

c) Fi t into the available space envelope, see Section 5.2.4.1.

d) Dissipate the heat generated in the motor .

e) Operate for 10,000 hours without maintenance.

f) Have a continuously var iable speed range from 10% to 100%,

g) Operate in a sodium vapor-helium environment.

The investigations covered hydraulic , e lec t r ica l , and bearing design p rob lems . It was the general consensus of the pump manufacturers con­tacted that a pump using two or three stages of mixed flow impel le r s would be required to provide the specified hydraulic per formance . The impel le r s would have inside and outside d iamete r s of approximately 1-1/2 inches and 3 inches respect ively . The full load speed would range between 10,000 and 12,000 revolutions pe r minute . A cover gas p r e s s u r e of 60 psig would be more than ample to supply the required net positive suction head.

The pump-dr ive motor would be from 8 to 24 inches long and the nominal power rating would be between 30 and 50 horsepower depending on the efficiency.

A cursory study of the problems involved in cooling the motor revealed that the heat generated within the motor could be dissipated by conduction to the reac tor water . The following assumptions were made: (a) motor operating in helium a tmosphere , (b) fluted double containment around the motor , (c) no thermal r e s i s t ance between the walls of the double contain­ment, and (d) velocity of the water outside the double containment equal to one foot per second. While assumption (c) is not s t r ic t ly valid, previous heat t ransfer work indicated that the contact p r e s s u r e between the walls is high enough such that the contact r e s i s t ance is very low compared to the other res i s t ances to heat t rans fe r .

- 13 -

A possible problem a r ea in the design of the pump motor combination is the cr i t ica l speed of the pump drive-shaft. Since the unit must be capable of operating over a wide range of speeds the rotating assembly should be ^ ^ designed such that the pump would always operate below the f i rs t cr i t ica l ^ P speed.

Since the space l imitat ions within the loop preclude the possibi l i ty of direct ly measur ing the liquid metal flow ra te , an indirect means will be required. A full size dimensional mock-up of the loop liquid metal system will be used to determine the flow rate of water as a function of input f r e ­quency and power. This data by application of "Reynold's Analogy" will permi t the determinat ion of the liquid meta l flow ra t e .

The problem requiring the most effort to solve, relevant to the develop­ment of a sat isfactory pump, appears to be concerned with the bear ings . This conclusion was drawn as a resul t of studying the requi rements placed upon the pump and consequently upon the bear ings . Briefly these requirenaents a r e (a) 10,000 hours of operation without maintenance, (b) high speed (10,000 to 12,000 rpm) as well as var iable speed operation, (c) high t empera tu re , (d) high level radiation environment, (e) occasional stopping and s tar t ing, and (f) operate in a liquid metal vapor environment.

The expected configuration of the pump and motor is s imi la r to that used in the Fas t Flux Loop, Reference 4, in which the pump is considerably below the motor , i .e . , the impel ler is mounted on the lower end of the shaft while the motor rotor is at the upper end. The liquid level of the sodium in the sump is between the pump and the motor .

It is expected that a combination rad ia l -and- th rus t bearing would be immediately above the motor and a radial bearing would be immediately beneath the motor, Reference 4. These two bear ings would operate in the helium and sodium-vapor a tmosphere . A second radial bearing might be needed in the vicinity of the impel ler and would, of necess i ty , operate in and be lubricated by liquid sodium.

The lower radial bearing (at the pump) is not considered to be an un-solvable problem. Liquid metal lubricated bear ings have been used in pumps before. If forced circulation of the lubricant is required, a small auxil iary impeller mounted on the main shaft would provide the neces sa ry circulat ion. During Title I, a complete investigation of design and ma te r i a l s would be performed.

The two upper bear ings in the Fas t Flux Loop, Reference 4, were tight-fitting and grease lubricated. Similar bear ings could be used in the thermal loop, but it is expected that their service lives would not extend to the design objective of 10,000 hours . Alternate bearing designs might u t i ­lize liquid metal or gas as the lubricant. The difficulties involved in apply­ing these a l ternate designs involve not only the factors of speed, t empera ture and environment, but a lso space l imitat ions and sealing r equ i r emen t s .

If the bearings a r e liquid metal lubricated, the motor lower bearing could be s imi la r to that used a t the impel le r . An auxi l iary impel ler could circulate sodium from the sump through the bearing and back to the s u m p . ^ ^ The motor upper bearing might be supplied with sodium from the sump ^m through the main shaft, but there is no room for installation of a line c a r r y ­ing sodium from the bearing back to the sump. Preventing the sodium from

- 14 -

leaking out of the bearing and down onto the motor is another problem. In addition, the radioactive sodium passing through the motor would produce radiat ion p rob lems .

Gas- lubr ica ted bear ings have also been used in cer ta in applicat ions, such as gas c o m p r e s s o r s . As applied to the the rma l loop motor bear ings , however, a major problem concerns sealing the clean lubricating gas from the contaminated (sodium vapor and fission products) gas that forms the motor environment . It is not believed that a seal exis ts today that could be guaranteed to prevent leakage between the two gas envi ronments .

If the lubricat ing gas is at higher p r e s s u r e than the sump, the sump mus t be vented to prevent gas p r e s s u r e build-up in the loop. If the sump is at higher p r e s s u r e than the lubricant , sump cover gas will bleed off. In e i ther case , sodium vapors and fission products would pass from the package loop into the external sys tems , producing radiat ion p rob lems .

It might be possible to pump the sump cover gas through the bear ings in a closed system within the loop. However, a near ly constant gas p r e s ­sure at the bear ing is usually requi red ; a main-staff mounted gas impel ler would not produce this constant p r e s s u r e since the pump motor mus t oper­ate over a wide speed range . In addition, it is not known whether or not the contaminated gas would be a sat isfactory lubricant .

It is apparent that, in order to a r r i ve at a feasible p u m p - m o t o r - b e a r ­ing design,considerable engineering, r e s e a r c h , and development work will be requ i red in Title I,

5,2.4,3 Sodiiam-to-Helium P r i m a r y Heat Exchanger

The conceptual design of the heat exchanger was developed as a resu l t of attempting to satisfy the following conditions:

a) Liquid sodium under conditions stated in the design objectives (Reference 2),

b) Helium coolant at 600 ps i .

c) Outside d iameter fixed at 5,125 inches to pe rmi t ATR refueling without moving the loop (see Section 5,2,4.1).

d) Maximize the heat re jec ted direct ly to r eac to r cooling water in o rde r to reduce the size of secondary coolant sys tem.

e) P a s s a g e s on the sodium side to be as la rge as possible to p r e ­vent plugging, enhance drainage, and provide minimum p r e s s u r e drop.

f ) Maintain the pump suction t empe ra tu r e as low as poss ible .

g) Consider t he rma l expansion due to hot and cold leg t empera tu re differences.

h) Allow passage of the leads from e lec t r ic hea te r s and specimen ins t rumentat ion.

- 15 -

The conceptual design of the p r i m a r y heat exchanger shown in F igure 6 at tains the the rmal per formance . Table IV, specified by the design objectiv^^. However, from Figure 6 it can be seen that the inner and outer section of t h ^ ^ heat exchanger a r e connected by the concentr ic tubes that form the the rmal b a r r i e r between the sodium s t r e a m s entering and leaving the test section. Since these two tubes a r e operating at different t empe ra tu r e s , the amount each expands will be different. If these tubes a r e fixed as shown, the induced s t r e s s e s will be excess ive . These s t r e s s e s could be reduced by redesigning the heat exchanger so that the differential growth of the the rmal b a r r i e r will be absorbed in the heat exchanger tubes. Additional analys is will be required during Title I to de termine the complete feasibil i ty of the heat exchanger con­cept presented .

Table IV

PERFORMANCE CAPABILITY SUMMARY HELIUM COOLED PRIMARY COOLANT HEAT EXCHANGER

I I I*

Load, kw P r i m a r y Coolant Inlet Tempera tu re , F P r i m a r y Coolant Outlet Tempera tu re , F P r i m a r y Coolant Flow Rate at Pumping

Tempera tu re , gpm Pumping Tempera tu re , F Secondary Coolant Inlet Tempera tu re , F Secondary Coolant Outlet Tempera tu re , F Secondary Coolant Flow Rate, Ib /hr Secondary Coolant In-Reactor P r e s s u r e

Drop, psi

*Minimum heat exchanger capacity within the design objective l imi ta t ions .

5.2.4.4 Hel ium- to-Reac tor Water Heat Exchanger

In o rder to reduce the design requ i rements of the ou t -of - reac tor equip­ment and consequently the HDW sys tem, the he l i um- to - r eac to r water heat exchanger, shown in Figure 6, is designed to t ransfer a maximum amount of heat di rect ly to the reac tor cooling water . The design is based on including as much heat t ransfer surface as possible in the l imited space available (See Section 5.2.4.1). The calculation assumed that a forced convection flow of 200 gpm was possible by using the core p r e s s u r e drop to dr ive the reac tor water down through a shroud which contains the 28 half-inch d iamete r tubes . This assumption will be checked during Title I. The per formance of this heat exchanger is shown in Table IV,

1,573 1,400 1,075

140

1,214 140 501

12,000 100

0,25 800 800 (Approx) 180

800 _ -

Zero _

- 16 -

Table V PERFORMANCE CAPABILITY SUMMARY

SECONDARY COOLANT TO REACTOR WATER HEAT EXCHANGER*

Load, kw 659 Secondary Coolant Inlet Tempera tu re , F 501 Secondary Coolant Outlet Tempera tu re , F 350 Secondary Coolant Flow Rate, Ib /h r 12,000 Reactor Water Inlet Tempera tu re , F 130 Reactor Water Outlet Tempera tu re , F 153 Reactor Water Flow Rate, gpm 200

* This performance is associa ted with the p r imary heat exchanger p e r f o r m ­ance presented as Case I in Table IV.

5.2.4.5 E lec t r i c Heater Capacity

During the t ime that the loop is submerged in water and the fuel test specimen is not producing sufficient power to prevent freezing of the sodium, a cer ta in anaount of heat must be added to make up for the heat loss from the loop. This loss was calculated to be about 40 kw with 15 kw being supplied by the pump when operating at full speed. The remaining 25 kw would be sup­plied by e lec t r ic immers ion hea te r s located in the sodium flow passage between the sodium-to-hel ium heat exchanger and the pump, see Figure 2. In al l probabil i ty e lec t r i c hea te r s will a lso be required in the sodium sump to p r e ­vent freezing. E lec t r i c power leads will be sized in Title I to accommodate the maximum power requi red . It is important that the d iamete r of these leads be kept to a minimum,

5.2.4.6 Loop Life

The ability of the loop to mee t the design objectives requir ing 10,000 hours life is influenced by the following fac tors :

1) The ability of the pump bear ings and lubr icants , and the naotor windings and insulation to operate in the loop environnaent for 10,000 hours .

2) The extent to which sodium oxide will build up in the loop and the effect of this build-up on loop operat ion.

3) The physical , chemical , and nnetallurgical changes that may take place in the loop m a t e r i a l s , i .e . , type 316 s ta inless s teel , over a 10,000 hour period under an t i c ­ipated loop operating conditions.

4) The combined p r e s s u r e and thermal s t r e s s e s in c r i t ­ical a r e a s under steady state and t ransient condi­tions, including an es t imated 162 thermal cycles from maximum tempera tu re to reac tor water t empera tu re .

- 17 -

The work performed up to the t ime the package loop concept for the the rma l loop was te rminated had not es tabl ished a sat isfactory solution ^ ^ any of the above prob lems . Additional study and r e s e a r c h and deve lopme^^ is requi red before the feasibility of a 10,000 hour design life can be es tab­lished,

5.2.4.7 Loop Re-Use

In o rder to r e - u s e the loop for i r rad ia t ion of a previously i r r ad ia t ed specimen, the loop must be designed for remote removal and instal lat ion of the specimen, for remotely making and breaking various service con­nections, for remote filling with sodium, and for remote purification of the sodium.

a) Loop End Closure

In o rder to remove and r e in se r t a test specimen in the loop a removable and replaceable bottom end c losure is required. A suggested design for such a c losure is shown in Figure 9. The details of and the problems associa ted with the removal and rep lace ­ment of this c losure and removal and inser t ion of the test specimen are descr ibed in Pa r t I of IDO-24043, Reference 5.

b) Service and Instrximentation Connections

In order to provide the various serv ices required to fill, operate and monitor the the rmal flux loop a great number of connections a re required . Due to the l imi ta­tions on available space it i s necessa ry to locate these connections on the top surface of the loop.

Connections a r e provided for the following se rv i ces : a) flushing and filling the loop with sodium, b) es tab­lishing helium buffer and cover gas regions, c) e lec ­t r i c a l power to the pump dr ive-motor and sodium hea t e r s , d) instrumentat ion leads for thermocouples , leak de tec tors , and sodium level indicators , and e) secondary coolant inlet and outlet l ines .

The design objectives of loop r e - u s e and r e - i r r ad i a t i ons of tes t specimens require that these connections be r e ­motely r e -used a number of t imes in the hot cell , A conceptual design of the loop top end is shown in F ig­ure 10. Mock-up tes t s will be required to demonstra te the utility of this scheme or that of a l t e rna tes . The problems associated with making and breaking these connections remotely in the hot cell are descr ibed in Pa r t I of IDO-24043, Reference 5,

c) Sodium Filling and Purification

The magnitude of the problem of sodium oxide build-up is one which can only be resolved by operational t e s t s of the loop. Since it does not appear that a suitable sodiiom

- 18 -

purification systena can be installed within the loop, con­nections have been provided at the top of the loop to p e r ­mit circulat ion of the sodium in a remotely operable external purification systeni . The satisfactory operation of this purification sys tem will be dependent on:

1) The ability to c i rcula te sodium through the smal l t he rmal loop fill l ines and internal passages with­out plugging.

2) The ability to heat the the rmal loop to at least 400 F while cooling the in-pile section of the loop.

3) The ability to remote ly open, connect to, and r e s e a l the the rmal loop fill l ines .

4) The longevity of the the rmal loop liquid level de t ec ­to rs that a r e needed to prevent inadvertent ove r ­filling.

5) The impuri ty level of the sodium within the the rmal loop pr io r to purification.

A sat isfactory solution of these problems will require the performance of full scale mock-up t e s t s . A detai l descr ip t ion of the proposed pur i f ica­tion sys tem is presented in IDO-24043, Reference 5.

5.2.4.8 Loop Shielding Requirements

Activated sodium is the major contributor to the high dose r a t e s from the the rmal flux loop during handling opera t ions . Since the loop sump is the highest point in the loop which contains activated sodium, it will be n e c e s s a r y to supply shielding within and possibly around the per iphery of the loop sump. This shielding is required to pernait reasonable loop handling p rocedures and to prevent excessive radiat ion damage to the loop motor windings, lubr icants , and bea r ings . It has been es t imated that the total integrated dose received by the motor windings should be l imited to 109 r a d s . P r e l i m i n a r y calculat ions, see Appendix N, indicate that approx­imately 3.5 inches of tungsten shielding instal led direct ly above the sump will satisfy this r equ i rement based on a loop operating life of 10,000 hours . Additional shielding will be required above the motor to prevent gamma s t reaming during handling opera t ions . It should be noted that these calcu­lations do not consider the dose rate to or from the adjacent fast flux loop pump-sump region which may establ ish the need for additional shielding around the pe r iphery of the motor region.

The above calculations a lso indicated the tungsten shield may require auxi l iary cooling to diss ipate the gamma heat generated. Additional ana l ­ysis and study of these shielding and gamma heat problems is required since the feasibil i ty of providing appropr ia te shielding in and around the loop sump had not been establ ished when the work on the package loop concept was te rminated .

5.2.4.9 Double Containment of High P r e s s u r e Gas

Cri t ical exper iments presen t ly being conducted for the ATR indicate that there a r e regions of the core which have positive void coefficients of

- 19 -

react ivi ty. Since the conceptual design of the thermal flux loop uti l izes high p r e s s u r e helium (greater than r eac to r operating p r e s s u r e ) as the 4 secondary coolant, an i n - r eac to r leak or rupture of this system would r e ­lease high p r e s s u r e gas into the reac to r p r i m a r y coolant. Gas bubbles could be formed and ca r r i ed through the reac tor core causing voids and their associa ted positive react ivi ty effects.

Provis ions to ensure against helium leakage were not included in the conceptual design of the the rmal flux loop since the extent of these pos i ­tive void coefficients of react ivi ty had not been defined.

The leakage of helium into the reac to r p r i m a r y coolant and subse ­quently into the core could be prevented ei ther by use of double contain­ment s imi la r to that provided for the sodium or by lowering the helium p r e s s u r e below the reac tor operating p r e s s u r e .

Lowering the p r e s s u r e of the secondary coolant helium while m a i n ­taining the conceptual facility configuration and performance capability is not considered prac t ica l because the helium compresso r horsepower requi rements would inc rease due to the inc rease in sys tem p r e s s u r e drop. The over-all system p r e s s u r e drop with the lower helium density could be reduced by reducing the sys tem the rma l per formance r e q u i r e ­ment . However, this is a lso considered an unacceptable solution.

The other possible solution to this problem is to double contain the helium inlet and outlet l ines , the p r i m a r y coolant heat exchanger and the he l ium- to - reac to r water heat exchanger. Since double containment must be achieved within the r e s t r a in t s of the existing space envelope, the size of the p r imary coolant heat exchanger and associa ted helium flow passages will be reduced with a net increase in loop p r e s s u r e drop. This p r e s s u r e drop would be reflected in increased horsepower and space requ i rements for the secondary system c o m p r e s s o r s .

The addition of double containment to the he l i um- to - r eac to r water heat exchanger would substantial ly dec rea se the amount of heat t r a n s ­f e r r e d to the reac tor p r i m a r y coolant, thus necessi ta t ing an inc rease in the heat removal capability of the secondary coolant sys tem.

Instrumentation to monitor for p r e s s u r e inc reases and /o r the p resence of water in the containment annuli would be required . Provis ion for such instrumentat ion is not considered a major problem. However, it would require additional connections at the top of the loop (see Section 5.2.4.7b).

The conceptual loop design allows the stagnant helium annuli in the in-core section to operate at the same p r e s s u r e as the p r i m a r y coolant heat exchanger, approximately 500 psi , in o rde r to simplify the design and fabrication of the loop. If the two sys tems must be separa te a complete redesign of the in -core section is required .

In summary, while it appears that double containment of all high p r e s ­sure helium may be achieved, a grea t deal of additional study and analysis will be required to determine its over-all effects on the loop design and ^ design objectives. In addition, this requi rement great ly inc reases the c o i ^ plexity of design and fabrication of an a l ready complicated and possibly marginal sys tem.

- 20 -

5.3 Out-of-Reactor Systems

The ou t -o f - reac to r section of the Thermal Flux Loop Faci l i ty consis ts of the following:

a) Secondary Coolant Loop - including f i l t e r s , heat exchanger(s), c o m p r e s s o r s , snubbers , valves and piping.

b) Helium (Commercial Grade) Supply System

c) Helium Drying System

d) Vacuum System

e) Instrumentat ion (See Section 5.4)

f) Util i t ies (See Section 5.5)

g) S t ruc tures (See Section 5.6)

Where possible c r i t e r i a for the Tit les I and I I design of these sys tems a r e presented here in together with just if icat ions for the c r i t e r i a where appropr ia te .

5.3.1 Shielding

The secondary coolant, commerc ia l grade helium, is circulated through both the in- and ou t -o f - reac to r sections of the Faci l i ty . It contains impur i ­t ies , mainly argon, that become activated during t rans i t through the in - reac to r portion of the secondary coolant sys tem. It may blow s ta in less s teel p a r t i ­c les , a s sumed to remain in the i n - r eac to r section after a ssembly and ins ta l ­lation, out of the i n - r eac to r section. These pa r t i c l e s will, of course , be radioact ive. Therefore , the secondary coolant will contain sources of activity for which shielding must be provided. A shielded fi l ter is provided in cubi­cle IB immediate ly downstream of the i n - r eac to r section of the Faci l i ty to filter out entrained p a r t i c l e s .

The shielding to be provided for secondary coolant system p rocess equipment and piping shall be adequate to linait the contact dose r a t e . Shifeld-ing will not be required for the helium supply sys tem or for the ins t rumenta ­tion and control sys tem.

The radiation source bases for the design of the shielding for the fil ter located immedia te ly downstreana of the i n - r e a c t o r section in Cubicle IB a re presented in Tables V I and V I I . The conceptual configuration of the fil ter not including shielding is shown in Figure 11. The assumed one gram of par t i c les constituting the source of activity is thought to be reasonable . P r e l i m i n a r y calculations indicate that the shielding requ i rements for the fi l ter will be relat ively modest , on the o rder of 2 to 3 inches of lead.

The bases for the design of shielding for other p r o c e s s equipment shall be the data presen ted in Tables V I I I , IX and X. Implicit in these data is an assumed sys tem a r rangemen t . The actual a r r angement , to be determined in Title I, will affect the relat ionship between t rans i t time in - r eac to r and ou t -of - reac tor , and therefore the activity induced in the secondary coolant.

- 21 -

Pre l imina ry calculations based on these data indicate that no shielding w L ^ be required for helium piping and equipment under normal operating cond^P tions.

Shielding requi rements under normal and emergency conditions shall be the subject of Title I study and shall follow from a detailed Faci l i ty hazards ana lys i s .

Table V I

BASIS FOR ESTIMATION OF ACTIVATION OF STAINLESS STEEL CHIPS

Irradiat ion Time, hr

Quantity of Chips, gm

10,000

1 (assumed)

Location of Chips During Irradiat ion

1 foot above top of ATR core (at bottom of heat exchanger)

Flux Trap Power , Mw 50

- 22

T a b l e V I I

STAINLESS S T E E L ACTIVATION IN H E A T EXCHANGERS O F ATR LIQUID M E T A L P A C K A G E L O O P S

10,000 HOURS IRRADIATION

Ac t iv i ty

Ac t iva t i on S o u r c e

I ron G a m m a 1 G a m m a 2 G a m m a 3

M a n g a n e s e G a m m a 1 G a m m a 2 G a m m a 3

C h r o m i u m G a m m a 1

C u r i e s / g m of S t a i n l e s s S tee l

1,0655 X 10"^

1.4587 X 10 - 1

5.7288 X 10 -2

G a m m a C u r i e s / g m of S t a i n l e s s S tee l

6.0731 X 10" 4 .5814 X 10" 2 .9833 X 10'

1.4587 X 10 4.3762 X 10' 2 ,9175 X 10'

5.6143 X 10 -3

G a m m a E n e r g y (Mev)

1.100 1.290 0.191

0.85 1.81 2 .13

0.323

Cobal t G a m m a 1 G a m m a 2

Nicke l G a m m a 1 G a m m a 2 G a m m a 3

2.8495 X 10

1.4289 X 10

2.8495 X 10 2.8495 X 10"

2.5863 X 10' 3.5580 X 10' 5.8584 X 10'

- 3 1 1

1728 33

1.12 1.49 0.37

Molybdenum Mo93 G a m m a 1 G a m m a 2 G a m m a 3 Mo99 G a m m a 1 G a m m a 2 Ganama 3 G a m m a 4 Ganama 5 M o l O l G a m m a 1 G a m m a 2 G a m m a 3 G a m m a 4

7.4690 X 10'

8.3887 X 10'

1.5100 X 10

7.4690 X 10 7.4690 X 10 7.4690 X 10"

7.3821 X 10' 8.3887 X 10" 9.2277 X 10' 8.3887 X 10' 9.2277 X 10'

1.4496 X 10 1.4119 X 10 5.2851 X 10 1.2835 X 10

- 6 - 6 - 6

- 4 -6 -5

0.26 0.68 1.48

0.140 0.181 0.372 0.741 0.780

0.515 0.083 0.896 2.08

Sulphur G a m m a 1

6.7079 X 10" 6.0371 X 10' 3.12

Si l icon G a m m a 1

To ta l

3.6808 X 10'

2 . 0 9 5 3 6 x 10 rr 2.5765 X 10'

2 . 3 4 8 1 6 x 10

23 -

-1 1.26

Table V I I I

COMPOSITION OF COMMERCIAL GRADE HELIUM

(Reference 7)

C onaponent

Carbon Dioxide

Argon

Hydrogen

Nitrogen

Methane

Helium

Weight Pe rcen t

0.000581

0,000049

0.00003

0.001763

0.000001

Balance

Table IX BASIS FOR ESTIMATION OF

SECONDARY COOLANT ACTIVATION

Secondary Coolant I rradiat ion Tinae, hr

Secondary Coolant Flow Rate, Ib /hr

10,000

2,290*

Secondary Coolant Residence Time, In-Pi le , 0.22 sec

Secondary Coolant Transi t Time, Compressor 5 Discharge to Suction, sec

Equilibriuna Activity Exists at Heat Exchanger Outlet

Flux Trap Power , Mw 50

* Gives naaximum in-core res idence time and t h e r e ­fore, maxinaizes impuri ty activation. See Appendix F .

- 24 -

HELIUM IMPURITY

Activation Source

Argon Oxygen

Ganama 1 Ganama 2 Ganama 3

Nitrogen Gamma 1 Gamma 2 Ganama 3

Total

Notes:

Table X

ACTIVATION IN ATR THERMAL 10,000 HOURS IRRADIATION

C u r i e s / c c gas

-11 4.0543 X 8.255 X

---

9.5698 X ---

4.0545 X

°- 6 10

10-1^

10-11

Activity

Ganama

4.0259

8.255 5.778 3,302

9.570 1,302 1.081

4,0262

C u r i e s / c c -1 1

X

-X

X

X

-X

X

X

X

10

-16 12-6 1°-17 10 ^'

- 1 6 1°- 6 1°- 7 10 ^'

10-11

FLUX LOOP

r~ln w-» w ^ o T7*n£aT"rrxr v ^ c

(Mev)

1.37 -

0.200 1.37 0.112

-6.13 7.10 2.70

1. Hydrogen and carbon produce no activation gananaas and, therefore , a r e not included above.

2. Helium density corresponding to above sources is 0.0036859 g m / c c .

5.3.2 Mechanical

5.3.2.1 General

The establishing of c r i t e r i a for the design of the secondary coolant systena r equ i re s detailed definition and descr ipt ion not only of a r e f e r ­ence exper iment , but a lso of the nature of other exper iments that may be instal led in the Faci l i ty . Design for a range of exper iments is outside the scope of these c r i t e r i a s tudies .

The maximum capacity of the systena is based on the naaxinaum heat t ransfer capability of the p r i m a r y coolant heat exchanger, this cor respond­ing to a maximum p r i m a r y coolant t empera tu re of 1,400 F , naaxinaum p r i ­m a r y coolant flow of 180 gpm, and i n - r e a c t o r section helium p r e s s u r e drop of 200 ps i . The reference capacity is defined by the per formance of the reference specinaen in a 50 Mw ATR outer flux t rap , this cor respond­ing to a naaxinaum p r i m a r y coolant t empera tu re of 1,400 F , p r i m a r y coolant flow of 140 gpna, and an i n - r e a c t o r section helium p r e s s u r e drop of 100 ps i . These data a r e sunamarized in Table XI .

It should be noted that the design of the secondary coolant sys tem for the thernaal flux loop shall be based on the reference capacity (Reference 8). The data presen ted in Table X I for the naaxinaum per formance of the p r i m a r y coolant heat exchanger is for infornaation only.

- 25 -

Table X I

DESIGN BASIS -THERMAL FLUX LOOP SECONDARY COOLANT SYSTEM

Basis

Secondary Coolant Secondary Coolant Mass Flow Rate,

Ib /hr Secondary Coolant Inlet Tenaperature

to Reactor , F Secondary Coolant Outlet T e m p e r a ­

ture F r o m Reactor , F Secondary Coolant Inlet P r e s s u r e

to Reactor , psig Secondary Coolant Outlet P r e s s u r e

F r o m Reactor , psig Secondary Coolant In-Reactor Section

P r e s s u r e Drop, psi Secondary Coolant Inlet P r e s s u r e to

Compresso r , psig Secondary Coolant Discharge P r e s ­

sure Frona Conapressor, psig Secondary Coolant Discharge Tem­

pera tu re Frona Conapressor Aftercooler, F

Main Conapressor, bhp Bypass Dryer Flow Rate, Ib /h r Secondary Coolant Dew Point, F HDW Inlet Tempera tu re , F HDW Outlet Tempera tu re , F Secondary Heat Exchanger Rating,

Btu /hr

A suggested e lementary flow diagram of the ou t -of - reac tor port ion of the helium secondary coolant sys tem is presented in Figure 7. Helium leaving the in-pile portion of the system flows through a shielded fi l ter to the secondary heat exchanger in which heat is t r ans fe r red to the high-p r e s s u r e demineral ized water sys tem. The helium then flows to the com­p r e s s o r s , through a fi l ter , and back into the in-pile section. Appropria te in te rcoo le r s , a f te rcoolers , and snubbers to danap out p r e s s u r e osci l lat ions a re provided with the conapressors . In-pile t empera ture control is a ccom­plished by autonaatically loading and unloading the c o m p r e s s o r s , thereby regulating helium flow to the in-pile section. A heliuna supply sys tem, a drying systena, a vacuum sys tem, the HDW sys tem, and appropr ia te i n s t r i ^ mentation support the main secondary coolant loop. Cr i t e r ia for the design and selection of components and auxi l iary sys tems a r e presented below.

Reference Specimen

Helium

12,000

140

350

600

500

100

450 (assumed)

650 (assume

140 720 120 -80 to -100 130 196

3,130,000

d)

Maximum Per fo rmance

of P r i m a r y Coolant Heat

Exchanger

Helium

16,950

140

363

600

400

200

350 (assumed)

650 (assumed)

140 1,250 170 -80 to -100 130 196

4,680,000

26

5,3.2,2 Secondary Coolant System Components

a) Piping, Valves , Mater ia ls of Construction

The out-of-pile portion of the secondary coolant system shall be fabricated of type 304 s ta inless s teel , in accord ­ance with ASA Code for P r e s s u r e Piping and the ASME Boiler and P r e s s u r e Vessel Code and appropr ia te nuclear c a s e s . The design t empera tu re shall be 400 F and the d e ­sign p r e s s u r e shall be 800 ps ig .

Jo in ts shall be of all-welded construction except at the f i l ters and c o m p r e s s o r s where flanged joints may be used to facil i tate installat ion and maintenance.

All valves shall be of be l lows-sea led or diaphragm type con­s t ruct ion.

P ipe s ize shall be consistent with ove r - a l l system p r e s ­sure drop linaitations,

b) Compres so r s

A naain conapressor capable of circulat ing helium at 12,000 pounds per hour and an auxi l iary compresso r capable of circulating helium at approximately 1,200 pounds per hour shall be connected in pa ra l l e l and shall operate on the line continuously. No spare compres so r is to be provided. The auxi l iary unit is provided for shutdown or emergency cooling and shall be sized in accordance with t ransient response requ i rements during Title I s tudies .

The conapressors shall be nonlubricated (teflon composition pistons) reciprocat ing units capable of circulat ing the secon­dary coolant without contanainating it in any way. The cona­p r e s s o r rat ings have been based on the i n - r eac to r section secondary coolant p r e s s u r e drop shown in Table XI and an assunaed ou t -of - reac tor section p r e s s u r e drop of 100 ps i . The c o m p r e s s o r s shall be equipped with af tercoolers and in te rcoolers of type 304 s ta in less steel al l -welded tube and shell construct ion.

The conapressors shall conform to the descr ipt ion presented in Table X I I , Space shall be provided for the conapressors in accordance with the data presented in Table X I I I . It will be noted that the physical size of the units is of such a magni­tude that sufficient room does not exist to pe rmi t their ins ta l ­lation in Cubicle IB, Accordingly, provisions have been naade for their installation in a compresso r vault outside the south wall of the ATR building together with the fast flux loop com­p r e s s o r s descr ibed in Reference 4, F i r m c r i t e r i a for the selection of the c o m p r e s s o r s , based upon detailed descr ipt ions of specinaens and exper iments to be installed in the Faci l i ty , shall be established during the Title I phase of the design.

- 27 -

The main compres so r shall operate off the ATR com­m e r c i a l bus and the auxi l iary unit off the ATR diesel bus . F r o m safety considerat ions, the use of these units in this manner is considered adequate since two s imul ­taneous fa i lu res , e i ther e lec t r ica l o r mechanical , must precede loss of coolant flow. This is considered naost unlikely.

Consideration has been given to the design of a secondary coolant systena common to both the Fas t and Thernaal Flux Fac i l i t i e s . This approach has been ruled out from safety considerat ions , and because system power r e q u i r e ­ments would be 75% to 100% higher for a conanaon systena than for two separa te sy s t ems . Operational flexibility would be lost if a common systena were to serve both Fac i l i t i e s ,

Table X I I SECONDARY COOLANT SYSTEM

COMPRESSOR CHARACTERISTICS

Mass Flow Rate, Ib /hr Suction and Discharge

Tempera tu re , F Suction P r e s s u r e , psig Discharge P r e s s u r e , psig No, of Stages Bhp Motor Rating, hp Power Source

HDW Requirements at 100 F, gpmi approx,*

Main

12,000 140

450 650

1 720 800

ATR Conanaercial

145

Auxiliary

1,200 140

450 650

1 72 80

ATR Diesel

15

* For a f te rcoolers , in te rcoo le r s , jackets and bear ings .

- 28 -

Table X I I I SECONDARY COOLANT SYSTEM

COMPRESSOR SPACE REQUIREMENTS

Main Auxil iary

Motor Rating, hp 800 80 Weight of Compresso r , Coolers

and Motor, lb, approx. 45,000 3,000 Concrete Pad Area , approx. 5' x 12' 2' x 3' Actual Floor Area 9 '5" x 22 '2" 5 ' 3 - l / 4 " x 8 '3" Floor Area Including Working

and Aisle Space 13'5" x 32'2" 9*3-1/4" x 14'3"

a - Snubbers

Conceptually, snubbers a r e to be instal led in the piping immediately ups t ream and downstreana of each compres so r . In addition to reducing vibrat ion, these units a r e a lso to be capable of reducing p r e s s u r e pu lsa ­tions to l ess than 1% of the line p r e s s u r e . Title I studies a r e to be ca r r i ed out to de termine whether snubbers a r e more suitable than r ece ive r s for this purpose .

b - Heat Exchanger(s)

The heat exchanger(s) shall be of all-welded tube and shell cons t ruc ­tion.

The helium shall be the tube side fluid. High-pressu re deminera l ized water shall be the shell side fluid.

Conceptually, one unit has been shown in Figure 7. The number of units to be installed in para l le l shall be determined from off-design p e r ­formance and safety studies ca r r i ed out during Title I to insure con­trollabil i ty over the des i red range of operating conditions.

The heat exchanger(s) shall be sized in accordance with the data p r e ­sented in Table X I V , F i r m c r i t e r i a for the selection of the exchanger(s) mus t await definition of Thermal Flux Loop Faci l i ty exper iments in Title I. Exchanger rat ings will be affected also by Title I studies that a r e to be ca r r i ed out to optimize loop heat losses to the ATR water .

- 29 -

Table X I V

SECONDARY COOLANT SYSTEM HEAT EXCHANGER CHARACTERISTICS

Helium Mass Flow Rate , I b /h r 12,000 Helium Inlet Tempera tu re , F 350 Heliuna Outlet Tenaperature, F 140 Heliuna Inlet P r e s s u r e , psig, approx. 475 Heliuna P r e s s u r e Drop, ps i , naax. 15 HDW Inlet Tempera tu re , F 130 HDW Outlet Tempera tu re , F 196 HDW Mass Flow Rate, Ib /hr 47,500

( -100 gpm) HDW P r e s s u r e Drop, ps i , max. 5 Heat Exchanger Rating, Btu /hr 3,130,000

c - F i l t e r s

F i l t e r s a r e to be provided immediate ly downstreana and ups t ream of the in-pile section of the sys tem. The f i l ters a r e provided to pro tec t the c o m p r e s s o r s , to t r ap par t icula te ma t te r which may be blown out of the in-pile section, and to prevent r e tu rn of par t icula te ma t t e r to the in-pile section. The need for and design of a fil ter capable of sepa­rating entrained sodium entering the sys tem as the resu l t of a leak of sodium into the secondary coolant shall be establ ished in Title I design and safety s tudies .

The f i l ters shall be of the s in tered s ta inless s teel type, capable of fi l tration in the 5-10 micron par t ic le size range, and shall be designed to pe rmi t cleaning and/or replacement of the fil ter e lements during ATR shutdown per iods . Spare capacity shall not be provided.

The design p r e s s u r e drop of the f i l ters shall be approximately 5 ps i .

The fi l ter immediately downstream of the in-pile section shall be shielded in accordance with c r i t e r i a p resen ted in Section 5,3,1 and shall be monitored as descr ibed in Section 5,4,2. The need and c r i t e r i a for a fil ter cask shall be established in Title I.

5.3.2.3 Secondary Coolant System Auxil iar ies

a - Helium Drying System

A bypass drying sys tem shall be provided to renaove mois tu re frona the secondary coolant sys tem pr io r to r eac to r s t a r t -up , and to naaintain the dew point of the secondary coolant at -80 F to -100 F during the ensuing reac tor cycle.

The drying agent shall be Linde Type 13X, 1/16-inch pel le ts , molec­ular s ieves or equal.

- 30 -

The capacity of the sys tem shall be a minimum of 100 g rams of naoisture while naaintaining the dew point of the secondary coolant at des i red leve ls .

The sys tem shall be sized and equipped to pe rmi t regenera t ion . The use of a dual absorpt ion d rye r and the optimum regenera t ion schenae shall be de termined during Title I.

The sys tem shall be designed for operation at 140 F and 650 psig and rated at not l e ss than 1% of the system flow r a t e .

b - Helium (Commercial) Supply System

A manifold of helium bottles shall be used to fill and provide m a k e ­up to the loop. The naanifold shall be sized during Title I.

P r e s s u r e regula tors shall be provided on each gas bottle, and a p r e s s u r e control valve downstreana of the manifold shall a lso be p r o ­vided. Make-up to the loop shall be under adminis t ra t ive control .

A p r e s s u r e controlled vent to the ATR stack shall be provided at the inlet to the in-pi le section to prevent loop over p r e s s u r e . Design of the ATR exhaust systena shall be modified as requi red by this Faci l i ty .

c - Vacuum System

A vacuum punap capable of evacuating the ent i re secondary coolant sys tem to a p r e s s u r e not g rea t e r than 10"^ mm Hg in 30 minutes shall be provided to p e r m i t purging and filling a t s t a r t - up . The punap shall be supplied with a high grade of high vacuum type oil (vapor p r e s s u r e l ess than 0.2 mic ron at 20 C) and shall be trapped to prevent back diffusion of oil vapors into the loop piping. The sys tem shall vent to the ATR stack,

5,4 Instruinenta.tion and Cc.v,-ols

5,4,1 General

The inst rumentat ion and control sys tem shall provide the means for continuous automatic and remote naanual control of Thermal Flux Loop p r o c e s s va r i ab l e s , shall indicate and r eco rd these va r i ab le s , and shall include appropr ia te a l a r m s and inter locks with the ATR safety sys tem requi red for both loop and reac tor safety. The sys tem is i l lus t ra ted conceptually in Figure 7. It shall be designed in a manner consistent with ATR p rac t i ce . Instrumentat ion shall be located on suitable display panels situated outside of Cubicle IB .

The bas ic sys tem control functions shall , for c r i t e r i a purposes , be the following:

a) The naaintenance of p r i m a r y coolant t empera tu re r i se a c r o s s the test specinaen at des i red levels by controlling p r i m a r y coolant pump speed and, t h e r e ­fore , p r i m a r y coolant flow.

- 31 -

b) The maintenance of p r imary coolant t empera tu re at specified levels by regulating secondary coolant flow,

c) The maintenance of secondary coolant p r e s s u r e at ^ ^ specified levels at the inlet to the in - r eac to r section.

It is recognized that the requirenaents of regulation of secondary cool­ant flow and maintenance of secondary coolant inlet p r e s s u r e may in terac t . Therefore , Title I effort will be required to study carefully the control problem and to devise a workable sys tem, A control schenae which may be capable of providing the des i red control is d i scussed in Section 5.4.3.

5.4.2 In-Reactor Instrumentation

The instrumentat ion provided for the in-pile package loop shall con­sist of thermocouples , sodium liquid level indicators and sodium leak detection devices located in the heliuna containing regions of the loop. If double containnaent of the secondary coolant is required, additional ins t ru -naentation will be provided in the containment annuli to indicate e i ther gas p r e s s u r e or in-leakage of reac to r cooling water .

The design philosophy that thermocouples and other sensing e lements will not penet ra te the walls of the liquid meta l sys tem, i .e . , the t h e r m o ­couple sheath shall not see liquid meta l , shall be the bas is for the design of the instrunaentation systena. This c r i te r ion is requi red for maintaining the integri ty of the liquid naetal sys tem. It will not preclude the m e a s u r e -naent of axial and radial t empera tu re profiles within the test specimen region. It will requi re that an exper imenter provide a test specinaen-end cap assembly for inser t ion into the in-pile package loop modified in such a way that the thermocouple wells welded into the end cap extend into those pa r t s of the test specimen region where t empera tu re m e a s u r e m e n t s a r e des i red . Sheathed thermocouples may then be inser ted into these wells and brought out of the in -core region in a manner analogous to that of other thermocouples . Thermocouples located in the buffer heliuna annulus s u r ­rounding the in -core region permi t the naeasurenaent of axial t empera tu re profiles along the outer edge of the test specimen region.

In addition to the thermocouples provided to m e a s u r e p roces s t e m p e r ­a tu re s , thermocouples shall also be provided to monitor motor s ta tor and bearing t empe ra tu r e s . The feasibil i ty of measur ing pr inaary coolant p r e s ­sure drop a c r o s s the test specimen shall be investigated during the Title I phase of design. The nunaber of thermocouples and leak detection devices to be provided for monitoring, control and safety purposes shall be d e -ternained during Title I. Level indicators will be provided in the loop sunap to pe rmi t system filling and external purification of the liquid meta l . The number and design of these indicators will be determined during Title I.

A naotor control center will be required to pernait both naanual and automatic control of pump speed, and therefore of p r i m a r y coolant t e m p e r ­a ture r i s e . Thermocouples naeasuring test section inlet and outlet t e m p e r ­a tures shall provide the control signal.

5.4.3 Out-of-Reactor Instrumentation ^ ^

In addition to providing the basic sys tem capabili t ies set forth in Section 5.4.1 the out -of - reac tor instrumentat ion and control sys tem shall a lso be capable of:

- 32 -

•a) Measuring g ross a , (3 and y activity in the secondary cool­

ant piping, the shielded fi l ter , Cubicle IB, and the compresso r vault .

b) Measuring and controlling secondary coolant mois tu re content at specified leve ls .

c) Alert ing the r eac to r opera tor to the exis tence of abnormal operating conditions such as ove r t empera tu re , over - or unde rp re s su re , loss of helium flow, presence of sodium or of other activity in the secondary coolant.

d) Shutting down the r eac to r in case of dangerous loop opera t ­ing conditions.

The design of the sys tem to be accomplished in Title I shall provide double tracking of instrumentat ion, a l a r m s and inter locks in accordance with ATR prac t i ce .

An important aspect of the design of the control system is the r e q u i r e ­ment that it be capable of operation over a wide range of loads. Of in te res t he re a r e the cha rac te r i s t i c s of the c o m p r e s s o r s and of the sys tem, and the sys tem control r equ i r emen t s . The head developed by the c o m p r e s s o r s is independent of load; system head losses inc rease approximately as the square of the flow. As a consequence, a var iable r e s i s t ance , in this case a compres so r suction thrott le valve, must be installed in the sys tem to balance sys tem losses against developed head for any des i red flow, t h e r e ­by controlling the p r e s s u r e at the inlet to the i n - r eac to r section. The main­tenance of p r i m a r y coolant tennperature level requ i res control of secondary coolant flow. This may be accomplished by a compresso r load control ler employing pneumatically operated suction valve unloaders , automatical ly controlled var iable c learance pockets, and automatical ly controlled bypass ­ing to prevent hunting at control ler nodal points corresponding to 25%, 50% and 75% load. This system will provide the des i red flow control and will be economical of power. The auxi l iary compresso r will always be in ope ra ­tion at its rated capacity. The system is conceptual and, as noted, in Section 5.4.1, its functions of flow and p r e s s u r e control a r e not independent. Title I efforts a r e therefore indicated to evaluate the suitability of this sys tem, and to design a control sys tem based on this or sonne other concept that will satisfy the control r equ i r emen t s .

5.4.4 Data Logging System

The main reac tor data logging center (Reference 3) provides normal scanning of 300 analog inputs, an independent logic check of the operat ions of the control system relay network, a memory of the past 5 minutes of operating history, acce le ra ted scanning of p rese lec ted quantit ies during t rans ien t s , and the calculation and in terpreta t ion of information. Twenty of the 300 analog points a r e allotted for the Thermal Flux Faci l i ty .

fThe design of the instrumentat ion and control system to be accomplished

ping Titles I and II shall provide for the t ransmi t ta l of appropr ia te data to e main reac to r data logging center , and for receiving and recording from

the data logging center information such as reac to r flux t rap power and control rod position, which may be of in teres t to the exper imente r .

- 33 -

5.5 Utilit ies

me The uti l i t ies required for the Thermal Flux Loop Faci l i ty shall inclui the following:

5.5.1 Elec t r ica l System

5.5.1.1 In-Reactor System

a) 480 V, 3 ph, 60 cy, failure free bus for 40 hp (approximate) sodium pump v a r i ­able frequency d r ives .

b) Sodium hea t e r s , 25 kw (approximate) .

5.5.1.2 Out-of-Reactor System

a) Vacuum System 3.7 kw, 480 v, 3 ph, 60 cy (5 hp)

b) Helium Drying System 1.5 kw, 110 v, 60 cy

c) Main Compressor ~563 kw ATR Com­m e r c i a l Bus*, 4,160 V, 3 ph, 60 cy (-720 Bhp)

d) Auxiliary Compresso r ~54 kw ATR Diesel Bus*, 480 V, 3 ph, 60 cy (-72 Bhp)

5.5.2 H igh -P re s su re Demineral ized Water System

a) Main Compressor ~145 gpm at 100 F *

b) Auxiliary Compressor ~15 gpm at 100 F*

c) Secondary Coolant ~100 gpm at 130 F * Heat Exchanger

*See Sections 5.3.2.2 b and d.

5.5.3 Ventilation

Cubicle IB and the compresso r vault shall be designed to be a pa r t of the ATR building gastight enc losure . The ATR ventilation systenn shall be sized to pe rmi t drawing 1,000 cfm of a i r from the basement a r e a adjacent to the cubicle and vault into each of these a r e a s , to maintain these a r e a s under a negative p r e s s u r e , and to d ischarge the a i r from these a r e a s to headers connected to the main exhaust sys tem. Butterfly type shutoff dampers a r e to be provided in the duct leaving each a rea to pe rmi t adjust­ment of the exhaust sys tem and isolation of the a r ea if nece s sa ry .

As specified in IDO-24041 (Ref. 4) the c o m p r e s s o r vault is to be p r l ^ vided with a 100 kw reci rcula t ion type a i r -handl ing unit incorporat ing a water cooling coil, and Cubicle IB with a 30 kw unit. These shall se rve

- 34 -

as sinks for heat generated by equipment within these a r e a s and shall be provided with cooling water frona the well water sys tem.

5.5.4 Instrument Air

100 psig at -30 F , Dew Point

5.6 Structures

Thermal Flux Loop Faci l i ty equipment shall be housed in three a r e a s within the ATR Faci l i ty:

a) The reac tor p roper in which the in-pile package loop and extension tube a r e located.

b) Cubicle IB on the f i r s t basenaent of the ATR building within which the f i l t e r s , heat exchanger(s) , helium drying sys tem and vacuum system a r e to be located, and outside of which the helium supply sys tem, pump motor control center , and instrumient control panels a r e to be located.

c) Compressor vault outside the south wall of the ATR building within which the c o m p r e s s o r s , snubbers and auxi l ia r ies a re to be located.

These a r e a s a r e related to one another as i l lus t ra ted in F igures 12 and 13. Piping and instrumentat ion leads will pass from the reac tor to Cubicle IB and the a r e a surrounding it through the mechanical pipe duct, nozzle access t rench, and outer shina rod service a r e a . A pipe t rench terminat ing in Cubicle IB is provided for passage of piping and ins t ru ­mentation leads from the compresso r vault and the control center .

Of importance in the equipment a r rangement is the location of the compres so r s in a vault and not in Cubicle IB. This a r rangement was d i c ­tated by the physical size of these uni ts . Their location outside the south wall of the ATR building, below grade , and on a level with the f i r s t b a s e ­ment, is in close proximity to the p roces s equipment they se rve , does not in terfere with the location of other ATR or loop equipment, is sufficiently r e s t r i c t ed from other a r e a s so as to minimize the consequences of unfore­seen occur rences , and from s t ruc tura l considerations pe rmi t s installation and operat ion of this equipment independent of al l other ATR equipment. Access to the compresso r vault is provided through a door and a knockout plug in the south wall of the ATR building and through a hatchway in the vault roof.

The requi rements for f ire protect ion sys tems and equipment in the cubicle and compresso r vault shall be determined during Title I.

The cubicle and compresso r vault shall be equipped with telephones, evacuation alarnns, and a paging sys tem. The lighting level in the c o m p r e s ­sor vault shall be 20 foot-candles; that provided in Cubicle IB is a lso 20 foot-candles.

The requ i rements for ho is t s , monorai l s and other ma te r i a l s handling equipment shall be evaluated in Title I.

- 35 -

The over -a l l a r rangement is sat isfactory from space and safety con­s iderat ions , and facil i tates installation, maintenance, and removal of equi | ment.

6. OTHER REQUIREMENTS

In addition to providing a complete set of design drawings and specifica­tions for all i tems mentioned in this repor t , the AE shall a lso p r epa re the following as par t of the Title I and Title I I design work:

a) P re l imina ry Proposa l

b) Safety Analysis

c) Test P rocedures

d) Operating Manual

7. INFORMATION TO BE SUPPLIED BY IDO AT THE START OF TITLE I DESIGN

1) Complete and fully detailed specifications for the test assembly(s) , drawings and dimensions defining the test specimen(s) , its (their) supports and instrumentat ion.

2) Complete and fully detailed nuclear , thermal and hydraulic opera t ­ing requi rements for the test specimen(s) .

3) Additional requirennents other than those l isted in Section 6.

36 -

REFERENCES

1. Bless ing , W. G., et a l . , Conceptual Design of an In-P i le Package Loop for Sodium Cooled Thermal Reactor Fuel Test ing, Atomic Power Development Assoc ia t e s , Inc. , APDA-145, September 15, 1961.

2. Le t te r from H. M. Leppich to R. H. Gordon, Apr i l 4, 1962, File No. RTR:JCMc. "ATR Liquid Metal Loop Object ives ."

3 . Advanced Test Reactor - ATR, P r e l i m i n a r y Safety Analysis Report , Ebasco Services Incorporated and The Babcock & Wilcox Company, IDO-24040, Apri l 1962.

4. Design Cr i t e r i a for a Fas t Flux Liquid Metal Loop in the Advanced Test Reac tor , Ebasco Services Incorporated and The Babcock & Wilcox Company, IDO-24041, July 1962.

5. Summary Report of Design Cr i t e r i a for the Liquid Metal Package Loop's Hot Cell Faci l i ty and Handling System at the Advanced Test Reactor , Ebasco Services Incorporated and The Babcock & Wilcox Company, IDO-24043, March 1963.

6. Design Cr i t e r i a for a Fas t Flux Liquid Metal Loop in the Advanced Test Reac to r . The Babcock & Wilcox Company and Ebasco Services Incorporated, IDO-24041 Supplement I, March 1963.

7. Air Reduction Company, Catalog 320, Page 1, September 1955.

8. Le t te r from H. M. Leppich to R. H. Gordon, September 28, 1962, File No. RTRrJCMc, "ATR Thermal Flux Loop P r i m a r y Heat Exchanger AT(1 0-1 )-l 075."

9. Scope Document for P repa ra t ion of Design C r i t e r i a , Liquid Metal Loops, Advanced Test Reactor , January 2, 1962, U.S .Atomic Energy Commiss ion , P . O. Box 2108, Idaho Fa l l s , Idaho.

- 37 -

p—

I o

SHROUD R I N S - HLXA&ONJkL SHROUD-

N l FLOW B&FFLE

090 OIK. WIRE

SHROUD RIN&

SCkLE.;-2nSlZt

SPfcCER 'WIRE.S-FUE.L P I N S

FIGURE 1

REFERENCE TEST SPECIMEN

^V5 TRt/MSM TA TiaU coHueerofK

6ECTION V-V

:iECT/ON \l-\J

NfO

^

^vsaiitn n/ae

HEL/UM-TO-REACTOR WATER HEAT EXC/^ANO'^R.

5KC-e95S-F

6ECTION Z - Z FIGURE 2

THERMAL FLUX LOOP

600

500

Q O m H 400 W H

«r o Q <; M X

H CO

• J

< O H

300

200

100

1—

—

1—

L_

• 1

^

y /

/

1

A /

80 100 120 140 160 180

LOOP SODIUM FLOW, GPM

FIGURE 3

HYDRAULIC CHARACTERISTICS

OF PRIMARY COOLANT SYSTEM

DROP TUBE

NT PRESSURIZED WATER LOOP

(CUBICLE ID)

r

NRLOOP ( A - 1 ) (CUBICLE 2E)

FUEL ELEMENTS

N T GAS COOLED LOOP

(CUBICLE IC)

FAST FLUX LOOP (CUBICLE lA )

THERMAL FLUX LOOP

(CUBICLE IB)

NR LOOP ( A - 5 ) (CUBICLE 2D) —" NR LOOP (A -1 )

(CUBICLE 2B) -NR LOOP(A-S)

(CUBICLE 2C)

-NR LOOP (A-5) (CUBICLE 2D)

FIGURE 4

LOCATION OF THE THERMAL FLUX LOOP \ij^

THE ADVANCED TEST REACTOR CORE

I fire cooe ^uet. £i.0M£KtT

rear spectma4 e-Losuae 4 .suppoeT -

i= ^\v^\\\

Na. - * -

/ \ I '

V I < V k \ ' l . i | . U ' ' • ' k k I. t I k I. M ( I? ' • • ' H.'I. <• 'H. t I. I. k I.

ixxxxxxxxxxxxxxwr--/'LUX TBHPBAFfi-E

XXXXXXXX'^

FU£L. TBa T - peeaAuae TUBS

ufPBB ocTive ruei. LIMB SL s* a-

LO'^Ba ACTfVM ruSL UNM CL 7a o

9oeeotrms < 070 wipe

SECT/ON •Z-'Z FIGURE 5

THERMAL FLUX LOOP IN-CORE SECTIONS

H£ATEm SeCTtOf^

urt.H/M--r»-M0AeTOM wATmm mur MitcHAti^mK

- NTLH/M'tN

COHOUIT

SOOHMt - TO- MffC lUM PUMAffY fMAT rtCMMMfrtfJt

HELIuy-TO-KBACTOfi WATMR He AT EXCHANCMR

SECT/ON Z-Z FIGURE 6

THERMAL FLUX LOOP HEAT

EXCHANGERS

PRlHVUfr COOLANT SYSTtM-'

( t * ) ( b * ) rrmj LoorCLMJw m »

l.«A< DfTIcnK^

TTKANANT WLIUM 4NNUU

MO TES:

I MPLICATIOM or imlUMt l lTkno i t TO BK DfTIKHMCe IN TIT1.I I '

3i M.lklCM3 AND IIIT«U.OC<t Tt M M T K M I M t B M TITLl t

4 14UI4 COMPMUAA UOA» CMfTk»l. I« 4CMMVX. CONT*SU,tf dVHfcTB* * i » - n « u VAUa*

%• TUM»WCO>m.n M l M i a i N a L M T IMRItKI>KB b K M l * IN tituUn AMNUU

FIGURE 7

ELEMENTARY FLOW AND INSTRUMENTATION

DIAGRAM

FIGURE 8

THERMAL

FLUX LOOP MAXIMUM

PHYS

ICAL

EN

VELO

PE

TEST SPECIMEN

CONTAINMENT TUBE END CLOSURE

PARTING GROOVE'

PARTING TOOL GRIP

Parting groove made in hot cell during specimen removal operation.

FIGURE 9

CONTAINMENT TUBE END CLOSURE

at/ri.Mr coMf/mcr/ot/

N*. FILL LINE CONNECTION

Na. FLUSH LINE. CONNE.CTION

Ha FILL LINE. CONKECTIOH

z

He JlkCKE.T PRESSURIZINe CONNE.CTION

SECTIOK X-X (SCRfiW TA^PtMA POK

CXTSNSIOH T\Mt CONMKCTIOM)

d ' /Vd? oMoave

SECTION Z -Z

P^CK^GE LOOP

SECTION W - W

t :

EXTENSION TUBE

FIGURE 10

SUGGESTED TOP END ARRANGEMENT FOR THE

THERMAL FLUX LOOP

r\iT€ft. £L6M*HT<

*5eCTKj>N A A

PlLTeREUEMENT

I NL6T-

OUTLST

ELEVATION

FIGURE 11

SECONDARY COOLANT SYSTEM FILTER

FIGURE 12

FACILITY ARRANGEMENT PLAN

e E . C T l O K J A - A

1 * FLoor. fcl ^»' -o

SAMnlE AKCA

Tnai0M4>, 6'Pm

fiffoT I'^ftfcMT. ftL T ' - «

P«lje:TK*T>eM«i, 6 - ^ P v p I S ^ EVES

'Z"*&fcMT ftCfeo'-O

e>ecrriow B-e,

ENLARGED SECTION

FIGURE 13

FACILITY ARRANGEMENT

APPENDIX A

Design Objectives and Fixed P a r a m e t e r s

Design Data Sheet No. M-5001 EO-AEC 9954 Contract No. AT(10-1)-1075 July 18, 1962 Rev. 1 - December 5, 1962 E, E. Van Brunt, J r .

EBASCO SERVICES INCORPORATED ADVANCED TEST REACTOR - DESIGN DATA

LIQUID METAL LOOP FACILITIES THERMAL FLUX LOOP

DESIGN OBJECTIVES AND FIXED PARAMETERS

PURPOSE AND SCOPE

To state the design objectives and fixed p a r a m e t e r s with r ega rd to the the rma l flux loop.

DESCRIPTION

A review of APDA-145 by the exper imen te r s and others was conducted to determine the design objectives and fixed p a r a m e t e r s for the the rma l flux loop. The attached is a l is t of the design objectives and fixed p a r a m e ­t e r s that were determined. This design data sheet will be rev ised from t ime to t ime during the prepara t ion of the design c r i t e r i a to add other i tems of information.

A - 1

Design Data Sheet No. M-5001 Liquid Metal Loop Faci l i t ies Design Objectives and Fixed P a r a m e t e r s December 5, 1962

NOMINAL DESIGN CONDITIONS

In-Pi le loop design and dimensions

Basic m a t e r i a l s of construct ion

P r i m a r y coolant

P r i m a r y coolant t empera tu re (max)

Operating P r e s s u r e , psig

Secondary coolant

Cover gas

Min. Loop Design Life, h r s .

Test Specimen

Type of fuel

F i s s i l e m a t e r i a l

F i s s i l e m a t e r i a l density, g /cc fuel

Type of spec t rum

Radial power depress ion, initial,

m a x / m i n

Angular power variat ion

Maximum power ra t io (0,625 ev cutoff)

f a s t / t h e r m a l 20 Fuel burnup, 10 f i s s ions /cc

fuel

Specimen power generat ion (fission), kw

Average pin power (fission), kw/in.

Maximum pin power (fission), kw/in.

Number of pins

DESIGN OBJECTIVES AND FIXED PARAMETERS

TBD*

TBD*

Na

TBD*

60 - 150

TBD*

TBD*

10,000

Ceramic (UC)

Uranium

TBD* (0.4-1.2)

Thermal

1.5-2,5

Uniform as possible

0.30

1-3

150-1500

5

9 5-7

A - 2

Design Data Sheet No. M-5001 Liquid Metal Loop Fac i l i t i es Design Objectives and Fixed P a r a m e t e r s December 5, 1962

Specimen length, in.

P in outside d iameter , in.

Fue l outside d iameter , in.

Bond m a t e r i a l

Cladding m a t e r i a l

Clad th ickness , in.

Geometry

Spacer outside d iameter , in.

Minimum tes t section length, ft.

Minimum tes t section d iameter , in.

Design Tempera tu re s

P r i m a r y coolant t empera tu re

at specimen outlet (max)

P r i m a r y coolant t empera tu re

at specimen inlet (min)

P r i m a r y coolant maximum

axial A T

Minimum heat exchanger capacity

Maximum allowable tes t specimen

surface t e m p e r a t u r e at zero

power (post - i r radia t ion)

Hydraulic Cha rac t e r i s t i c s

Maximum flow, gpm

Specimen p r e s s u r e drop, ft.

(at max flow)

48 (max)

0.560

0.500

Sodium

Stainless s teel

0.020

Tr iangular pitch, sp i ra l space r s

0.090

TBD*

2.35

1400 F

800 F

350 F

TBD*

1000 F

TBD*

180

100

TBD = To be de termined

A . 3

APPENDIX B

Secondary Coolant System

Design Data Sheet M-5002 E.O.-AEC 9954 Contract No. AT(10-1)-1075 December 1, 1962 E. E. Van Brunt, J r .

EBASCO SERVICES INCORPORATED ADVANCED TEST REACTOR - DESIGN DATA

LIQUID METAL LOOP FACILITIES THERMAL FLUX LOOP

SECONDARY COOLANT SYSTEM

PURPOSE AND SCOPE

To presen t the conceptual design bas i s for the out-of-pile portion of the secondary coolant sys tem which supports the Thermal Flux Loop to be instal led in the Advanced Tes t Reactor .

The design is consistent with the per formance of the in-pi le sect ion of the loop d iscussed in B&W Design Data Sheet No. R-5001. It has been c a r r i e d forward only far enough to es tabl ish space and utility r e q u i r e ­ments for sys t em components . It has not, at this t ime, been analyzed for controllabil i ty nor has it been optimized. These do not appear to be major problems and, therefore , can be solved during Title I work.

DESIGN BASIS

The conceptual design of the secondary coolant sys tem is based upon Case I of Table I in B&W Design Data Sheet No. R-5001, which is the maximum the rma l capability under the following conditions:

Helium Mass Flow Rate 12,000 Ib /h r

Helium Inlet Tempera tu re to In-Pi le Section 140 F

Helium Outlet Tempera tu re from In-Reactor Section 350 F

Helium Inlet P r e s s u r e to In-Pi le Section 600 psig

Helium P r e s s u r e Drop - In-Reactor Section 100 psi

GENERAL DESCRIPTION OF LOOP

An e lementary flow d iagram of the out-of-pile portion of the helium secondary coolant sys tem which supports the The rma l Flux Loop is p r e ­sented in SKC-9954-1. Typical p roces s conditions a r e presen ted in Table I.

Helium leaving the in-pi le portion of the sys tem flows through a shielded fi l ter to the loop heat exchanger in which high p r e s s u r e de -minera l i zed water (HDW) is used to cool it to a t empera tu re of 140 F ,

B - 1

Design Data Sheet M-5002 Liquid Metal Loop Fac i l i t i es Secondary Coolant System December 1, 1962

It then flows to the compres so r , through a fi l ter , and back into the in-pile sect ion. Appropria te in te rcoolers and af tercoolers a re provided with the c o m p r e s s o r s . Loop t empera tu re control is accomplished by a c o m p r e s s o r load control ler employing pneumatical ly operated suction valve unloaders , automatical ly controlled var iable c learance pockets, and automatical ly controlled bypassing to prevent hunting at control ler nodal points corresponding to 25, 50 and 75% load. This sys tem will provide the des i r ed flow control and will be economical of power. Snub-be r s a r e provided ups t ream and downstream of the c o m p r e s s o r s to damp out p r e s s u r e and flow osci l la t ions . The secondary coolant loop is p r o ­vided with a hel ium supply sys tem, a bypass drying sys tem, and a vacuum sys tem used to purge the loop. The HDW sys tem supplies cooling water for the heat exchanger and the c o m p r e s s o r in te rcoole rs and a f te rcoole rs . The ins t rumenta t ion shown is bas ic and will be ult imately designed in a manner consis tent with es tabl ished s tandards for ATR.

Table I - The rma l Flux Loop Secondary Coolant System*

Secondary Coolant Helium

Helium Mass Flow Rate, Ib /h r 12,000

Helium Inlet Tempera tu re to Reactor , F 140

Helium Outlet Tempera tu re from Reactor , F 350

Helium Inlet P r e s s u r e to Reactor , psig 600

Helium P r e s s u r e Drop - In-Pi le Section, ps i 100

Helium Inlet P r e s s u r e to Compresso r , psig 450 (assumed)

Helium Discharge P r e s s u r e from Compres so r , psig 650 (assumed)

Helium Discharge Tempera tu re from Compres so r

Aftercooler , F 140

Main Compresso r , bhp 720

Bypass Dryer Flow Rate, Ib /h r 120

Helium Dew Point, F -80 to -100

HDW Water Inlet Tempera tu r e , F 130

Secondary Heat Exchanger Rating, B tu /h r 3,130,000

* Based on requ i rements for reference specimen.

The f i l t e r s , heat exchanger, bypass d rye r and vacuum sys tem will be located within the confines of Cubicle IB of the ATR F i r s t Basement Plan . The c o m p r e s s o r s , i n t e rcoo le r s , a f te rcoolers and snubbers mus t be located outside of this cubicle, since the shielded space provided is inadequate. Location of these components in an unshielded a r ea is not

B - 2

^ ^ s i g n Data Sheet M-5002 Ljiquid Metal Loop Fac i l i t i e s Secondary Coolant System December 1, 1962

considered hazardous because hel ium is non-act ivat ing, i s a secondary coolant and, on leaving the in-pi le section, is f i l tered through a shielded f i l ter . P r e l i m i n a r y calculat ions based on the information p resen ted in Babcock & Wilcox Company Design Data Sheet No. R-5004, "The rma l F lux Liquid Metal Package Loop Secondary Coolant Impur i t ies Act iva­t ions" indicate that the c o m p r e s s o r s and thei r assoc ia ted piping will not requ i re shielding. The vacuum sys tem and hel ium supply sys t em will be common with the F a s t F lux Loop and therefore will be located at points convenient for joint u se .

Loop components and supporting sub - sys t ems a r e descr ibed below.

PIPING

Loop piping will be fabricated of Type 304 s ta in less s teel and will be of all welded construct ion except at the f i l ters and c o m p r e s s o r s where flanged joints may be used to facil i tate instal la t ion and m a i n t e ­nance.

Pipe s ize will be selected after location of the c o m p r e s s o r s and component p r e s s u r e drops have been de termined. It will probably be in the range of 4 in. Schedule 40.

All valves will be of bellows sealed type construct ion.

COMPRESSORS AND SNUBBERS

Conceptually, two one-s tage , non- lubr icated (teflon composit ion pis tons) , horizontal ly opposed, rec iprocat ing c o m p r e s s o r s equipped with in te rcoo le r s , a f tercoolers and snubbers a r e considered for use in the loop. The c o m p r e s s o r s a r e connected in pa ra l l e l and a r e ra ted and operated as shown in Table I I . No spare capacity is being p r o ­vided.

Table I I - Helium Compres so r Charac te r i s t i c s

Mass Flow Rate

L b / H r

12,000

1,200

Suction and Di s ­

charge T e m p e r ­a tu re - F

140

140

Suction P r e s s u r e

Ps ig

450

450

Di P

scharge r e s s u r e Ps ig

650

650

Motor Rating

Hp

800 80

ATR Power Source

Commerc Diesel

i a l

HDW* Requ i re ­

ment at 100 F

Gpm

145

15

Unit

Main

. ^ ^ i l i a r y

*"Tor Aftercoolers , In te rcoo le rs , Jackets and Bear ings

B - 3

Design Data Sheet M-5002 Liquid Metal Loop Fac i l i t i e s Secondary Coolant System

Both c o m p r e s s o r s a r e operated on the line continuously. The auxi l iary unit is provided for shutdown or emergency cooling. F r o m safety con­s idera t ions the use of these units in this manner is considered adequate since two simultaneous fa i lures , e i ther e lec t r ica l or mechanical , must p recede loss of coolant flow and this is considered unlikely. Appropr ia te means a r e to be provided for unloading the units to facili tate s tar tup and p r e s s u r e control .

Appropriate snubbers a r e instal led in the piping ups t r eam and downst ream of the c o m p r e s s o r s to reduce p r e s s u r e pulsations to l ess than 1% of the line p r e s s u r e .

The floor a r e a requ i rements for the c o m p r e s s o r s a r e such that they cannot be instal led with other loop equipment in Cubicle IB, Alternat ive unshielded locations a r e outlined in Design Data Sheet M-4003.

HEAT EXCHANGER

The heat exchanger is to be of al l -welded tube and shell cons t ruc ­tion and is fabricated of Type 304 s ta inless s tee l . The secondary coolant will be the tube side fluid and HDW the shell side fluid. The unit will be ra ted in accordance with maximum limiting p roces s conditions de t e r ­mined from Design Data Sheet R-5001 during Title I s tudies . No spare capacity is being provided. Conceptually, one unit is deemed adequate. However, for control purposes , the nvimber requi red will be determined from off-design per formance studies during Tit le I which may indicate the need for para l le l units ,

FILTERS

F i l t e r s a r e to be provided immediate ly downst ream and ups t ream of the in-pi le section of the loop to remove par t icula te ma t t e r which may be picked up from the loop, thereby protect ing the c o m p r e s s o r s and to prevent r e tu rn of par t icula te ma t t e r from the out-of-pile section to the r eac to r . The fi l ter immedia te ly downstream of the in-pi le section will be shielded.

Two types of f i l ters a r e presen t ly under considerat ion for use: (1) a s in tered s ta in less s tee l unit capable of f i l t rat ion in the 5-10 micron range , and (2) a high t empe ra tu r e absolute f i l ter whose utility will depend upon its abili ty to withstand the p r e s s u r e drop that will be developed.

HELIUM DRYER

A bypass d rye r operated at 140 F and 650 psig and ra ted at 1% of M the sys tem flow ra te is provided to maintain the helium dewpoint at -80 to -100 F . Regenerat ion will be accomplished during r eac to r shutdown.

December 1, 1962

B - 4

Design Data Sheet M-5002 Liquid Metal Loop Fac i l i t i es Secondary Coolant System December 1, 1962

HELIUM SUPPLY SYSTEM

A manifold of helium bottles will be used to fill and provide makeup to the loop. The manifold p r e s s u r e will be maintained at a p r e s s u r e approximately 50 psi g rea t e r than that of the connpressor suction naani-fold by using p r e s s u r e regula tors on each gas bot t le . System makeup will be adminis t ra t ively control led.

This sys tem is also used to regenera te the molecular sieve d r y e r .

Loop o v e r - p r e s s u r e at the inlet to the in-pi le sect ion is maintained through a p r e s s u r e controlled vent.

VACUUM SYSTEM

A vacuum sys tem capable of evacuating the loop will be provided to pe rmi t purging and filling at s ta r tup . This sys tem will be s ized during Title I.

INSTRUMENTATION

The bas ic ins t rumentat ion requ i rements of the sys tem a r e p resen ted in SKC-9954-M-1. The sys tem will be designed in a manner consistent with ATR p rac t i ce .

UTILITY REQUIREMENTS SUMMARY

a. Power - 480 Volt, 60 Cycle, 3 Phase

1 - Main Compresso r

2 - Auxil iary Compresso r

3 - Vacuum Pump

4 - Dryer Heater

b . High P r e s s u r e Demineral ized "Water

1 - Main Compresso r 145 Gpm at 100 F

2 - Auxil iary 15 Gpm at 100 F

3 - Heat Exchanger Equivalent to 3,130,000 Btu /hr at 130 F inlet t e m p e r a t u r e .

720 Bhp

72 Bhp

5 Hp

1.5 Kw

B - 5

ii 0 J

t- t

J s s H

f

ivJ^- mil

'(V"-^>

-

"t^O^n^}"

-a^

qg

nn

gy

.

f—

"^

l

ll >

0

* i

H

(^

-a© X

I '>!

13 I

H^

-C

»-4)

IftN^V

^

^'

-8M

D

--n

^

g>-c (?>—

C)

«)-^

i)—

^i) (l>

^

ti

(1>

-^8

M3

>-

)-

^

(D

(r> r>

(^ ^

iJ^

^ r: }i

rJ fr

-:^^

t

r"

——

j-

T^

i—

'—

I—

\

z o

v-

-M-

5

I- F

«

II if Q

J

IN

1 i

ll m IN

-si,

ll

Q

I I

L_J| r-i-n

I I

W

(l>-®

®--«) -

*!

! I

.

I-4SU

M

i>-

(i>--V

-^v

I I

I

88

•"Si

If r H

i

u

APPENDIX C

The rma l Analysis of Helium Cooled Heat Exchanger

Design Data Sheet R-5001 59-3075-30 Contract No. AT(10-1)-1075 May 14, 1962 T. Speidel Revision 1, August 24, 1962 H. Honig

THE BAB COCK AND WILCOX COMPANY FOR EBASCO SERVICES INCORPORATED

PRELIMINARY DESIGN DATA ADVANCED TEST REACTOR

THERMAL FLUX LIQUID METAL PACKAGE LOOP THERMAL ANALYSIS OF HELIUM COOLED HEAT EXCHANGER

PURPOSE AND SCOPE

To repor t the resu l t s of a study made on a conceptual heat exchanger that will satisfy the Design Objectives.

GENERAL DESCRIPTION

To accommodate the space l imitat ions and to minimize the com­plexities that accompany the fabrication, operation and handling of the Thermal Flux Loop seve ra l different heat exchanger designs were examined.

The f i r s t design considered embraced the pr inciple of var iable liquid level control as proposed in APDA-145. This scheme was r e ­jected for reasons stated in Design Data Sheet R-5008.

The second design studied and repor ted in the original i s sue of this data sheet consis ted of two tube sheets connected with sp i ra l tubes to c a r r y the sodium up from the specimen to the pump. The pximp d i s ­charged the sodium downward through an annulus formed by the shell of the heat exchanger and the p r e s s u r e tube wall . Helium coolant was directed around the outside of the sp i ra l tubes to remove the heat from the sodium. This design was rejected for two r ea sons : (a) space l i m i ­tat ions, and (b) complexit ies involved in direct ing the helium coolant to and from the shell side of the heat exchanger.

The th i rd design which appears to satisfy all of the conditions both implied and stated in the design objectives i s descr ibed below.

This design was based on two major r equ i r emen t s . It had to con­form with the space l imitat ions of 2 feet in length and 5-1 /8 inches out­side d iameter . It a lso had to have the capability of t r ans fe r r ing 1573 kw of heat from the sodium to the helium with an inlet sodium t empera tu re to the heat exchanger of 1400 F . With these two requ i rements and the select ion of a reasonably smal l outside d iameter tube, the above condi­tions could be satisfied only after inser t ing the maximum heat t r ans fe r a r ea in the available space . A schematic view of the heat exchanger is shown in F igure 1 of this appendix. The full s ize drawing is shown in F igures 2 and 6 in the body of this repor t .

C-1

Design Data Sheet R-5001, Rev. 1

The heat exchanger consis ts of a shell and tube design with the sodium making two passes through the tubes, and the helium making one pass around the tubes . The bottom tube sheet is divided into two concentric tube sheets to compensate for the difference in axial tube expansion be ­tween the inner port ion and the outer portion of the heat exchanger.

The shell side of the heat exchanger contains a cylindrical baffle that prevents mixing of helitim between the inner and outer portions of the heat exchanger . The bottom of the baffle is orificed to provide the p roper hel ium flow through the inner and outer por t ions . The inner hel ium s t r e a m mixes with the outer helium s t r e a m before leaving the heat exchanger .

Provis ions to accommodate s t ruc tu ra l requ i rements , such as main­taining allowable s t r e s s e s in the var ious pa r t s of the heat exchanger, will be made during the Title I effort. Design problems ar i s ing from manufacture and assembly techniques will also be solved at this t ime .

The hot helium leaves the bottom of the shell side of the sod ium-to-helium heat exchanger by way of a collection header connected direct ly to and located on the eas t side of the i n - r eac to r port ion of the loop. One side of the header forms a tube sheet for the 28 one-half inch outside d iameter tubes . These tubes form the he l ium- to - r eac to r coolant water heat exchanger . The tubes t e rmina te in a tube sheet and header that is s imi l a r to the aforementioned assembly . This header also forms the two 90 degree bends that a r e neces sa ry to direct the helium into the d ischarge piping that is located in the extension tube.

The outside of the tubes a r e enclosed by a shell that d i rec ts the r e ­actor coolant water down around the tubes and down into the flux t r ap utilizing the core p r e s s u r e drop to maintain the flow. The shell is lo ­cated on the eas t side of the loop and extends downwards to a point that is approximately 4 inches below the sodium-to-hel ium heat exchanger. Here the shell extends completely around the loop and enters the flux t r a p .

LOOP DESIGN OBJECTIVES*

F i s s ion power range - 150 to 1500 kw

Sodium Tempera tu re , Max. - 1400 F

Sodium Tempera tu re , Min. - 800 F

Sodium Flow, Max. - 180 gpm

Axial Tempera tu re Difference, Max. - 350 F

* Based on Design Objectives from a le t te r from H. M. Leppich - IDO, to R, H. Gordon - Ebasco Services Incorporated. Subject: ATR Liquid Metal Loop Objectives - 4 -4-62 .

C-2

Design Data Sheet R-5001, Rev. 1

ASSUMPTIONS USED IN THE REFERENCE DESIGN

(1) The helium inlet t empera tu re to the i n - r e a c t o r portion of the loop was a r b i t r a r i l y se lected as 140 F .

(2) Calculations a s sume the distr ibution of the hel ium flow ra tes in both port ions of the shel l side is propor t ional to the f r i c -tional r e s i s t ance offered by the tubes only. It is anticipated that R&D investigations will be neces sa ry to deternaine the configuration needed to obtain the p roper flow ba lances .

(3) F o r the reference case (Case I), the combination of 12,000 Ib /h r hel ium flow and 100 ps i hel ium p r e s s u r e drop was a s sumed in o rder to produce a reasonable hel ium pumping power within the heat exchanger space l imi ta t ions . In determining the nnaxi-mum heat exchanger capacity (Case II) a 200 psi p r e s s u r e drop was a rb i t r a r i l y assumed. Applied to the heat exchanger con­figuration a r r ived at for the reference case , this 200 psi p r e s ­su re drop pe rmi t s a hel ium flow of 16,950 Ib /h r .

(4) Because of the heat exchanger configuration, a negligible amount of heat pa s se s a c r o s s the outer wall of the p r i m a r y heat exchanger from the helium to the r eac to r wate r .

PHYSICAL DATA FOR SODIUM-TO-HELIUM AND HELIUM-TO-WATER HEAT EXCHANGERS

Sodium to Helium

Niimber of Tubes, Outer P a s s - 47

N\imber of Tubes, Inner P a s s - 54

Tube Outside Diameter , inches - 0.25

Tube Wall Thickness , inches - 0.020

Tube Length, feet - 2

Helium to Water

N\imber of Tubes - 28

Tube Outside Diameter , inches - 0.5

Tube Wall Thickness , inches - 0.035

Tube Length, feet - 8-1/2

RESULTS

^ ^ Table I i s a summary of the per formance c ha ra c t e r i s t i c s calculated ^ ^ f o r the two pass sodium-to-hel ium heat exchanger in conjunction with the

he l ium- to -wa te r heat exchanger .

C-3

Des ign Da ta Sheet R - 5 0 0 1 , Rev . 1

T A B L E I

CASE I CASE II CASE III CASE IV CASE V**

Na - He - Hx

q, kw

^ N a i^l^*' ^

^Na °^*1^*' ^

^ N a ' 1^ /^^

^ N a ' SP"^

@, F

"^He ^''^^^' ^

Tj^ out le t , F

^ H e ' 1 ^ / ^ -

He - W a t e r Hx

q, kw

^ H e i"l^* ' ^

T^^ ou t le t , F

Wj^^, I b / h r

Tj^ Q in le t , F

T „ _ ou t le t , F

^ H ^ O ' gP"^

I n - R e a c t o r P o r t i o n of

PTT in le t , p s i a

P j j out le t , p s i a

T^^ i n l e t , F

T j j ^ ou t le t , F

Wj^^, I b / h r

( R e f e r e n c e C a s e )

1573

1400*

1079

55,800

140

1214

140

501

12,000

659

501

350

12,000

130

153

200

F a c i l i t y

600

500

140

350

12,000

2100

1400*

1062

70,800

180*

1215

140

482

16,950

730

482

363

16,950

130

155

200

600

400

140

363

16,950

1600

1150*

800*

52,000

125

954

140

400

16,950

530

400

302

16,950

130

148

200

600

400

140

302

16,950

1465

1022

800*

75,200

180*

896

140

378

16,950

513

378

295

16,950

130

147.8

200

600

400

140

295

16,950

0.25

800

A p p r o x . 8 0 0 *

76,400

1 80*

800

-

-

0

-

-

-

0

130

Approx . 1 30

200

600

600

-

-

0

* B a s e d on Des ign Objec t ives f r o m a L e t t e r f r o m H. M. L e p p i c h - IDO, to R. H. Gordon - E b a s c o S e r v i c e s I n c o r p o r a t e d . Subject : A T R Liquid Meta l Loop O b j e c t i v e s , 4 - 4 - 6 2 ,

** M i n i m u m h e a t e x c h a n g e r c a p a c i t y wi th in the D e s i g n Objec t ive l i m i t a t i o n s .

C - 4

PUMP y^iO

F 1 G 7 . 1

Na, TO H B . HE-^T E .XCHKNbER AND

He TO HzO H t M EXCH^^^GtR

JOB NO. p<b - 5 0 7 ^ CUSTOMER

SUBJECT TUhRKl<\L F ^ J X L i a u i P M L T A L P^ZKK<^i L o o p ^ H E M ^ X C ^ ^ N G E T ^ S BY W.KT. He C

DATE Ku& 27**- I9G2 FORM »OS7-8M-10-a«

APPENDIX D

Gamma and Neutron Heat Generation Rates

Design Data Sheet R-5002 59-3075-30 Contract No. AT(1 0- l ) -1075 May 11, 1962 D. R. Whitaker Revision 1, August 20, 1962

THE BABCOCK AND WILCOX COMPANY FOR EBASCO SERVICES INCORPORATED

PRELIMINARY DESIGN DATA ADVANCED TEST REACTOR

THERMAL FLUX LIQUID METAL PACKAGE LOOP GAMMA AND NEUTRON HEAT GENERATION RATES

Purpose and Scope

To determine the gamma and neutron heat that must be removed from the in-pi le pprtion of the the rmal flux loop.

General Descript ion

Figure 1 shows the heat generat ion ra tes in the var ious m a t e r i a l s in the loop ve r sus axial distance from core naidplane. Since the axial distr ibution is symmet r i ca l about the core midplane, the resu l t s shown on Figure 1 can be used both above and below core midplane . The total gannma heat load that mus t be removed is shown in Table I. Drawing SKC-2947A shows the configuration of the in -co re port ion of the loop on which this Design Data Sheet is based.

Since pa r t of the heat in the p r e s s u r e tube must be removed by the sodium coolant, the heat generation ra tes and total gamma heating in the p r e s s u r e tube a r e shown on Figure 1 and Table I, respect ive ly . The total heating in the p r e s s u r e tube is indicated separa te ly and thus is not in­cluded in the total shown in Table I because it is not known at this t ime how much of the heat will be removed by the sodium coolant and how much by the ATR coolant wa te r .

Neutron heating values also contribute to the total heating for the in-core portion of the loop, but these values a re unimportant when compared to the gamma heating va lues . Therefore , only gamma heating values a r e repor ted h e r e . Title I work will cover a detailed analysis and repor t of the neutron and gamma heating.

D-1

Table I

GAMMA HEAT IN ATR THERMAL FLUX LOOP

Reactor Power - 250 Mw

Loop Component

Fuel

Sodium up flow

Tes t Specimen Steel (Baffles, Cladding)

Test Specimen Containing Tube

Sodium downcomer

Subtotal for in -core portion of loop

Bottom 6 inches of loop

Por t ion of loop above core (1.5 feet)

Mater ia l

Uranium Carbide

Sodium

Stainless Steel, Type 316

Stainless Steel , Type 316

Sodium

Stainless Steel , Type 316 and Sodium

Average Heat Generation

Rate (Watts /cc)

147.481

4.057

42.499

48.731

5.116

Total Gamma Heating in Loop

Components (Kw)

166.98

7.94

33.78

49.21

15.52

273.43

6.33

12.53

Total 292.29

P r e s s u r e tube Stainless Steel , Type 316 59.765 145.01

D-2

L

tii-f

e o H

-01

g o to 1'

^ H

o

3 3 H-

1

•rt

o

•a

o

» 5 (D "C^

s v_

^

R

o ^

• o M

- -

—

n /* _

M

1 1

1 1

1 1

1 J_

1

1 1

1 1

1 1

£

1 1

1 1

1 1

1 u

[ 1

1 1

{ I 1

q;

1 m

ff

j 1

1 '

i 1

1 1

r !

1 1

1 1

1 1(

T

O

1 1

1 1

1 1

1 *£

1

1 I

! M

11

I T

I [

1 1

IIT

y

! 1

1 1

1 1

1 1

1 1

1 1

1 1

1^

E

I 1

1 1

1

1 li

^

T

! 1

1 1

1 I

i>

]<!

I LL

H

, L-M

Hn

u

fn]""

"" it'"

^"'

;: '1

^4

"^

11-1

1---

—t

fl fl

1414

1

ftll^

tr

•

1

T

Ml'

.''

1

1 1

' 1

' 1

1

1'.

^''

1

jy"

TI

1

* I

U|-H

J

|+{4

y [t

rrn

ffit"

H

-ffl;

LU

IJ-

iTrrT

1 1

1 i"*

\jr

[\

\ \\\

\\

1 1

1 1

1

rfifi

UU

-" r

r r

r If

" 0

1 1

Ml

f rri

r

rf

III'

' I

I

' '

1

\^\-

\-^

•• f

l }-

11 -

1

-1J

1 1

1

1 [1

i

^ 1

1

I [r

T

V

\\ '

" Mn

f"

-lff

l-4

1 fll

|T

L

11

tt

^ ^

jjl

tllL

t

i

11"'

' "l

,±

''' i

ll 11

-

' '' 1

'

11

'' ''

^ jj

' ;:>

J

j 1

i- "'

1

1

"T

I •

1

1 }

J

11-

-r I

Y G

amm

a H

ea

t

_ —

>

\

"TT

riT

r

j^\

1 11

1

' T

11

Ge

ne

rati

on

R

at*

0 u

>

>•

Mlll

ll H

ill

ILlm

1

Ui [I

N

ttft

tli

\'- r

rrr

1111

111

.<:

1111

tnlB

J""

1;

tttl

1 1

1

I 1 1

1

ki

II

• ¥

' 1

1 1

1 M

l ' i£

1

tilJj

T""

1-1

IT

tl

tt 1H

•'

rt

tl

1 it

n'

" f

1III'

\

\\\A

J 1

>u

IL

UJ

J j

U k

Sl

III

.1

X

AA

L

'• '

v\V

11

1 1

^'

- .'

' Iv

rT 1

J"

*/''

' 1

ix r

]i'].

='

'"' r

itf

iTy-

'f f,

pn

Tj-

t-

i! ffl-t

iilJJ-

j

.:l!U

i|

4-LL

LU---

+:

i U

ii

11)1

1

''

ill!

'

* 1

II1

+

Ni-

- II

JH

II

ffl-1

-—J

-li j

lfl-

t' 11

1

"•:!

ii

Tji

f •'

1; Iti

lirr

44J-J

1- -•

-

-rt]f[

-f]-

-Tj]-

• • -

• t

p-f ,

/wa

tts^

,j ^

^.

. 1

.^

.

JIJ

r I'

•• •

nil]"

•

• it

" 1

1::

::

:J,,

: 1

1 I

t 1

1 1

ll

J—

- -

| 1

•!•

-\:

^'A

:\\v

'y-

\—

. .a

J i

f 1

(,

(,

ll

. ..|

..

1 1 ?

"

[

1 '

^ »''"

1

1II

1.

( 1

W

^i^

'

ki,

,!'+

/

Ijt

;' |t

1

• L

^^

, k

l

U^

/ S

i'7

^ J

li,'

^ 2

*^''i

IJ

i 1

^. •

1 If

'i

ll

n .•

i t

J V

,

[ 1

1 ^s

.'

H»^

^''^

;

\-\

'i t

n-'

,--

••',

jtt

F'' I

T-

III

M

!•'

I '

1

W\'

IT

ii1"'

'[!""

lt II

lit'

' ''

}' L

1'

1 'J

i

1 ;

:. ^

J _|_

L

M

"nil

1 r•

'' •

'''

1

H

r :i

"""

"[

•- •

••

I'll]

.;

T 1

T^^

" ""

T i

i

\ --

.J

-j-

J1 11

-!-

1-1 s \ ' ,1

/ \ rt

{6 t°

h- *x

•-(

fli

0) C

(]> %

L_ /

M

U

4fc

i i n 1

^ 1

1 1

1 1

1

/ / Siv V\ hp

PT

S

i 1

1-

^ 1

1 1

1 1

1

? 9 1

IM

^ 1

1 1

1 II

TIT

IT iii

111

\\

i JJ

J J

L

£

' ok

' in

JJj

-.flj

ii 11

1 r

! 1

r

\ ^

\v

\ JI

UJI

...!

.,....

, L

ii

h

!''

1 '

\\\\

4±|:i

.;::

ns;

T

1 |_

M

t_L

! J

IN

01

m

<>j

CD

«

1

ill

11

•<—

' T

jlj]

"

T1

] 11

1]

1 T

"

-T+

- T

R

11J

B» '

j1

Ij J

1 IT

H '

t 1

-33

L _

., .

.-

ia I'

""" !

r '

|!1

0

r 1

H

|] '•T

I |tt'

]

'-v~

iiii

r '^

::

lj'

!|

iHri

t '

Ji

g n

a

] •

'• B

1

U

'' J.

•'

J

IT'

111

1 '

T'-

u\ 0

0

r

fi! ? C I 0 n I i \

\t

W

"^

APPENDIX E

Tempera tu re Distr ibution in In-Core Tubes

Design Data Sheet R-5003 59-3075-10 Contract No. AT(10-1)-1075 May 29, 1962 R. V. S t r aub /D . F . Judd Revision 1, August 21, 1962

THE BABCOCK AND WILCOX COMPANY FOR EBASCO SERVICES INCORPORATED

DESIGN DATA ADVANCED TEST REACTOR

THERMAL FLUX LIQUID METAL PACKAGE LOOP TEMPERATURE DISTRIBUTION IN THE IN-CORE TUBES

Purpose and Scope

The purpose of this Design Data Sheet is to p re sen t the maximum tempera tu re distr ibution in the in -core p r e s s u r e tube and flow baffles.

Descript ion

An analysis was miade to determine the maximum radia l t empera tu re distr ibution in the in -core p r e s s u r e tubes and flow baffles so the max i ­mum the rmal s t r e s s e s could be computed. The maximunn t empera tu re gradients in the p r e s s u r e tubes occur at the core midplane. The max i ­mum tempera tu re gradients in the flow baffles occur at the specimen sodium outlet. The factor controlling the location of the maximum tenn-pera tu re gradients in the p r e s s u r e tubes is the gamma heating, while the sodium tempera tu re difference is the controlling factor for the location of the maximum tempera tu re gradients in the flow baffles.

The per t inent data used in the analysis a re shown on the following tabulation.

ANALYSIS DATA

1) Flux Trap Power , Mw 50 (Nominal)

2) Gamma Heating Rate in Core Midplane 10 P r e s s u r e Tubes and Top and Bottom Flow Baffles - w / g - of Core 5

3) Sodium Flow Rate, Gpm 140

4) Sodium T e m p e r a t u r e s , F Test Section Inlet 1079

5) Reactor Inlet Water 130 T e m p e r a t u r e , F

E - 1

Design Data Sheet R-5003 Revision 1, August 21 , 1962

ANALYSIS DATA (Cont'd)

6) Contact P r e s s u r e Between Double Walled Fluted Containments , psi 400 to 800

7) Surface Finish at the Contact Surfaces , Micro- inch 30

8) Ar rangement as Shown on Sketch SKC-2811A-1

The average in ternal heat generat ion ra tes were obtained from Design Data Sheet R-5002. The contact coefficient for the p r e s s u r e tubes inner face and the flow baffle inner face was obtained from NACA Report TN-3295. The sodium tempera tu res were es t imated considering the heat convected to the reac tor coolant and the heat t r ans f e r r ed through the flow baffle. A tempera tu re r i se of 2 F was es t imated to occur in the r eac to r coolant from the flux t rap baffle inlet to the midplane of the r eac to r . The tempera ture of the coolant will not affect the t empera tu re gradients or t empera tu re level s ig­nificantly.

The t empera tu re dis tr ibut ions a re the sum of the t empera tu re gradient result ing fronm internal heat generation plus the t empera tu re gradients due to heat t r ans f e r r ed by convection at the boundar ies . In the case studied h e r e , the t empera tu re of the inner face of the inside p r e s s u r e tube is lower than the sodium t e m p e r a t u r e . The resu l t is that heat is conducted from the sodium to the reac tor wate r . If the inner t empera tu re had been higher than the sodium t e m p e r a t u r e , some of the heat generated in the p r e s s u r e tubes would have been absorbed by the sodium.

The t empera tu re distr ibution in a solid body in steady state with an in ternal heat source is descr ibed by the general differential equa­tion:

-K y ^ t = Q"'

Making all of the c lass ica l assumpt ions , neglecting axial conduc­tion, assuming no angular var iat ion and rewrit ing the equation in cyl indrical coordinates gives:

d^t ^ 1 dt _ • Q'" J 2 r dr k dr

Where t = Tempera tu re r = Radius

Q'" = Internal Heat Generation Rate k = Thermal Conductivity

E - 2

•

Design Data Sheet R-5003 Revision 1, August 21 , 1962

This equation was solved using the appropr ia te convection boundary conditions for the p r e s s u r e tubes and the flow baffle. The resul t ing mean and ex t reme t empera tu re s a r e tjhown on sketch SKC-281 l A - 1 .

Discussion

The discuss ion of the choice of the configuration used for the in -co re portion of the loop is presented in Design Data Sheet R-5008.

The model used for calculating the t empera tu re dis tr ibut ions has been simplified to facili tate a d i rec t approach to a solution. The Title I effort will include a detailed analys is which will a lso consider axial and angular va r i a t ions .

A very important considerat ion in the analysis is the contact r e s i s t ­ance between the p r e s s u r e tubes and at the interface of the flow baffle. The effects of radiat ion damage on the interface have noi. been e\ ;il':a.ted. This a r e a will be investigated in Title I.

•

E - 3

X^TAGN^m HELIUM:

90 FLUTE

TOP OF ACTIVE CORE

TEST SPECIMEN

v ^ ^ F L U X TR/ \P BA.FFL&

Ti'/^J^ -'-*\-'O70"

: ^

•OJO

TYPICAL SECTJON THROUGH FLUTED TUBE

V r ^ r t l / COKE MID-PLANE

2 - 6 2 / -RA.DH

T I

Tz l3 T * I f T e

IT

I 4 0 0 1379 134-3 1 2 6 2 N 4 5 l O S i I O T 9

• F " 'F "F •F *F

•F

T « T 9 Tio Til Ti2 T i5 Ti4.

1 0 6 5 I O E 4 9 3 9 7 1 9 ^ 0 2

191 1^52

»F •F "F "F ' F *F

T ^ 2-3 Tm 3-4 T»», ^-6 Jm 9-»0 Tm 10-11 Tfo 12-13

J362'*F 151 2 -F M 2 1 "F

*985''F S 3 9 ' F 351 -F

NOT TO SCALE

cuOTOMPt EBASCO SERVICES I N C . FOfg USAEC - IDO JOB N o . ^ 9 307^

^KC ^QWAZL •u^ttcrTHLKt^k^l Itfl^UjP MgTAL IK PILE LOOP -

DATE 2t- A U " ^ i r W j ^ j | |

APPENDIX F

Secondary Coolant Impuri t ies Activations

Design Data Sheet R-5004 59-3075-30 Contract No. AT(10-1)-1075 June 4, 1962 D. R. Whi taker /R. A.Benedic t Revision 1, August 20, 1962

THE BABCOCK AND WILCOX COMPANY FOR EBASCO SERVICES INCORPORATED

DESIGN DATA ADVANCED TEST REACTOR

THERMAL FLUX LIQUID METAL PACKAGE LOOP SECONDARY COOLANT IMPURITIES ACTIVATIONS

Purpose and Scope

To de termine the activation of impur i t ies in the secondary coolant.

General Descript ion

The secondary coolant (a commerc ia l grade of helium) contains impur i t ies that a r e activated during t rans i t through the in-pi le portion of the loop. These impur i t ies a r e * , in percent by weight:

Carbon Dioxide 0.000581

Argon 0.000049

Hydrogen 0.00003

Nitrogen 0.001763

Methane 0.000001

* F r o m Air Reduction Company Catalog 320, page 1, September 1955,

In addition, it is assumed that one gram of s ta inless s teel chips r e ­mains in the bottom of the in-pile heat exchanger after fabricat ion, is i r rad ia ted and then is blown out of the in-pi le portion of the loop by the secondary coolant.

Design Conditions

Secondary Coolant

Total t ime of secondary coolant i r rad ia t ion 10,000 hours

Equil ibrium activity exis ts at heat exchanger outlet. Nominal reac to r power applicable to the eas t flux t rap in which the loop is located is 50 mw.

F - 1

Design Data Sheet No. R-5004 Revision 1, August 20, 19<

Helium flow rate 2290 Ib /h r

Helium t empera tu re 500 F

Helium p r e s s u r e (average) 600 psia

Helium res idence time in r eac to r 0.22 s e c .

Total helium t r ans i t t ime , from 5 s ec . c o m p r e s s o r d ischarge to compres so r suction

Stainless Steel

Total t ime of i r rad ia t ion 10,000 and 2,000 hours

Location 1 foot above top of ATR core ,

The p r i m a r y coolant t r ans i t t imes vary as the inverse of the cool­ant flow ra tes for a given flow configuration. These t rans i t t imes , m turn , influence the coolant activation exponentially, i . e . , (b-dj ( ' ' ^~\r f \

where t , = core t rans i t and t . = total or loop t rans i t t ime . Thus for an inc reased flow rate there is a smal l dec rease in flux but because of the form of the exponential this is quite insignificant unless there is an unusually l a rge change in flow r a t e .

The p r e s s u r e and t empera tu re of the p r i m a r y coolant influence the density of the coolant which, in turn , direct ly influences the activation source s t rengths or flux. However, for helium, a change of p r e s s u r e and t empera tu re from 600 ps ia , 500 F to 500 psia , 300 F (the expected operating conditions) changes the density by l e s s than 4% which for these calculations is unimportant .

Results

Helium impur i t ies activation is repor ted in Table I.

Stainless Steel activation is repor ted in Tables I I and I I I .

Definition of G a m m a - C u r i e : Gamma activity in c u r i e s , equal to

3.7 X 10 gammas pe r second.

F - 2

Design Data Sheet No. R-5004 Revision 1, August 20, 1963

Table I

HELIUM IMPURITY ACTIVATION IN ATR THERMAL FLUX LOOP

Activity -, * Gamma Gamma Energy

Activation Source C u r i e s / c c Gas C u r i e s / c c (Mev)

Argon 4 . 0 5 4 3 x 1 0 " ^ ^ 4 . 0 2 5 9 x 1 0 " ^ ^ 1.37

Oxygen 8.255 x 10"-^^

Gamma 1 - 8.255 x 10"-^^ 0.200

Gamma 2 - 5.778 x 10" 1.37

Gamma 3 - 3.302 x 10"^^ 0.112

Nitrogen 9.5698 x lO"'^^

Gamma 1 - 9.570 x 10"^^ 6.13

Gamma 2 - 1.302 x 10"-^^ 7.10

Gamma 3 - 1.081 x 10"^^ 2.70

Total 4.0545x10"-^^ 4 . 0 2 6 2 x 1 0 " ^ ^

Notes:

1. Hydrogen and carbon produce no activation gammas and, t h e r e ­fore , a re not included above.

2. Helium density corresponding to above sources is 0.0036859 g m / c c .

F - 3

Design Data Sheet No. R-5004 Revision 1, August 20, 1 9 6 | ^

Table I I

STAINLESS STEEL ACTIVATION IN HEAT EXCHANGERS OF ATR LIQUID METAL PACKAGE LOOPS -

10,000 HOURS IRRADIATION

Activation Source

Iron Gamma Gamma Gamma

Manganese Gamma Gamma Gamma

Chromium Gamma

Cobalt Gamma Gamma

Nickel Gamma Gamma Gamma

1 2 3

1 2 3

1

1 2

1 2 3

Molybdenum MoV^ Gamma Gamma Gamma Mo99 Gamma Gamma Gamma

1 2 3

1 2 3

Gamma 4 Gamma MolOl Gamma Gamma Gamma Gamma

Sulphur Gamma

Silicon Gamma

Tot

5

1 2 3 4

1

1

al

Curie s /Gm of Stainless

1.0655 _ _ -

1.4587 ---

5.7288 -

2.8495 --

1.4289 ---

-7.4690

-• _

8.3887 _ ----

1.5100 ----

6.7079 -

3.6808 -

2.09536

X

X

X

X

X

X

X

X

X

X

X

s tee l

10-^

10-1

10-^

10-^

10-^

10-6

10-4

10-4

10-9

10-5

10-1

Activity

Gamm a Curie s /Gm of s ta in less Steel

., 6.0731 4.5814 2.9833

_ 1.4587 4.3762 2.9175

_ 5.6143

_ 2.8495 2.8495

-2.5863 3.5580 5.8584

--

7.4690 7.4690 7.4690

. 7.3821 8.3887 9.2277 8.3887 9.2277

. 1.4496 1.4119 5.2851 1.2835

-6.0371

• 2.5765

2.34816

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

< l°-5 1 0 ^

10:2 10-2 10 '

10-^

10:3 10 ^

10-5 10 ^

1 0 1 loi 10-6

1 0 : :

10-5 10 4 10 5 1 0 ^

1 0 : :

10 A 1 0 : 5 10 ^

10-9

10-8

10-1

G a m m a E n e r g y (Mev)

^ 1.100 1.290 0.191

• 0.85 1.81 2.13

-0.323

_ 1,1728 1.33

_ 1.12 1.49 0.37

--

0.26 0.68 1.48

-0.140 0.181 0.372 0.741 0.780

-0.515 0.083 0.896 2.08

-3.12

_ 1.26

F - 4

Design Data Sheet No. R-5004 Revision 1, August 20, 1962

Table I I I

STAINLESS STEEL ACTIVATION IN HEAT EXCHANGERS OF ATR LIQUID METAL PACKAGE LOOPS -

2,000 HOURS IRRADIATION

A c t i v a t i o n S o u r c e

I r o n G a m m a 1 G a m m a 2 G a m m a 3

M a n g a n e s e G a m m a 1 G a m m a 2 G a m m a 3

C h r o m i u m G a m m a 1

Coba l t G a m m a 1 G a m m a 2

N i c k e l G a m m a 1 G a m m a 2 G a m m a 3

M o l y b d e n u m Mo93 G a m m a 1 G a m m a 2 G a m m a 3 Mo99 G a m m a 1 G a m m a 2 G a m m a 3 G a m m a 4 G a m m a 5 M o l O l G a m m a 1 G a m m a 2 G a m m a 3 G a m m a 4

Su lphur ^ ^ G a m m a 1

TKlicon G a m m a 1

T o t a l

C u r i e s / G m of S t a i n l e s s

7 .7060 X . _ -

1.4587 X -. -

5.0114 X -

6.0474 X . -

1.4289 X . . -

.

7.4690 X -. .

8.3887 X -. . . .

1.5100 X . . . -

6.7079 X -

3.6808 X -

1.9982 X

S t e e l

10 -4

10-1

10 -^

10-4

10-3

10-6

10 -4

10-4

10-9

10-5

10-1

Ac t iv i ty

G a m m a C u r i e s / G m of S t a i n l e s s

. 4.3924 X 3.3136 X 2.1577 X

.

1.4587 X 4.3762 X 2.9175 X

. 4.9112 X

. 6.0474 X 6.0474 X

. 2.5863 X 3.5580 X 5.8584 X

-.

7.4690 X 7.4690 X 7.4690 X

. 7.3821 X 8.3887 X 9.2277 X 8.3887 X 9.2277 X

. 1.4496 X 1.4119 X 5.2851 X 1.2835 X

.

6.0371 X

_ 2.5765 X

2.2932 X

S tee l

io:t 10 5 1 0 ^

l o : ^

10 .2 1 0 ^

10-3

10- . : 1 0 ^

1 0 - :

10 5 1 0 ^

i o 1 10"6 10-6

l o : :

1 0 . 5 10 4

10 ^

1 0 - :

10 t

1 0 ^

10-9

10-8

10-1

G a m m a E n e r g y (Mev)

. 1.100 1.290 0.191

-

0.85 1.81 2 .13

. 0.323

-1.1728 1.33

_ 1.12 1.49 0.37

.

. 0.26 0.68 1.48

. 0.140 0.181 0.372 0.741 0.780

. 0.515 0.083 0.896 2.08

-

3.12

.. 1.26

F - 5

APPENDIX G

Sodium System Control and Safety System Cr i t e r i a

Design Data Sheet No. R-5005 59-3075-30 Contract No. AT( 10-1)-1075 August 28, 1962 J. D. Carlton

JUS BABCOCK AND WI LCOX COMPANY FOR EBASCO ^ERVICEE INCORPORATED

DESIGN DATA ADVANCED TEST REACTOR

THERMAL FLUX LIQUID METAL PACKAGE LOOP SODIUM SYSTEM CONTROL AND SAFETY SYSTEM CRITERIA

Purpose and Scope

The purpose of this design data sheet is to p resen t the genera l sodium system control c r i t e r i a and safety system c r i t e r i a for the Thermal Flux Liquid Metal Loop to be instal led in the Advanced Test Reactor .

The control and safety systenn design c r i t e r i a descr ibed here in a re conceptual, and are intended to serve as a start ing point for Title I work.

References

1. Design Data Sheet R-5001, Revision 1, The rma l Flux Liquid Metal Package Loop - The rma l Analysis of Helium Cooled Heat Exchanger, August 24, 1962.

General Descript ion of Thermal Flux Loop

Typical p rocess information (from R.eference 1) for the the rma l flux loop is l is ted in Table I below. This table shows the control ranges of the var iab les .

TABLE I

Thermal Flux Loop P r o c e s s Infoinnation (Based on Case 1, Ref. 1)

P r i m a r y Coolant Sodium

Secondary Coolant Helium

Heat Sink HDW System

( Demineral ized Water]

Sodium System Cover Gas Helium

Sodium Tempera tu re Range, F 250 - 1400

G - 1

TABLE I (Cont'd.)

Design Data Sheet No. R-5005

Sodium AT Range, F

Sodium Flow Range, gpm

Sodium Cover Gas P r e s s u r e Range, psig

Helium Coolant P r e s s u r e , psia

Helium Coolant Tempera tu re Range, F

Heliiim Coolant Flow Range, Ib /h r

0 - 350

0 - 180

0 - 6 0 (approximate)

600 S lu-i-i'lc oection inlet

140 - 1400 (P r imary Heat Exchanger)

140 - 350 (Reactor Inlet to Outlet)

0 12,000

P r o c e s s Information Required

P r o c e s s information requ i red for control, safety and exper imental information of the Thermal Flux Test Loop includes the following:

A. TEMPERATURES

Sodium System Tempera tu res In general , sodium tennpera-tu re s will be obtained from adjacent nnetal t empera tu re s )

Pump Discharge Punnp Suction Pump Sximp Tes t Specimen Inlet Tes t Specimen Tempera tu re Profile (via thermocouples

innbedded in inner helium annulus wall adjacent to the tes t specimen)*

Test Specimen Outlet ( same as P r i m a r y Heat Exchanger Inlet)

P r i m a r y Heat Exchanger Outlet

Sodium Piunp Tempera tu re

Bottom of Pump Motor Casing Pump Motor Stator Tempera tu re Pump Motor Bearings

G - 2

Design Data Sheet No, R-5005

TEMPERATURES (Cont'd.)

Stagnant Helium Tempera tu re

Inner Heliiom Annulus (Adjacent to specimen) Outer Heliiom Annulus (Adjacent to specimen)

Secondary Helium Tempera tu res

P r i m a r y Heat Exchanger Inlet P r i m a r y Heat Exchanger Outlet In-Pi le Inlet In-Pi le Outlet

* Feasibi l i ty of d i rec t specimen tempera tu re m e a s u r e m e n t s to be studied during Title I.

B. FLOW

Sodium System Flow

Sodi\am system flow is m e a s u r e d from cal ibra ted curves which re la te sodiiim flow to pump power and input frequency.

Secondary Helium System Flow

In-Pi le Flow

C. PRESSURE

Sodium System P r e s s u r e

Test Specimen Inlet and Outlet P r e s s u r e (Feas ib i l i ty to be studied during Title I)

Secondary Helium System

In-Pi le In le t In-Pi le Outlet

ELECTRICAL

Sodium Pump Motors

Power Consumption Input Frequency Line Voltage Stand-By Drive Output Voltage Motor Line Current

A.

D.

G - 3

Design Data Sheet No. R-5005

D. ELECTRICAL (Cont'd.)

Sodium Leak Detec tors

Output of Inner Helium Annulus Probe Output of Outer Helium Annulus Probe

E. HEAT BALANCE

Heat balance information may be derived from the p r o c e s s var iables l is ted above.

Genera l Control Cr i t e r i a , F igure 1

General control r equ i rements for p rocess var iables a re l is ted as follows:

A. TEMPERATURES

Sodium System Tempera tu re s

Sodium system t empera tu re s must be controlled (1) to maintain sodium AT a c r o s s specinnen, and sodium tempera ­ture at specimen outlet or inlet to des i red values for a given experiment, (2) to maintain sodium tempera tu re at specimen equal to or l e s s than 1000 F during periods when the reac tor is shut down and(3) to maintain sodium t empera ­ture above the freezing point when the reac tor is down or when the sodium loop is being removed or replaced.

• Secondary Helium System Tempera tu re s

Helium system inlet tennperature to the p r i m a r y heat exchanger mus t be controlled to approximately 140 F .

B. FLOW

Sodium System

The sodium system flow is to be controlled to obtain the des i red sodium AT a c r o s s the tes t specimen.

Secondary Helium Systenn

Heliiim flow is to be controlled in o rder to maintain the proper t empera tu re level of the sodium system.

G-4

Design Data Sheet No. R-5005

C. PRESSURE

Helium Coolant System P r e s s u r e

Helium coolant system p r e s s u r e must be controlled to approximately 600 psia at the in-pile inlet. Helium p r e s s u r e mus t be maintained at this value to provide the heat t r ans fe r capability requ i red to remove heat from the sodium sys tem.

Automatic Control of P r o c e s s Conditions

The automatic p roces s control system will be designed with suf­ficient flexibility to mee t reasonable demands of var ious exper imen t s . The control system will be able to provide di rect control of any two of the following:

Sodium inlet t empera tu re to specimen Sodium outlet t empera tu re from specimen Sodium delta T ( AT) a c r o s s specimen Sodium flow ra te Sodium inlet t empera tu re to p r i m a r y heat exchanger Sodivim outlet t empera tu re from p r i m a r y heat exchanger Sodi\im delta T ( AT) a c r o s s p r i m a r y heat exchanger

Fo r example, a control mode to maintain the sodium outlet t empera ­ture from the specimen and the sodiiim A T a c r o s s the specimen at constant values would function a s follows:

The specimen A T is continuously sensed and t r ansmi t t ed to a control ler where the incoming signal is compared to a set value of AT for a given exper iment . If the AT differs from the set point, the con t ro l l e r acts to adjust the frequency control rheo­stat on the pump drive assembly . This action in turn i nc rea se s or dec r ea se s the pump output in the direct ion to bring the specimen

A T back to the set point.

Sodium outlet t empera tu re from the tes t specimen is con­tinuously sensed and t r ansmi t t ed to a con t ro l l e r where the signal is compared to a set value. If the incoming signal differs from the set value, the control ler acts to inc rease or dec rease heliiun flow in the p r i m a r y heat exchanger and r e tu rn the sodium outlet t empera tu re to the proper value.

ATR Reactor Power Reductions

'The safety system c i rcu i t s for the The rma l Flux Liquid Metal Loop should be designed to reques t the appropr ia te r eac to r power reduction for abnormal conditions.

G - 5

D e s i g n Da ta Sheet No, R - 5 0 0 ^

T h e r e a r e four t y p e s of p o w e r r e d u c t i o n a v a i l a b l e for the r e ­a c t o r :

a) s c r a m b) fas t r e c o v e r y s c r a m c) fas t s e t b a c k o r s low s e t b a c k d) r e v e r s e

Of t h e s e , the s c r a m i s i n i t i a t e d f rom the safe ty s y s t e m , and the r e m a i n i n g c o r r e c t i v e a c t i o n s a r e c a r r i e d out by the r e a c t o r c o n t r o l s y s t e m s .

L i s t e d be low a r e a b n o r m a l p r o c e s s cond i t ions for which a l a r m c i r c u i t s should be p r o v i d e d to p r e v e n t d a m a g e to the t e s t s p e c i m e n , the i n - r e a c t o r p o r t i o n of the loop o r the r e a c t o r . Some of t h e s e con ­d i t i ons nnay be u s e d to i n i t i a t e r e a c t o r p o w e r r e d u c t i o n s if r e q u i r e d . The d e t e r m i n a t i o n of a l a r m and power r e d u c t i o n se t po in ts wi l l be m a d e du r ing T i t l e I .

a) L o s s of, o r low sod ium flow

b) High sod ium t e m p e r a t u r e

c) High t e m p e r a t u r e in o u t e r o r i n n e r he l ium annul i

d) L iqu id m e t a l l e ak into i n n e r o r ou t e r he l ium annul i

e ) L o s s of, o r low h e l i u m flow

f) L o s s of, o r low h e l i u m p r e s s u r e

g) L o s s of e x p e r i m e n t c o n t r o l power

h) L o s s of r e a c t o r c o n t r o l power

i) L o s s of, or low HDW flow

j) High t e m p e r a t u r e of he l i um c o m p r e s s o r m o t o r b e a r i n g s

k) High heliiom c o m p r e s s o r m o t o r power c o n s u m p t i o n , l ine vo l tage o r l ine c u r r e n t

1) High o r lov/ sod ium p u m p m o t o r p o w e r

m ) High sod ium p u m p m o t o r l e a d c u r r e n t

n) Low sod ium p u m p d r i v e output vo l tage

o) Open sod ium p u m p m o t o r c i r c u i t s

G - 6

Design Data Sheet No, R-5005

Although controls and a l a r m s on the helium c o m p r e s s o r s and HDW system a r e not essent ia l to operat ion of the in-pile portion of the loop, those indicated above serve as anticipatory failure signals to help prevent deleter ious conditions from existing before other signals a r e effective.

G-7

In-Pile

I Outer Stagnant Helium Annulus

LEGEND

T - Terr^erature

L - Lsvel

In-Pile Loop Instrumentation Diagram

FIGURE I

Glosiire Plug

APPENDIX H

Maximum Physica l Envelope

Design Data Sheet R-5006 59-3075-30 Contract No. AT(10-1)-1075 November 15, 1962 W. S. Har r i s

THE BABCOCK & WILCOX COMPANY FOR EBASCO SERVICES INCORPORATED

PRELIMINARY DESIGN DATA ADVANCED TEST REACTOR

THERMAL FLUX LIQUID METAL PACKAGE LOOP MAXIMUM PHYSICAL ENVELOPE

Purpose and Scope

The purpose and scope of this Design Data Sheet is to define the physical envelope within which the Thermal Flux Liquid Metal Package Loop must fit. The limitations a re defined as those peculiar to but not necessar i ly limited to the east outer flux t rap location in the Advanced Test Reactor which has been designated as the position for this package loop.

Description

Several features of the ATR reac tor determine and dictate the dimensions of the Thermal Flux Loop:

1. The in-core section up to the top of the flux t rap baffle is res t r ic ted to a maximum diameter of 4 in. by the water gap required according to the physics calculations for the neutron spectrum des i red .

2. The a rea for 180 degrees of the periphery of the loop for 6-1/2 ft above the top active fuel line of the reac tor fuel elements is l imited to a radius of 2-9/16 in. from the center of the ver t ical center line of the loop. This r e ­striction is to allow removal and replacement of the ATR fuel elements with­out disturbing the loop.

3. The other 180 degrees of the per iphery of the loop i s l imited to the flared a r ea directly over the beryll ium core reflector and to a maximum radial distance out from the ver t ica l centerline of the loop of 7-1/2 in. This 7-1/2 in. dimension is derived from a maximum diametra l distance of 12 in. for the loop thimble in the canal t ransfer tube,

4. Above the section in (3) the loop for the same 180 degrees as in (2) is r e s t r i c t ed to a maximum radial dimension of 4 -1 /4 in. This res t r ic t ion is i m ­posed by the proximity of the three (3) adjacent loops (8-1/2 in. on centers) for a distance of 40 in .

5. The other 180 degrees of the per iphery of the loop in the region men­tioned in (4) may be l a rge r than 4 -1 /4 in. radius , but is r es t r i c ted to the l imits in (3) for a distance of 40 in,

6. Above the section described in (4) and (5) the loop necks down to a c i r ­cular c ross section 5-15/16 in, in diameter with i ts center coincident with the in -core section center l ine. This right cylindrical section must extend for a minimum distance of 10-1/2 in, to allow the loop to be pulled through the loop access hole (6 in, diameter) in the reac tor vesse l top closure plate so that loop services may be made up or broken when the loop is installed or removed,

7. The total length of the loop assembly is dictated by the clear distance between the bottom of the top closure plate (El. 100 ft-0 in,) and the top of the shim drum drive gear box (El. 83 ft-2 in.) . Allowance must be made for han­dling clearance (min, 3 in,) and a handling extension on top of the loop (approx­imately 8-1/2 in . ) . The maximum loop length without the handling extension therefore is 15 ft-11 in.

These dimensions and limitations a re graphically shown on the attached sketch (SKC-2934-C),

H-1

•

APPENDIX I

P re l imina ry Phys ics Per formance of the Reference Fuel Test Specimen

•

Design Data Sheet No. R-5007 59-3075-30 Contract No. AT( 10-1)-1075 August 8, 1962 E. J, P ie rczynsk i /R . J, Neuhold

THE BABCOCK AND WILCOX COMPANY FOR EBASCO SERVICES INCORPORATED

DESIGN DATA ADVANCED TEST REACTOR

THERMAL FLUX LIQUID METAL PACKAGE LOOP PRELIMINARY PHYSICS PERFORMANCE OF THE

REFERENCE FUEL TEST SPECIMEN

Purpose and Scope

The purpose of this design data sheet is to present the r e su l t s of a p re l iminary physics study of an ATR in-pi le sodiuxn loop for test ing r e ­ac tor fuels in a high the rma l flux environment and to re la te the tes t de ­sign objectives to the resu l t s of the calculat ions.

General Descript ion

Although it is p resen t ly expected that the tes t specimen will be located in the east flux t r a p of the ATR core between two outer lobes operating at 60 xnw and 40 mw, al l of the calculations a r e per formed for a one-diinensional (cylindrical geometry) outer single lobe model using ei ther the B&W Multi-Group (16 groups) Diffusion P r o g r a m or the B&W Two-Group Lifetime P r o g r a m . With this model, the tes t specimen power is obtained by normaliz ing the core sect ion of the single outer annular lobe model to 50 mw. The decision for such normal izat ion and the use of an outer single lobe model a r e based on the resu l t s of a full core l ife­t ime study for the ATR which included therxnal exper iments in outer and ex ter ior lobes. See Section C "Two-Dimensional Analys is ,"

I - l

Design Data Sheet No. R-5007

A. Configuration and Composition

The radia l configuration and composition used for the reference single lobe calculations a r e shown in Table I. Region 1 in Table I r ep resen t s the tes t section obtained by homogenizing seven s ta inless s teel clad uranium carbide fuel pins with sonne s t ruc tura l mate r ia l , space r s , and sodium. A reference uranium enr ichment of 2.54 a /o is used, derived from the 1,500 k^v pow^er requi rement and an es t imated density of 13.6 g m / c c for uranium carbide . The resul tant f issi le nnaterial density (grams of f issi le mate r ia l per unit volume of uranium carbide) is 0.324 g m s / c c . Region 2 is the outer portion of the loop. Region 3 is the water gap between the loop and the flux t r ap baffle (Region 4). Region 5 is the ATR core . Regions 6 and 7 a r e the ref lector regions . Elennental nunnber densi t ies a r e a toms in-24 r e ­fe r red to the par t icu la r region. cc

F igure I desc r ibes the radial geometry . The dimensions of Region 4 through 7 a r e set by ATR design. The flux t r ap water thickness is about the minimum that will pe rmi t attaining the required f a s t - to - the rma l power ra t io . The Region 1 radius is that required for the reference tes t spec i ­men geometry . The thickness of Region 2 is based on the space remaining after considerat ion is given to the other regional space requ i rements .

Table I

Radial Configuration and Composition for Reference Full Length One-Dimensional Calculation

Region Number

1

2

3

4

5

6

7

Outer Radius, cm

2.606

5.080

6.667

7.4613

14.1288

33.6055

67.2605

Physical Tempera tu re , ]

1,000

700

160

155

163

187

130

<" Elements

C Na F e

Na

H

A l

H B(nat)

0

H Be

H 0

Number Densit ies

0.013646 0.012015 0.0043231

0.011205

0.06560

0.060275

0.036767 0.00002549 0.01907

0.006860 0.11000

0.062312 0.031156

Elements

U-235 U-238

F e

0

A l U-235 U-238

0

A l

Number Densit ies

0.0003467 0.0132995

0.033902

0.03280

0.02493 0.0002417 0.00001713

0.003430

0.0031928

1-2

Design Data Sheet No. R-5007

B. One-Dimensional Analysis

All one-dimensional calculations a r e done for a single outer lobe model with tes t powers normal ized to a single lobe core power of 50 nnw. Axial leakage effects a r e accounted for in al l of the regions .

A summary of one-dimensional calculat ions is included in Tables I I A and I I B. The peaking factors in Table I I A a r e based on the total power d is t r ibu­tion while the peaking factors in Table I I B a r e based on only the t h e r m a l power distr ibution. The core power radia l max imum- to -ave rage l is ted in Table I I A compares favorably with values calculated with A - 1 , A - 3 , and A-5 exper iments* in a single lobe model. In ei ther case , however, axial peaking effects and the detailed effects such as azimuthal core peaking a re not included. Such effects will be studied in Tit le I.

Fuel Pin Enrich­

ment a/o

Fissile Material Density,

g/cc Fuel

Single

Lobe Keff

Table II A

One-Dimensional Power Summary Single

Test Power

Kw

Exterior Lobe Core Power

Avg. Test Power Avg. Density Rod

Kw/cc of Power Test Region Kw/in.

= 50 Mw

Power Ratio Fast to

Th, {,625 ev. cutoff)

Test Power Radial

Max/Avg.

Test Power Radial

Max/ Min

2.54 0.324 1.068 1,500 0.57 4.46 0.32 1. 15 1.34

Core Power Radial

Max/Avg.

1.51

Table 11 B

One-Dimensional Thermal Peaking Factors

Test Radial Max/Avg.

Th. Pwr. Only

1.22

Test Radial Max/Min Th. Pwr. Only

1.51

Core Radial Max/Avg.

Th. Pwr. Only

1.56

* The A-1 , A-3 , A-5 thermal tests are water cooled tests rather than sodium and differ from each other only in the U-235 concentration. These tests contain H2O, Zr-2, and U-235 in a lattice with a metal-to-water ratio of 1.0. The homogenized elemental number densities are:

N atoms -24 -x 10

Element

H 0

Zr-2 U-235

cc

0.022946 0,011473 0.02125 0.00002449(A-1); .00004889(A-3); . 00007 348{ A-5)

The flux t r aps surrounding the A - 1 , A - 3 , and A-5 exper iments a r e 3.015 cms thick and may contain varying fractions of H2O and Al. In this study the A-1 and A-5 flux t r aps a r e essent ia l ly H2O.

1-3

Design Data Sheet No. R-5007

C. Two-Dimensional Analysis

Full core studies with an ex te r ior flux t r ap containing the test have not been done. However, full core studies with tes t locations containing A - 1 , A - 3 , and A-5 t es t s can be used as a bas is for single lobe power normal iza t ion and to de termine lobe a symmet ry effects. This study shows that the ra t io of the average power density of an A-5 tes t in an exter ior flux t r ap (between an A-1 tes t in an outer lobe operating at approximately 60 mw and an A-5 test in an outer lobe operating at approximately 40 mw) and an A-5 test in an outer lobe operating at approximately 40 mw is about 1.25. Consequently, a core power of 50 mw is used for nornnalization of the one-dimensional single outer lobe calculat ions. The configuration of lobes is shown in F igure I I . In the ATR design effort the reference design is based on a 60-50-40 mw power split . In Figure I I , a 60-50-40 power split means that the center fliix t r ap operates at 50 mw, two outer lobes operate at 60 mw each and the remaining two outer lobes operate at 40 mw each. F igure I I shows that the the rmal flux test , in the east flux t rap , is subjected to effects of adjacent lobes and center flux t r a p s imi la r to those effects to w^hich the A-5 tes t mentioned above is subjected.

Lobe a s y m m e t r y effects can be seen in Table I I I which shows the tes t maximum to minimum surface power density and the maxinnum to min i ­mum four-group tes t surface flvixes for the A-5 tes t in an ex te r ior flux t r a p of the ATR core . These data and the resu l t s of the data in Table I I A show^ that the design c r i t e r ion on the radial tes t power depress ion will not be exceeded even if full lobe a symmet ry is included. The sodium cooled 48-in. t he rma l tes t will have no objectionable effects on adjacent exper i ­ments due to i ts nuclear and geomet r ica l s imi la r i ty to A - 1 , A-3 , and A-5 t e s t s .

Angular Power

Max/Min

Group 1 Angular Flux

(.01 Mev - lOMev) Max/Min

Table I I I

Lobe Asymmetry (Exterior Flux Trap)

Group 2 Angular Flux

(.045 Kev - . 01 Mev) Max/Min

Group 3 Angular Flux

(. 625 ev - . 045 Kev) Max/Min

Group 4 Angular Flux

(Thermal) Max/Min

1.13 1.29 1.20 1.15 1.14

1-4

Design Data Sheet No. R-5007

D. Lifetime Calculation

A one-dimensional lifetime calculation using the B&W Combined Cycle Lifetime P r o g r a m was done for the single lobe reference con­figuration descr ibed in Table I. In this calculation the sample ' i s i r ­radiated for a total t ime of 34 days with the single lobe core section operating at approximately 50 mw. After 17 days of operat ion (one ATR core cycle), the annular single lobe core is replaced and the sample i r ­radiated for an additional 17 days.

The data in Table IV show how the total lobe power of 51.5 mw (core + test) is shared during sample i r radia t ion. These resu l t s indicate no severe shifting of power between the core and tes t as a function of i r rad ia t ion t ime. However, the calculations were only run for two core l ives (34 days). Fo r much longer i r rad ia t ion t imes shifting of power may occur. This ef­fect will be further investigated in a Tit le I effort.

F igure I I I shows the integrated fission r a t e s per cc of uranium carbide in the tes t for each of the fissionable tes t e lements including the Pu-239 produced from the U-238 absorpt ions . The total in tegrated f ission ra te , i .e. , the sum for all the elements is shown in F igure IV.

Table IV

Power Distr ibution

Time (Days)

0

1

5

10

17

25

30

34

F r a Lobe

ction of Total Power in

0.9577

0.9581

0.9594

0.9597

0.9627

0.9623

0.9619

0.9613

Core F r a

Lobe ction of Total Power in Test

0.0423

0.0419

0.0406

0.0403

0.0373

0.0377

0.0381

0.0387

1-5

Design Data Sheet No. R-5007

E. Design Objectives and Summary

A comparison of the design objectives a/ pertaining to the physics p e r ­formance of the test specimen and the resul ts of this pre l iminary analysis a re shown in Table V. In Table V, TBD means "to be determined by the loop designer."

Table V

7

8

9

10

11

12

13 14 15 16 17 18 19 20

21 22

Type of Fuel F iss i le Material F i ss i le Material Density,

g/cc fuel Type of Spectruin Radial Power Depression,

initial, Max/Min Angular Power Variation

Maximum Power Ratio, Fas t / Thermal (0.625 ev cutoff)

Fuel Burnup 10^0 f iss ions/ cc fuel

Burnup Increment, 10^0 f i s -s ions /cc fuel

Specimen Power Generation, (Fission) kw

Average Pin Power (Fission), kw/in.

Maximum Pin Power ( F i s ­sion), kw/in.

Number of Pins Length, in. Pin Outside Diameter, in. Fuel Outside Diameter, in. Bond Mater ia l Cladding Material Clad Thickness, in. Geometry

Spacer Outside Diameter, in. Minimum Test Section

Diaineter, in.

Design Objectives gV

Ceramic (UC) Uranium

TBD (0.4-1.2) Thermal

Results of Analysis

Ceramic (UC) Uranium

0.324 by Thermal

1.5-2.5 1.34 Uniform as possible 1.07 (Surface Max/Avg.)

(1.13 Max/Min)

0.30

5-15

1-3

150-1,500

5

9 5-7 48 (max) 0.560 0.500 Sodium Stainless St 0.020 Triangular

Spiral Spa 0.090

2.35

^. M. Leppi

ee l

Pitch, cers

ch to R.

0.32 5-15 ( ~ 4,100 QJ to

12,300 cj h r s @ 1,500 kw) 1-3 ( - 8 2 0 to 2,500 c/

h r s @ 1,500 kw)

1,500

4.5

8.5 (hot spot) 7 48 0.560 0.500 Sodium Stainless Steel 0.020 Triangular Pitch,

Spiral Spacers 0.090

2.425 (max)

H. Gordon. a/ F rom April 4, 1962 le t ter from H

b / The specified range of fissile mater ia l density leads to excessive specimen power generation. Discussion with the AEC indicated that a reduction in the fissile mater ia l density (below the specified range) would be a satisfactory method of reducing test power, so long as the density remained grea te r than that of natural uranium fuel which has a density of about 0.1 g m / c c .

cJ These fuel burnup t imes a r e est imated values and a r e based on an extrapola­tion of F igures I I I and IV.

1-6

3 " n 01

o>

3 Lobe and S p e c i m e n C e n t e r l i n e

o T e s t S p e c i m e n - R e g i o n 1 a-

rf>- Loop Structure - Region 2

ui Flux T r a p Water - Region 3 00 -J

-J Flux T r a p Baffle - Region 4 1

'<y- ATR Core - Region 5 00

- J Reflector - Region 6

C3

'(y. Reflector - Region 7

V p

tu )—' 0 ft 0

3 a> rt-H ^

^ 1—1

0 a w M 1—1

N

f

FIGURE H f^TR LOBB CONPKSU/Zf^TlON

<SOf/IV/ LOBB. 4-0 MW LOBE

SOMW PLUX TRAP

40 MW LCie>£ <iO MW LOBE

i

APPENDIX J

Heat Trans fe r Control Philosophy

Design Data Sheet No. R-5008 59-3075-30 Contract No. AT( 10-1) - 1075 August 13, 1962 J. E. Lemon/D. F . Judd

THE BAB COCK AND WILCOX COMPANY FOR EBASCO SERVICES INCORPORATED

DESIGN DATA ADVANCED TEST REACTOR

THERMAL FLUX LIQUID METAL PACKAGE LOOP HEAT TRANSFER CONTROL PHILOSOPHY

Purpose

To present the resu l t s of an investigation to de te rmine : (1) if the var iable liquid level method of heat t r ans fe r control, as proposed in APDA-145, is feasible with a liquid other than m e r c u r y and (2) if the var iable liquid level nnethod is the best scheme for controlling the heat t r ans fe r ra te in the heat exchanger and in -core portion.

Introduction

The package loop concept in APDA-145, "Conceptual Design Of An In-Pi le Package Loop F o r Sodium - Cooled The rma l Reactor Fue l Test ing," uses a controllable level of m e r c u r y in the heat exchanger section and the in -core portion of the loop to vary the heat t r ans fe r ra te from the p r i m a r y coolant sodium to the reac to r water . The var iable level concept is p r o ­vided (as stated in APDA-145) in the in -core port ion to reduce specimen t empera tu re t r ans ien t s during abnormal operating conditions such as the loss-of-sodium flow. This is done by rapidly filling the annulus with m e r c u r y or helium to inc rease or dec rease heat t ransfer , as requi red . The same var iable level concept is used also in the heat exchanger s e c ­tion to vary the effectiveness of the heat exchanger and thereby control the t empera tu re level of the p r i m a r y sodium coolant.

Since mercu ry readily at tacks aluminum, a decision was made to el iminate mercu ry* as the in te rmedia te heat t r ans fe r fluid due to the possibi l i ty that it might escape into the main coolant water and thereby attack the ATR aluminum fuel e lements . An investigation was conducted to determine if the var iable liquid level method of heat t r ans fe r control is feasible with a liquid other than mercu ry . An investigation was also made to de termine an a l ternat ive method of heat t r ans fe r control .

* Scope Document for P repara t ion of Design Cr i te r ia , Liquid Metal Loops, Advanced Test Reactor - USAEC-IDO, January 2, 1962.

J - 1

Design Data Sheet No. R-5008

Discussion

Fluids for Variable Level Control

The approach taken in this investigation to find a substitute for m e r c u r y was to sea rch through l i t e ra tu re (see References at end of this Design Data Sheet) for fluids -whose p roper t i es might permi t them to be used for this application. The proper t i es receiving most considerat ion were melting point, boiling point, t he rma l conductivity, thermal stability, radiation stability, cor ros ive effect on 316 s ta inless steel, react ion rate with other coolants, and handling precaut ions .

As a resul t of the l i t e ra tu re search only one c lass of fluids, liquid meta ls , was found to have proper t i es making it suitable for consideration. Other c l a s ses of fluids such as liquid organics , gas - so l ids , and l iquid-solids were eliminated for var ious r easons . The liquid organics a r e not stable under i r rad ia t ion and high t empera tu re conditions. T a r s a r e formed which could foul heat t r ans fe r surfaces and plug the flow passages thus preventing a change of liquid level . The gas -so l ids and l iquid-solids were not considered due to the unknown technology. Considerable r e s e a r c h and development would be requi red before any of these fluids could be considered.

A p r i m a r y considerat ion in the selection of a liquid metal as the in te r ­mediate heat t r ans fe r fluid is the melting t empera tu re . Ideally, the mel t ­ing t empera tu re should be below room t empera tu r e . However, since the p r i m a r y coolant sodium must be kept above 208 F (melting point), then it may be possible to use a fluid with a melting point up to about 200 F by utilizing some of the heat supplied to the sodium. A la rge number of the liquid metals were not considered because thei r melting t empera tu re s were considerably above 200 F, requiring supplementary heating for which space must be provided.

Some liquid metals have desirable melting points but were eliminated for other r easons . Fo r example, the liquid metal cesium was eliminated due to the inconclusive study of its at tack on 316 s ta inless s teel . More r e s e a r c h would be requi red on this problem. The liquid metal gallium has a sat isfactory melting point but it is highly cor ros ive . It a t tacks nearly all meta ls except tantalum, zirconium and tungsten^. Some experiments have shown that the addition of inhibitors such a s indium will reduce this problem^.

As a resu l t of the l i t e ra tu re sea rch two liquid meta l s , NaK-44 and rubidium, were selected for comparison as in termedia te heat t ransfe r fluids. Table 1 l i s ts the proper t ies of these l iquids. The following presen ts the important aspects of each liquid:

J - 2

Design Data Sheet No. R-5008

Table 1

P rope r t i e s of Intermediate Heat Trans fe r Fluids

Physical P roper t i e s * 3

Density, lb/ft

Viscosity, Ib /hr- f t

Vapor p r e s s u r e , nnm Hg

Thermal P rope r t i e s

Melting point, F Boiling point, F

Thermal conductivity, B tu /h r - f t ,F *

Heat Capacity, Btu/ lb - F *

Heat of vaporization, Btu/ lb

General P roper t i e s

Corros ion of s ta in less s teel , type 316 at 1,200 F

Reaction w ith water

Toxicity

F i r e or explosive hazard in a i r

NaK-44

49

0.5

Negligible

66

1,518

15.1

0.25

Rubidium

84

0.5

134

102

1,290

14.3

0.0877

Negligible

Violent

Moderate

High

363

Negligible

Violent

Moderate

High

* P r o p e r t i e s a r e at 1,000 F .

J - 3

D e s i g n Data Sheet No. R - 5 0 0

N a K - 4 4

T h i s l iquid m e t a l i s an a l loy of sod ium and p o t a s s i u m . The t echno logy of i t s u s e in t h e r e a c t o r f ie ld i s we l l known. N a K - 4 4 m e l t s a t 66 F which e l i m i n a t e s the need for h e a t e r s at r o o m t e m p e r a t u r e . The boi l ing t e m ­p e r a t u r e of NaK is 1,518 F which m e a n s tha t l i t t l e , if any, o v e r p r e s s u r e would be r e q u i r e d to p r e v e n t bo i l ing . NaK is not a f fec ted by high r a d i a ­t ion and h igh t e m p e r a t u r e s . S t a i n l e s s s t e e l h a s good r e s i s t a n c e a g a i n s t a t t a c k by NaK. The m o s t s e r i o u s p r o b l e m with NaK is the v io lent r e ­a c t i o n tha t o c c u r s when it c o m e s in con tac t wi th w a t e r . T h i s would n e c e s ­s i t a t e a double w a l l e d c o n s t r u c t i o n to p r o v i d e a n annu lu s b e t w e e n the NaK a n n u l u s a n d the ATR r e a c t o r w a t e r .

Rub id ium

/

T h i s l iquid m e t a l h a s p r o p e r t i e s s i m i l a r to NaK excep t tha t it i s m o r e r e a c t i v e in a i r and the t echno logy of i t s u s e i s not a s we l l known. I t s b o i l ­ing point (1,290 F ) i s s i i ff icient ly high so that l i t t l e o r no o v e r p r e s s u r e would be r e q u i r e d to p r e v e n t bo i l ing . I ts m e l t i n g t e m p e r a t u r e of 102 F i s h i g h e r t h a n for NaK and e x t e r n a l heat m a y be r e q u i r e d to k e e p it in the l iquid s t a t e at r o o m t e m p e r a t u r e dur ing hand l ing o p e r a t i o n s . T h i s a m o u n t of h e a t would be s m a l l , h o w e v e r , due to i t s s m a l l hea t c a p a c i t y (0.0877 B t u / l b F ) . Rub id ium is s i m i l a r to NaK with r e s p e c t to i t s r e a c t i o n with w a t e r and the r e q u i r e m e n t for a double wa l l ed c o n s t r u c t i o n to p r e v e n t i t s con tac t with w a t e r . T h e cos t of r u b i d i u m is high in c o m p a r i s o n to NaK ( a p p r o x i m a t e l y $ 7 9 5 / l b c o m p a r e d to $ 0 .16 / Ib ) . *

B a s e d on the above c o m p a r i s o n , i t s e e m s r e a s o n a b l e to s e l e c t NaK-44 o v e r r u b i d i u m b a s e d on the t echno logy of the u s e of NaK-44 be ing g r e a t e r and b a s e d on the l o w e r cos t of NaK-44 a s c o m p a r e d to r u b i d i u m with both having s i m i l a r p r o p e r t i e s . T h u s , the v a r i a b l e l e v e l concep t d e s c r i b e d in A P D A - 1 4 5 could be r e t a i n e d with the u s e of NaK in the p l a c e of m e r c u r y .

I n - C o r e P o r t i o n

T h e r e a r e d i s a d v a n t a g e s i n h e r e n t in the i n - c o r e v a r i a b l e l iquid l eve l s c h e m e .

1) The c o v e r g a s p r e s s u r e m u s t be i n c r e a s e d o r d e c r e a s e d t o v a r y t h e l iquid l e v e l . T h i s c o n t r o l , inc luding g a s feed and b leed , would be h ighly c o m p l e x .

2) T h e r e i s a n add i t iona l t i m e de l ay involved f rom the t i m e a s i g n a l i s i n i t i a t e d t o i n c r e a s e o r d e ­c r e a s e the hea t t r a n s f e r r a t e unt i l the l iquid l eve l i s p h y s i c a l l y changed . Any lag in hea t r e m o v a l f r o m the s p e c i m e n i n c r e a s e s the chance of the s p e c i m e n o v e r h e a t i n g .

* R e c e n t m a n u f a c t u r e r ' s e s t i m a t e .

J - 4

Design Data Sheet No. R-5008

3) The external control connections inc rease the difficulty in handling the loop in and out of the r eac to r .

4) The NaK system requ i res complete double con­tainment, including al l ex ternal feed and bleed piping, valves, fittings, tanks, etc .

5) The var iable level concept is shown in sketch SKC-2948-A. This sketch shows that only a 1.469-inch space is available for the specimen. This is not sufficient space to accomnnodate the reference tes t specimen which requ i res 2.425 inches. The th icknesses for the var ious annuli and baffles shown on SKC-2948-A were sized as follows:

Flux t r ap baffle - The d iameter and thickness were establ ished from de­sign work on the ATR core .

Flowing water annulus - This thickness was sized to give the proper neutron spec t rum in the tes t specinnen.

Level Control annulus (NaK and helium) -This thickness was designed to prevent plugging and to minimize retent ion of NaK on the walls of the annulus v/hen the level is lowered.

Flow^ing sodium annulus - This was sized to give a reasonable sodium p r e s s u r e drop.

Grooves or flutes - Sized to permi t the routing of thermocouple w i r e s and to keep heat t r ans fe r r a t e s high.

316 s ta inless s teel walls - All th icknesses a r e es t imates based on approximate t e m ­pe ra tu re s and p r e s s u r e s .

The only annulus that could possibly be dec reased in size would be the flowing sodium annulus. However, even by eliminating this space completely, the spec i ­men space could not be increased sufficiently to p e r ­mit the instal lat ion of the reference specimen.

A new design has been conceived that will per form the requi red function and will not have the disadvantages of the var iable level concept. A plan c ro s s section of the new design is shoMOi in sketch SKC-2947-A.

The new^ design rep laces the var iable level concept with two walls in |close contact. F lu tes a r e provided between the tw o walls to contain s tag­n a n t helium which se rves as the in termedia te fluid. The fins formed b e ­tween the two walls pe rmi t a high ra te of heat t r ans fe r from the sodium coolant to the ATR water . The annuli and wall th icknesses were s ized as descr ibed previously.

J -5

Design Data Sheet No. R-5008

The fluted design is super io r to the var iable level concept for the following r ea sons :

1) The fluted design allows a l a rge r heat t r ans fe r ra te between the sodium and the water during emergency operat ions than that attained with the var iable level scheme.

2) The fluted design t r ans f e r s more heat direct to r eac to r water during normal operat ion than the var iable level concept thus allowing the use of a sma l l e r heat exchanger.

3) No liquid level control system is requi red for the fluted design.

4) No external connections pertaining to liquid level control a r e requi red for the fluted scheme.

Heat Exchanger Por t ion

The disadvantages d iscussed under the In-Core Port ion apply a lso to the heat exchanger, except #2 (control t ime delay) which is not con­s idered to be a ser ious disadvantage in the heat exchanger. In addition, the control of level is difficult in that smal l var ia t ions in gas p r e s s u r e resul t in relat ively large changes in liquid level . As the liquid level va r i e s , t he rma l gradients through the containing walls change, p roduc­ing a fatigue condition that may resul t in p r e m a t u r e fai lure of the s t ruc tu re .

A new heat exchanger design has been conceived that has none of the disadvantages of the var iable level concept. The heat exchanger conceptual design consis ts of a number of para l le l tubes through w^hich the p r i m a r y coolant sodium flows and around -which secondary coolant helium flows. A re la t ively simple control system, external to the reac tor , va r i e s the helium flo-w rate to maintain the required sodium t empera tu re levels .

The details of the heat exchanger design have not yet been established, but will be repor ted, together with heat exchanger per formance , in a future revis ion to Design Data Sheet R-5001.

J - 6

Design Data Sheet No. R-5008

Recommendations

On the bas i s of the discuss ion in this repor t , the following recommenda ' tions a r e made:

1) Flowing helium should be used as a secondary coolant in the heat exchanger port ion with the heat exchanger and helium piping a r r anged to maximize heat re ject ion from the helium to the reac to r water .

2) A fluted annulus containing stagnant helium should be used in the in -core port ion of the loop.

Unless otherwise directed, further work on the var iable level con­cept will be discontinued. The performance of the flowing helium and fluted annulus concepts -will be determined and will be repor ted in future design data shee ts .

J - 7

Design Data Sheet No. R

REFERENCES

1 - Reactor Handbook, Volume I - Mater ia l s , I960, In ters ta te Pub l i shers , Inc., New York.

2 - Conceptual Design Of An In-Pi le Package Loop For Sodium -Cooled The rma l Reactor Fuel Testing, APDA-145, AEC Contract No. AT(11 - 1)-865, September 15, 1961, Atomic Pow^er Development Assoc ia tes , Inc.

3 - Liquid Metals Handbook, Sodium - NaK Supplement, July 1, 1955, Atomic Energy Commission, Dept. of the Navy, Wash., D. C.

4 - Liquid Metals Handbook, June 1952, Atomic Energy Commission, Dept. of the Navy, Wash., D. C.

5 - Organic Coolant Databook, Technical Publication No. A T - 1 , July 1958, Monsanto Chemical Company, St. Louis 24, Missour i .

J - 8

:l

r [P

6 z CD O

I -J liJ<

111 l-

:2

I ^

^ lA

U i

V A fo

rl

n

U U

) m

I 3 (II I i C i z o i c 8!

APPENDIX K

System Per fo rmance Charac t e r i s t i c s

Design Data Sheet R-5009 59-3075-30 Contract No. AT(10-1)-1075 August 21, 1962 H. Honig/D. F . Judd

THE BABCOCK AND WILCOX COMPANY FOR EBASCO SERVICES INCORPORATED

DESIGN DATA ADVANCED TEST REACTOR

THERMAL FLUX LIQUID METAL PACKAGE LOOP SYSTEM PERFORMANCE CHARACTERISTICS

Purpose and Scope

To presen t a tabulation of the loop performance cha rac te r i s t i c s for operating conditions closely approximating the conditions specified in the Design Objectives.

General Description

The mate r ia l presented in this Design Data Sheet is a summary of information calculated on the performance of each of the loop sys tems and components in the nominal 50 mw eas t outer flux t r ap .

The case considered was the reference fuel specimen together with a maximum axial sodium tempera ture difference of 350 F and a max i ­mum sodium tempera ture of 1400 F .

Tabulation

In-Pi le Loop Design and Dimensions

Basic Mater ia ls of Construction

P r i m a r y Coolant

Max. P r i m a r y Coolant Tempera ture , F

Cover Gas Operating P r e s s u r e , psig

Secondary Coolant

Cover Gas

Test Specimen

Number of Pins

Test Specimen Length, in.

Pin Outside Diameter , in.

Fuel Outside Diameter , in.

K-1

Reference Specimen

SKC-2957E SKC-2933E

Stainless Steel, type 316

Sodium

1400

60

Helium

Helium

7

48

0.560

0.500

Tabulation (Cont'd)

Test Specimen

Cladding Material

Bond Material

Cladding Thickness, in.

Geometry

Spacer Wire Outside Dia., in.

Max. Test Specimen Length, in.

Max. Test Specimen Dia., in.

Sodium Conditions

Pump Flow, gpm @ 1214 F

Pump Total Developed Head, ft.

Specimen Flow, gpm

Leakage Flow (specimen bypass), gpm

Specimen P r e s s u r e Drop, ft.

Specimen Inlet Temp. (Mixed Mean), F

Specimen Outlet Temp. (Mixed Mean), F

Heat Exchanger Inlet Temp., F

Heat Exchanger Outlet Temp., F

P r i m a r y Coolant Axial Temp. Difference, F

Helium Conditions

Design Data Sheet R-5009^ August 21, 1962

Reference Specimen

Stainless Steel

Sodium

0.020

Trangular Pi tch

0.090

48

2.425

140

296

135

5

245

1050

1400

1400

1079

350

Flow, Ib/hr 12,000

Loop Inlet Temp, (at top of extension tube), F 140

Heat Exchanger Inlet Temp., F 140

Heat Exchanger Outlet Temp., F 501

Loop Outlet Temp, (at top of extension tube), F 350

Loop Inlet P r e s s u r e (at top of extension tube), psia 600

Loop Outlet P r e s s u r e (at top of extension tube), psia 500

K-2

Design Data Sheet R-5009 August 21, 1962

Tabulation (Cont'd)

Reference Specimen

Heat Balance

Heat Load

Specimen F iss ion Power , kw 1500

Tes t Section Gamma Heating (including Specimen, Sodium and St ructure) , kw 208,7

Gamma Heating F r o m Heat Exchanger

Outlet to Test Specimen Inlet, kw 228.6

Heat Input F r o m Sodium Pump, kw 12.0

1949.3

Heat Loss

In-Core Section to Reactor Water , kw 376.0

Heat Exchanger to Flowing Helium, kw 1573.3*

1949.3

* 659.2 kw loss to r eac to r water from hel ium re tu rn tubes.

Discussion

The performance cha rac t e r i s t i c s for the Thermal Flux Liquid Metal Package Loop cor respond to the nominal design conditions. The Title I effort will extend these cha r ac t e r i s t i c s to include the entire operating range of the per formance va r i ab le s .

K - 3

K

.1

u [

o

N

N

1

APPENDIX L

System Hydraulic Cha rac t e r i s t i c s

Design Data Sheet R-5010 59-3075-10 Contract No. AT(10-1)-1075 August 26, 1962 R. V. S t raub/D. F . Judd

THE BABCOCK AND WILCOX COMPANY FOR EBASCO SERVICES INCORPORATED

DESIGN DATA ADVANCED TEST REACTOR

THERMAL FLUX LIQUID METAL PACKAGE LOOP SYSTEM HYDRAULIC CHARACTERISTICS

Purpose and Scope

The purpose of this Design Data Sheet is to p resen t the liquid metal hydraulic cha r ac t e r i s t i c s of the Thermal Flux Liquid Metal Package Loop with the reference specimen.

General Description

The configuration used for the analysis is shown on the following drawings:

Thermal Flux Liquid Metal Package Loop Figure 2 in Body of Repor t

Thermal Flux Liquid Metal Package Loop,

Fuel Test Specimen SKC-2972E

The reference tes t specimen used in the analysis is as follows:

Specimen Length, in. 48

Number of Fuel Pins 7

Fuel Pin Outside Diameter , in. 0.560

Fuel Pin A r r a y Triangular

Wirewound Spacer Outside Diameter , in. 0.090

Spacer Wire Pi tch, in. 6

The salient feature of the shroud design is the 10 mil l radia l c l e a r ­ance between the shroud r ings and the 2.425-inch bore of the internal flow baffle. This c learance was judged to be sufficient for re inser t ing an i r r ad ia ted specimen. This subject will be reviewed in detail in Title I.

The total specimen p r e s s u r e drop cons is t s of fuel pin p r e s s u r e drop ^ n d drag losses due to the spiral wound space r s . The rod p r e s s u r e drop is calculated with the Darcy-Weisbach equation and Moody's Fr ic t ion Fac to r .

L - 1

Design Data Sheet R - ^ 1 0 August 26, 1962 ^

The spacer head loss was calculated using the method repor ted in NUCLEONICS, June 1961.*

Discussion

The specimen and shroud design is cons idered to be p re l iminary and should be establ ished in Title I. Considerat ion should be given to the hot cel l operat ions that must be performed on it as well as normal ope ra ­t ions. The specimen support s t ruc tu re , specimen instrumentat ion, r a d i a ­tion damage coolant leakage, e tc . , mus t all be cons idered in establishing the final design.

The at tached curve . Figure 1, shows the p r e s s u r e drop for the loop with the reference specimen. Pump cha rac t e r i s t i c s a re not available at t ime of i ssuance of this Design Data Sheet.

* "Drag Coefficients for Fuel E lement S p a c e r s , " A. N. de Stordeur , Beige Nuc lea i re , S. A., B r u s s e l s , Belgium, June 1962, "Nucleonics , " Page 74.

L - 2

^

pi m

^

"w-

l•4-"^

M iJT;

s? 1 •1T

±

-1 J

t-

a o

It

—

I UJ

f^.L^i T:

O

SHROUD R l M G -HiX^GOMKL SHROUD

NA FLOW SAFFLt,

0 9 0 Dl<K. wrRC.

SHROUD RIN&

SCALE:-2KS»Zt

I 3 | 4 5 * 7 • f l l O

I I 13 11 14 i 13 l« 17 !• 19

MATERIAL LIST

FUE-L P I N S

• I W I t l O B t

THERMAL FLUX

LIQUID METAL PACKAGE LOOP

FUtL TEST SPECIMEN

THE BABCOCK ft WILCOX CO. ATOMC ENCMY DIVISKW

LVIKMUK. VA.

EBASCO SERVICES INCORPORATED

ARCHrrECT - EWGWeDI

U.S. AT6MIC E N ^ Y COMMrSSION IDAHO OPERATIONS OFFICE

IDAHO FALLS. IDAHO

JKI. ? 0 7 7 E |iO

APPENDIX M

Functional Requirements for Auxil iary Services

Design Data Sheet No. 5011 59-3075-30 Contract No. AT(10-1)-1075 August 31, 1962 H. Honig

THE BABCOCK AND WILCOX COMPANY FOR EBASCO SERVICES INCORPORATED

DESIGN DATA ADVANCED TEST REACTOR

THERMAL FLUX LIQUID METAL PACKAGE LOOP FUNCTIONAL REQUIREMENTS FOR AUXILIARY SERVICES

Purpose and Scope

To define some of the functional requ i rements external to the in-

reac to r portion of the the rma l flux loop.

Requirements

All requ i rements a r e based on the re ference tes t specimen operating

range .

Helium Coolant (Ref; Design Data Sheet No. R-5001, Rev. 1)

Loop Inlet P r e s s u r e 600 psia Loop Inlet Tempera tu re 140 F Flow Range 0 to 12,000 Ib /h r Maxiinum Tempera tu re out of the

reac tor vesse l 350 F Maximum p r e s s u r e drop, extension

tube inlet to outlet 100 psi Pu r i t y -Commerc i a l (nominal purity

99.9976%)

Helium Cover Gas - Atmospher ic helium

Sodium - Oxygen content not to exceed 50 ppm. Other impuri ty l imitat ions to be d e t e r m i n e d by the exper imente r .

E lec t r ica l Requirements for pump motor to be issued at a l a te r date .

Instrumentat ion Requirements

The requi rements for the control sys tem a r e given in Design Data

Sheet No. R-5005.

M-1

APPENDIX N

Thermal Loop Shielding Analysis

APPENDIX N

THERMAL LOOP SHIELDING ANALYSIS

The following c r i t e r i a required that an analysis be made of the

radiation levels emanating from the thermal flux loop:

I. Pe r sonne l working around the reac tor should not be subjected

to excess ive doses of radiat ion. The following conditions a r e

of concern in this study.

A. Loop removal from reac tor

1. Pe r sonne l atop the reac to r may work through open

refueling ports when the loop is in the in -core position.

2. Pe r sonne l atop the reac tor work to make and break loop

connections. Loop is r a i sed to i ts uppermost position.

Refueling ports a r e closed.

3. Pe r sonne l atop the reac to r , loop being moved horizon­

tally to the drop tube.

4. Loop being handled in canal .

B. Pe r sonne l atop the reac tor work through open refueling

ports to change capsule exper iments . Loop is in its in-

core posit ion. Reactor water level is at elevation 92-93

feet.

C. Access to capsule t rench, water level in reac to r is at eleva­

tion 92-93 feet.

I I . Radiation damage to

A. Loop motors

B. Top head seals of adjacent loops

I I I . In both I and I I above, considerat ion mus t be given to both

unruptured and ruptured tes t specimens in the thermal loop.

At the t ime this r epor t was p repared , the analysis had not been com­

pleted. Considerable redes ign and reana lys i s must be done during Title

I in o rde r to a r r i v e at a loop design that will satisfy all these c r i t e r i a .

N - 1

Descr ibed here in a r e the problems encountered in attempting to mee t the

various design c r i t e r i a and some possible solutions to these problems . ^ B

The principal sources of after-shutdown radiation in the thermal flux

liquid meta l package loop a r e gamma rays from fission products result ing

from fuel sample defects and induced activit ies in the loop s t ruc tura l

ma te r i a l s and p r imary coolant. Calculations show that the p r imary

coolant sodium activation is the major contributor to the after-shutdown

dose ra tes when the fuel pin cladding is not ruptured. F iss ion product

activity and s t ruc tu ra l ma te r i a l s activation a r e negligible in comparison

to the coolant activation. For the ruptured fuel pin case , p r imary coolant

activation is sti l l the l a rges t contributor to the dose r a t e s , with the

fission products dis t r ibuted in the coolant contributing about 20%-30% to

the dose r a t e .

The sodium sump is the uppermost location of sodium when the loop

is in its normal operating (vertical) position. Therefore , this sump is the

source of controlling radiations in all cases since the p r imary coolant or

distr ibuted fission products a r e the major contr ibutors to the after-shut­

down dose r a t e s . This analysis i s , therefore , chiefly concerned with general

sump a r rangement s including sump shielding, result ing dose ra tes and

heat generation due to sump sou rce s . All calculations a r e based on an

i r rad ia t ion t ime of 10,000 hours .

I - A - 1 . This condition was not analyzed.

I -A-2 . When the loop is in its uppermost position, the top of the sodium sump is shielded at the side by 4 ft. 3 inches of reactor water and by the 8-inchthick top shield plate . A shield plug above the sump at tenuates d i rec t s t reaming ver t ical ly upward from the sump.

The calculations were based on equilibrium sodium-24 activity in the

sodium sump. This sodium activation occurs at 1225 F with a sodium

density of 0.809 g m / c c . The fission product calculations for the ruptured

fuel pin case were based on distr ibution of fission products in the sodium

with an assumed concentration of 100% of gases , 10% of hal ides , and 1% of

solids in the sodium sump.

N - 2

The calculation of the equil ibrium sodium activity was made using the equation:

s act th

where N = pN I . / ^ • •, .. r •> ^ s o e a t o m s / c c = atomic density of coolant

A s

p = coolant density

NQ= Avagadro 's number

I = atomic abundance of parent isotope

A = atomic weight of element

o act ~ activation c ro s s section for sodium cor rec ted for t empera tu re and Maxwellian distr ibution

6th = activating the rmal flux

X = decay constant of sodium-24 (T. ,,, = 15.1 hrs)

t, = i r radia t ion t ime

t , = decay t ime

Equil ibr ium sodium-24 activity in the sump was found to be 6.1 x 10

d i s / c c - s e c at one hour after shutdown.

F iss ion product source s t rengths were computed with the use of

digital computer p rog rams which solve coupled t ime-dependent differ­

ential equations descr ibing fission product buildup and decay in a reac tor

c o r e . The p rog rams compute the concentration in a reac to r core of each

isotope in each of a number of specified fission product chains , and group

the act ivi t ies by energy. The p rograms a r e l imited to a maximum of five

isotopes in a chain, six fission product buildup t imes , and six decay t imes

after shutdown.

The attenuation calculations through water above the sump were made

by assuming a s e r i e s of ver t ica l line sources to r ep resen t the annular

sodium sump. A gamma yield of one photon per disintegrat ion of 2.76

Mev and of 1.38 Mev energy from SDdium-24 decay was considered. The

fission product act ivi t ies were grouped by energy on the computer p rogram

in six groups of average energies 0.35 Mev, 0.90 Mev, 1.40 Mev, 1.80 Mev,

2.30 Mev, and 2.60 Mev. The line source equation for the dose ra te used

w a s :

N - 3

B K S^ DR ^ ^ T F (0^, b) - F (02) b) I

4 na

where B = Dose buildup through water

K = Conversion factor; flux to dose ra te at specific energy

S^ = Source s t rength of l ine - source

It should be noted that self-attenuation of the source .ma te r i a l is not

allowed for in this equation. Dose ra tes from activated sodium in the

heat exchanger region and in the tes t section were found to be lower than

those from the sump by factors of approximately 100 and 2,000, r e s p e c ­

tively.

The initial after-shutdown radiation calculations were performed

assuming the loop in a ra i sed position with no allowance for the 8-inch

thick s tee l plate in the vesse l head and were calculated for one hour after

r eac to r shutdown. These calculations indicated unreasonably high dose

ra tes at the water surface of 1.94 x 10 m r e m / h r with unruptured fuel

pins and 3.03 x 10 m r e m / h r with ruptured fuel pins. These numbers ,

when reduced by an est imated attenuation factor of 200 for the 8-inch 4

s teel plate , resul ted in dose r a t e s of approximately 9.7 x 10 m r e m / h r 5

and 1.5 x 10 m r e m / h r , respect ively, for the unruptured and ruptured

fuel pin c a s e s .

All of the above calculations were made with an original sump con­

figuration in which the top of the sump was located 5 '9" above the top of

the tes t specimen and the sump was assumed to be 1 foot long with an out­

side d iameter of 5 inches and an inside d iameter of 1-5/8 inches . There

was no shielding in or around the sump. Because of the high dose ra tes

found for this loop position with these assumpt ions , no calculations were

made for other loop positions with these same assumpt ions .

The next analysis considered 3 days shutdown t ime and all s teel

shielding in place atop the r eac to r . This longer shutdown t ime is much

n e a r e r the actual shutdown t ime at which loop handling operations might

occur , and therefore all subsequent calculations for loop handling used

this assumpt ion. Another assumption for these la ter calculations was

that the equil ibrium sodium-24 activity in the sump is 12.4% of the equil

r ium sodium-24 activity in the loop. This was based on diffusion

N - 4

calculations referenced in P r a t t & Whitney Aircraf t r epor t TIM No. 558.

Using these l a t t e r assumpt ions , and further assuming that a tungsten

shield plug above the sodium sump effectively at tenuates the d i rec t loop

s t reaming, the dose ra te at the top of the 8-inch steel plate where handling

operations occur was calculated to be approximately 30 m r e m / h r for the

unruptured fuel pin case and approximately 37 m r e m / h r for the ruptured

fuel pin ca se .

All of these la ter calculations were made by assuming self-attenuating

line sources to r ep re sen t the sump annulus. This line source equation is

as follows:

(M' C ) , ,. DR - B K S^ ,, „ r L M ^ C ,

—7L e s - b , 4 Tta 1

where B = Dose buildup for iron applied for ent i re b of water and s teel

K = Conversion factor; flux to dose ra te at specific energy

Dose ra tes from the hea ter section, heat exchanger section and test section

were found to contribute approximately half of the quoted dose r a t e s .

I -A-3 . When the loop is lowered two feet from its uppermost position (see I-A-2) for horizontal t r anspor t to the drop tube, there is about one foot of water over the top of the loop. With the same sump configuration (12" high, 5" O.D., 1-5/8" I.D., 3 -1 /2" thick tungsten plug over the sump) and with refueling por t s open, the dose r a t e s through the po r t s a r e 367 m r e m / h r for the unruptured fuel pin case and 542 m r e m / h r for the ruptured pin case . If a shielding viewing window equivalent to 4" of s teel is placed in the open port , these dose r a t e s drop to 23 m r e m / h r and 29 m r e m / h r .

After analyzing cases I-A-2 and I-A-3 with 12.4% of equil ibrium

sodium-24 activity, and after studying expansion volumes, t rans ien t t e m ­

p e r a t u r e s , pumping effect of motor shaft, e tc . , it was decided that this

12.4% is too optimist ic for the the rmal loop. The actual number will probably

be somewhere between 12.4% and 100% activity. New calculations were made

using 100% activity, giving maximum after-shutdown dose r a t e s .

This m o r e conservat ive assumption resu l ted in the following dose r a t e s

for the ruptured fuel pin cases descr ibed above.

N - 5

E q u i l i b r i u m N a - 2 4 Ac t iv i ty in S u m p

100% 12.41

C a s e I - A - 2 N a - 2 4 145 m r e m / h r F i s s i o n P r o d u c t s 7 m r e m / h r

152 m r e m / h r 37 m r e m / h r

C a s e I - A - 3 N a - 2 4 1,632 m r e m / h r (open p o r t s ) F i s s i o n P r o d u c t s 175 m r e m / h r

1,807 m r e m / h r 542 m r e m / h r

C a s e I - A - 3 N a - 2 4 102 m r e m / h r ( sh ie lded p o r t s ) F i s s i o n P r o d u c t s 6 m r e m / h r

108 m r e m / h r 29 m r e m / h r

D o s e r a t e s f r o m t h e h e a t e r s e c t i o n , h e a t e x c h a n g e r s e c t i o n , and t e s t s e c t i o n

c o n t r i b u t e a p p r o x i m a t e l y 10%, 11% and 12%, r e s p e c t i v e l y to the above c a s e s .

I - A - 4 Cana l hand l ing w a s not a n a l y z e d .

I - B V e s s e l E n t r y wi th L o w e r e d W a t e r L e v e l

With the loop in i t s i n - c o r e pos i t i on , the s o d i u m s u m p s i t s wel l above

the c o r e and wi l l c o n t r i b u t e to the d o s e r a t e a b o v e the w a t e r if the w a t e r

l e v e l i s l o w e r e d to p e r m i t e n t r y for loop weld ing o p e r a t i o n s , e t c . , i n s i d e

the v e s s e l . With the w a t e r l e v e l 10 fee t o v e r the top of c o r e , d o s e r a t e s

a t the w a t e r s u r f a c e f r o m N a - 2 4 in the s u m p a r e a s f o l l o w s :

1 Day Shutdown 320 R / h r 3 Days Shutdown 35 R / h r

With 11 fee t of w a t e r o v e r the c o r e , the n u m b e r s a r e :

1 Day Shutdown 70 R / h r

3 D a y s Shutdown 8 R / h r

With r u p t u r e d fuel p ins and the s a m e d i s t r i b u t i o n of f i s s i o n p r o d u c t s

a s a s s u m e d e a r l i e r l o c a t e d in the s u m p , the d o s e r a t e wi th 10 fee t of w a t e r

o v e r the c o r e b e c o m e s 45 R / h r a f t e r 3 days shu tdown.

B e c a u s e of t h e s e h igh d o s e r a t e s , the t h e r m a l loop should be r e m o v e d

f r o m the r e a c t o r d u r i n g any o p e r a t i o n s tha t r e q u i r e r e a c t o r v e s s e l e n t r y .

I -C C a p s u l e E x p e r i m e n t R e m o v a l - L o w e r e d W a t e r L e v e l

R e m o v a l of c a p s u l e e x p e r i m e n t s r e q u i r e s t ha t the r e a c t o r w a t e r l e v e l

be l o w e r e d to a p p r o x i m a t e l y 10' to 1 1 ' o v e r the top of the c o r e . Dur ing

th i s o p e r a t i o n p e r s o n n e l m a y be r e q u i r e d above the open r e a c t o r h e a d

a c c e s s p o r t s and a c c e s s in to the c a p s u l e a c c e s s t r e n c h wi l l be r e q u i r e d . '

N - 6

Therefore , dose r a t e s at both locations were checked to de te rmine allow-

ible access t i m e s .

After the work descr ibed in previous sections was done, m o r e detailed

analyses were made of the the rmal conditions in the loop. These indicated

that a 12- inchdeep sump would not accomodate the t he rma l expansion of

the sodium under al l operating conditions. Therefore , a new sump config­

urat ion was es tabl ished: 18 inches deep, sodium O.D, of 4 -1 /2 inches ,

sodium I.D, of 1.8 inches (the pump shaft and a baffle run down through the

center of the sump), and a 3-1 /2- inch thick tungsten shield between sump

and moto r . Equil ibrium Na-24 activity was assumed in the sump. Shielding

credi t was taken for the loop wal ls , r eac to r coolant wate r , and vesse l wal ls ,

as applicable.

The result ing dose ra tes a r e as follows at 1 hour after shutdown:

Capsule Access Trench Reactor Head Access Po r t

Na-24 452 m r e m / h r Na-24 11.4 r e m / h r F i s s ion F i s s ion

Products 61 m r e m / h r Products 2.1 r e m / h r

513 m r e m / h r 13.5 r e m / h r

It was apparent that dose ra tes from a ba re (unshielded) sump would

be too high. In studying the feasibili ty of shielding the sump, the use of

external shielding was considered but not studied, s ince it was felt that

in terna l shielding might be sa t i s fac tory . However, with in ternal shielding

taking up some of the radia l space in the sump, the depth of sodium mus t

necessa r i ly i nc rease to accommodate the requi red expansion volume.

With 1/2-inch thick tungsten shielding immediate ly inside the loop wall

around the sump, the sump configuration i s :

Shield O.D. 4 -1 /2 inches

Shield I.D. 3-1/2 inches

Sodium O.D. 3-1/2 inches

Sodium I.D. 1.8 inches

With this radial configuration, the sodium extends up into the 8-inch

d iamete r portion of the loop. The dose r a t e s a r e :

N-7

Capsule Access Trench Reactor Head Access Por t

Na-24 5 r e m / h r Na-24 9 r e m / h r F iss ion Fiss ion

Products 1 r e m / h r Products __2 r e m / h r

6 r e m / h r 11 r e m / h r

These calculat ions show that adding the tungsten shielding inside the

sump d e c r e a s e s the r eac to r access por t dose ra te about 20% but i nc r ea se s

the capsule access t rench dose ra te by a factor of approximately 12. The

reduction in ra te at the top of the reac tor is real ized because the slant

height of the shielding and, therefore , i ts effective thickness is g rea t in

compar ison to the r i s e in the level of the sodium source . The inc rease in

ra te at the acces s t rench is due to the r i se in sodium level which causes

the radiat ion to pass through the shielding at a reduced slant height angle,

thus reducing the effective thickness of the shielding.

It is apparent that a g rea t deal of study remains to be done to overcome

these radiat ion p rob lems .

11-A Motor Winding Protec t ion

It has been es t imated that insulation in the motor winding should be 9

l imited to an integrated dose of 10 Rads. Therefore , a tungsten shield plug

was placed over the sump to pro tec t the windings. It was found that 3.5" of

tungsten between the sodium sump and motor windings pe rmi t s operation for 9

grea te r than 10,000 hours without exceeding the 10 Rad dose. Additional

shielding was placed above the motor to prevent gamma s t reaming up the

loop for personnel protect ion during handling.

Since absorption of gamma rays produces heat in shield m a t e r i a l s , heat

generation r a t e s in the tungsten shield plug were computed. Original calcu­

lations were quite conservat ive and indicated heat removal from the plugs

may presen t a problem. Therefore , a different shield configuration was

t r ied . This configuration was a laminated tungsten shield located in the sump

The tungsten would be 1/4" thick slabs with sodium above and below these

slabs with the las t slab above al l the sodium. This configuration proved

unfeasible when it "was found that the motor windings would receive the allow-9

able dose of 10 Rads in approximately 400 hours-

N - 8

Because of the high dose r a t e s with the laminated shield, m o r e detailed

| Jk ;u la t ions of the heat generat ion in the solid tungsten shield plug and motor

snaft were pe r fo rmed . These detailed calculat ions gave heat generat ion

r a t e s considerably lower than the original conservat ive calcula t ions . An

analysis was s ta r ted to de te rmine if the heat generated in the solid tungsten

shield plug and motor shaft is removed without producing excess ive t e m p e r a ­

tu re s in these pa r t s and in the motor shaft bearing which should probably be

maintained below about 250 F . Sxifficient t ime was not available to complete

a detailed ana lys i s . However , p re l iminary calculations indicate that the heat

is removed without ra is ing the t empera tu re above allowable l imi t s . The

calculat ions a lso indicated that heat in the tungsten shield plug and pump

shaft due to t he rma l radiat ion from high t empera tu re sodium in the sump

could be higher (depending on sodium tempera tu re ) than the heat due to the

absorption of gamma r a y s . This problem could possibly be solved by (1)

using the rma l radiat ion heat shields between the sodium and the pump shaft

and the tungsten or (2) devising a means to minimize ci rculat ion between the

sodium in the sump and sodium in the loop thereby lowering the t empera tu re

of the sodium in the sump. These problems will have to be analyzed in detai l

in Title I. F rom a shielding viewpoint the or iginal 3.5" tungsten plug proved

feasible and the bet ter a r r angemen t .

A further considerat ion of motor damage concerns the effects of the

radiat ion from the t he rma l loop sump upon the adjacent fast loop motor . If

the original height of the PW-19 loop is maintained, the fast loop motor will

be at very near ly the same elevation as the the rma l loop sump. It is expected,

but has not been calculated, that the fast loop motor will be damaged fair ly

quickly. If the fast loop is lengthened (see IDO-24041, Supplement 1), i t s

motor will be somewhat above the the rma l loop sump and, the re fore , the

radiat ion damage the fast loop motor incurs will be l e s s . Fu r the r study of

this problem is requi red during Title I work.

I I -B

Damage to seals of adjacent loops and other m a t e r i a l s seems unlikely

from loop handling opera t ions . For example, if the loop is r a i s ed at 3 days

shutdown, dose ra te over 4 ' 3 " H^O a t vesse l head is 35 R /h r from equilibrium

Na-24 activi ty in the sump.

N - 9

If init ial damage to the adjacent seals occurs at 2 x 10 Rad, this would 6 4

allow 2 x 1 0 = 5.7 x 10 hours from Na-24 in loop handling (loop in ra ised^^ ~35 W

position) p r io r to initial damage. Summary

It appears that shielding is requi red in or around the sump to protect

the motor windings and bear ings from radiat ion damage and to protect

operating personnel from overexposure during loop handling and other

reac tor opera t ions . At this t ime no firm shielding a r rangement s can be

made; further design work mus t be pursued.

N . 1 0 * U.S. GOVERNMENT PRINTING OFFICE ; 1963 O—695-802