advanced nuclear fuels corporation - nrc.gov · *values at 200 seconds into loca transient. 0ii. 6...

ADVANCED NUCLEAR FUELS CORPORATION

ANF-87-159 Supplement 1

Issue Date: 4/28/88

H. B. ROBINSON UNIT 2

LARGE BREAK LOCA/ECCS ANALYSIS

WITH AN INCREASED ENTHALPY RISE FACTOR

Prepared by

R. C. Gottula, Team Leader PWR Safety Analysis

Licensing & Safety Engineering Fuel Engineering & Technical Services

April 1988

Q(II

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PLEASE READ CAREFULLY

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ANF-3145.472A (12/87)

IANF-87-159 Supplement 1

TABLE OF CONTENTS

Section Page

1.0 INTRODUCTION...... ............ . . . . . 1

2.0 SUMMARY OF RESULTS. . ................ . . . 2

3.0 LOCA ANALYSIS MODEL AND ASSUMPTIONS ...... ...... ... 5

4.0 RESULTS .............. .............. 10

5.0 CONCLUSIONS. . ............. . . . . . .... . .. 29

6.0 REFERENCES .............. ........ ....... 30

ii ANF-87-159 Supplement 1

LIST OF TABLES

Table PAst

2.1 H. B. Robinson Unit 2 LOCA/ECCS Analysis Results...... . . 3

3.1 H.B. Robinson Unit 2 System Data.............. . . . 7

3.2 Fuel Design Parameters. ................ . . . . . 8

4.1 H.B. Robinson Unit 2 LOCA/ECCS Analysis Event Times . . . . . . 11

4.2 H. B. Robinson Unit 2 LOCA/ECCS Analysis Conditions and Results............. . . . .... . . 12

iii ANF-87-159 Supplement 1

LIST OF FIGURES

Figiure Pq

2.1 H.B. Robinson Unit 2'K(Z) - Normalized Axially De endent Power Peaking Limit (FQ(Z)) Curve, FQY - 2.32 . . . . . . . . . . . . . . . . . . . . . . . . . . . 4

3.1 H.B. Robinson Unit 2 FO(Z) vs. Core Height Used in the Large Break LOCA/ECCS Analysis . . . . . . . . . . . . . . . 9

4.1 Accumulator (Intact) Flow Rate during Refill and Reflood Periods, 0.8 DECLG Break . . . . . . . . . . . . . . . . 13

4.2 Accumulator (Broken) Flow Rate during Refill and Reflood Periods, 0.8 DECLG Break . . . . . . . . . . . . . . . . 14

4.3 HPSI (Intact) Flow Rate during Refill and Reflood Periods, 0.8 DECLG Break . . . . . . . . . . . . . . . . 15

4.4 HPSI (Broken) Flow Rate during Refill and Reflood Periods, 0.8 DECLG Break . . . . . . . . . . . . . . . . 16

4.5 LPSI (Intact) Flow Rate during Refill and Reflood Periods, 0.8, DECLG Break . . . . . . . . . . . . . . . . 17

4.6 LPSI (Broken) Flow Rate during Refill and Reflood Periods, 0.8 DECLG Break . . . . . . . . . . . . . . . . 18

4.7 Reflood Core Mixture Level, 0.8 DECLG Break, Cosine Power Shape . . . . . . . . . . . . . . . . . . . . . . . 19

4.8 Reflood Downcomer Mixture Level, 0.8 DECLG Break, Cosine Power Shape...... ..... ..... .. . . . .. 20

4.9 Reflood Upper Plenum Pressure, 0.8 DECLG Break, Cosine Power Shape. .............. . . . . . . . .. 21

4.10 Core Flooding Rate, 0.8 DECLG Break, Cosine Power Shape . ..... ..... ..... ..... .. 22

4.11 Reflood Core Mixture Level, 0.8 DECLG Break, Top-Skewed Power Shape . . .... ..... ..... ..... 23

4.12 Reflood Downcomer Mixture Level, 0.8 DECLG Break, Top-Skewed Power Shape ........ ..... .... 24

iv ANF-87-159 Supplement 1

LIST OF FIGURES

4.13 Reflood Upper Plenum Pressure, 0.8 DECLG Break, Top-Skewed Power Shape......... . . . . . . .... 25

4.14 Core Flooding Rate, 0.8 DECLG Break, Top-Skewed Power Shape. .............. . . . . .. 26

4.15 Cladding 'Temperature during Refill and Reflood Periods, 0.8 DECLG Break, Cosine Power Shape..... . . . .. 27

4.16 Cladding Temperature during Refill and Reflood Periods, 0.8 DECLG Break, Top-Skewed Power Shape . ..... .. 28

1ANF-87-159 Supplement 1

1.0 INTRODUCTION

This document presents the results of a postulated large break loss-of-coolant

accident (LOCA) analysis for the H. B. Robinson Unit 2 reactor with only one High Pressure Safety Injection (HPSI) pump available. The large break LOCA

analysis reported in Reference 1 assumed availability of 2 HPSI pumps. It has

been determined that for a condition of loss of offsite power, which is assumed for a LOCA transient and where startup of diesel generators is

necessary to operate HPSI pumps, it is possible that only one HPSI pump may be

available. The refill and reflood calculations for the limiting double-ended

cold leg guillotine break (0.8 DECLG) identified in Reference 1 were repeated

assuming only one HPSI is available. Peak cladding temperatures were

calculated for both a cosine and a top-skewed axial power shape. The analysis

supports a nuclear enthalpy rise factor (FAH) of 1.7, a total peaking factor

(Fq) of 2.32, and the existing axially dependent power peaking limit provided

in Amendment 109 to the H. B. Robinson Unit 2 Technical Specifications. The

analysis was performed at a power level of 2346 MWt (2300 MWt plus 2%

uncertainty) with ANF fuel and a steam generator tube plugging level of 6%. All plant operating conditions were the same as given in Reference 1 except

for the HPSI flow rate.

SII

2 ANF-87-159 Supplement 1

2.0 SUMMARY OF RESULTS

The results of the analysis demonstrate that the requirements of 10 CFR

50.46(b) are satisfied for H. B. Robinson Unit 2 with the axially dependent power peaking limit curve (K(z)) shown in Figure 2.1. The analysis supports a

maximum total power peaking factor (Fq) of 2.32 and a nuclear enthalpy rise factor (FAH) of 1.70.

The analysis was performed for the previously identified double-ended cold leg

guillotine break with a discharge coefficient of 0.8 (0.8 DECLG).

Calculations were performed for both a center-peaked chopped cosine axial power shape and an end-of-cycle (EOC) top-skewed axial power shape. Both

cases utilized the maximum stored energy predicted to occur at an exposure of 1.8 MWd/kgU. The peak cladding temperature (PCT) for the chopped cosine axial

power shape was predicted to be 1926'F. The PCT for the top-skewed axial power shape was predicted to be 1982*F. The results of the analysis are summarized in Table 2.1.

Sll


Table 2.1 H. B. Robinson Unit 2 LOCA/ECCS Analysis Results

Peak Peak X/L=0.50 X/L=0.81

Peak Rod Avg. Exposure Range, MWd/kgU 0-49 0-49

Peak Cladding Temperature, PCT, *F 1926 1982

Total Core Zr-H 20 Reaction, % <1.0 <1.0

Local Zr-H 20 Reaction, %* 2.53 2.50

*Values at 200 seconds into LOCA transient.

0II


3. An additional 5 second delay time from 25 seconds to 30 seconds was

considered for the Low Pressure Safety Injection (LPSI) start time.

Since LPSI flow doesn't begin until about 50 seconds into the transient,

this change has no impact on the analysis.

4. An additional 5 second delay was not added to the start of containment

sprays because the original start time would result in a lower and more

conservative containment pressure. Also, the additional delay time would

have a minimal effect on the containment pressure.


Table 3.1 H.B. Robinson Unit 2 System Data

Primary Heat Output, MWt 2300 (2346)*

Primary Coolant Flow, lbm/hr 100.3 x 106

Primary Coolant Volume, ft3 9186**

Operating Pressure, psia 2250

Inlet Coolant Temperature, OF 546.2.

Reactor Vessel Volume, ft3 3684

Pressurizer Total Volume, ft3 1300

Pressurizer Liquid Volume, ft3 780

Accumulator Total Volume, ft3 (each of three) 1200

Accumulator Liquid Volume, ft3 825

Accumulator Pressure, psia 615

Steam Generator Heat Transfer Area, ft2 (one) 40859***

Steam Generator Secondary Flow, lbm/hr (one) 3.37 x 106 (3.428 X 106)*

Steam Generator Secondary Pressure, psia 800

Reactor Coolant Pump Rated Head, ft 266

Reactor Coolant Pump Rated Speed, rpm 1190

Reactor Coolant Pump Rated Torque, ft-1bf 22363

Moment of Inertia, lbm-ft2/rad 70000

*Values within the parenthesis are for 102% power.

**Includes the pressurizer total volume, and 6% SG tube plugging.

***Includes 6% SG tube plugging.

SII


Table 3.2 Fuel Design Parameters

Parameter ANF Fuel

Cladding 0.D., in. 0.424

Cladding I.D., in. 0.364

Cladding Thickness, in. 0.030

Pellet O.D., in 0.3565

Diametral Pellet-to-Clad Gap, in. 0.0075

Pellet Density, % TD 94.0

Active Fuel Length, in. 144

Enriched U02, in. 132

Upper Blanket, in. 6.0

Lower Blanket, in. 6.0

Cell Water/Fuel Ratio 1.76

Rod Pitch, in. 0.563

2.5

1.5 *%

2.0- 14

- - - COSIN

------ EOC AXIAL

0.5

0.0 0 2 4 8 8 10 123

ELEVATION IN CORE (FT) >

TI

8-nc

Figure 3.1 H.B. Robinson Unit 2 FQ(Z) vs. Core Height UsedaI in the Large Break LOCAJECCS Analysis

0.5k


4.0 RESULTS

The reduced HPSI flow rate resulted in a slightly delayed beginning-of-core

recovery (BOCREC) time from 46.17 to 46.29 seconds. The reduced HPSI flow

rate also resulted in a slightly reduced reflood rate due to a slightly lower

water level in the downcomer. Plots of the accumulator, HPSI, and LPSI flow rates and reflood parameters are shown in Figures 4.1 through 4.14. Event

times for the analysis are shown in Table 4.1.

The peak cladding temperature for the chopped cosine axial power shape was not significantly changed from that reported in Reference 1. The peak

cladding temperature was predicted to be 1926*F, and the elapsed time to its occurrence was reduced from 63.84 seconds( 1 ) to 49.64 seconds. The heat transfer coefficient from the FCTF correlation for the slightly reduced reflood rate caused the PCT to occur at a different node and at a slightly lower value than that reported in Reference 1.

The peak cladding temperature for the limiting top-skewed axial power shape was predicted to be 1982*F. The time of occurrence of peak cladding

temperature increased from 126.74 seconds( 1 ) to 140.2 seconds, and the peak clad temperature predicted was increased relative to that reported in . Reference 1.

Plots of peak cladding temperature during the refill and reflood portions of the transient are shown in Figures 4.15 and 4.16. Temperature, timing, channel blockage fraction, and metal-water reaction values are tabulated in Table 4.2.


Table 4.1 H.B. Robinson Unit 2 LOCA/ECCS Analysis Event Times

Event Time (sec)

Start 0.0

Initiate Break 0.1

Safety Injection Signal 0.7

Accumulator Injection (Broken Loop) 3.2

Accumulator Injection (Intact Loop) 11.9

End-of-Bypass (EOBY) 21.84

Safety Pump Injection, HPSI 25.70

Accumulator Empty (Broken Loop) 43.44

Safety Pump Injection, LPSI (Broken Loop) 43.59

Start of Reflood (BOCREC) 46.29

Accumulator Empty (Intact Loop) 50.58

Safety Pump Injection, LPSI (Intact Loop) 51.30

Peak Clad Temperature Reached:

Cosine 49.64

Top-skewed 140.2

f


Table 4.2 H. B. Robinson Unit 2 LOCA/ECCS Analysis Conditions and Results

Calculational Basis

License Core Power, MWt 2300

Power Used for Analysis, MWt* 2346

Break Size, DECLG, Cd 0.8

Nuclear Enthalpy Rise, FAH 1.70

Steam Generator Tube Plugging, % 6.00

Maximum Peak Rod Average Exposure, MWD/kgU 49.0

Peak Peak X/L=0.50 X/L=0.81

Peak Rod Avg. Exposure Range, MWD/kgU 0-49 0-49

Exposure at Time of Peak Stored Energy, MWD/kgU Peak Rod Average Exposure 1.8 1.8

Average LHGR, kw/ft 5.98 5.98

Peak Linear Heat Gneration Rate (LHGR)* 14.16 13.50

Total Peaking Factor, FQT 2.32 2.21

Axial Peaking Factor, FEZ** 1.365 1.301

Local Peaking Factor, FL 1.07 1.07

Peak Cladding Temperature, PCT, "F 1926 1982

Peak Cladding Temperature Location, ft 6.04 11.0

Peak Cladding Temperature Time, sec 49.64 140.2

Hot Rod Burst Location, ft 6.04 9.75

Hot Rod Burst Time, sec 37.19 50.74

Channel Blockage Fraction 0.268 0.307

Total Core Zr-H20 Reaction, % <1.0 <1.0

Local Zr-H20 Reaction Location, ft 6.04 11.0

Local Zr-H20 Reaction, %*** 2.53 2.50

*Including 1.02 factor for power uncertainty. **Including 1.03 for engineering uncertainty.

***Values at 200 seconds into LOCA transient.

4000

C

S2000

I

-oc 00

EnD

00

Time after EOBY (sec) M

Figure 4.1 Accumulator (Intact) Flow Rate during Refill and Reflood <+t Periods, 0.8 DECLG Break

1500

C-)

o 1000

0 c

0

0 loo 200 30ooS-n

-oo

08

Time after EOBY (sec) M

=3Z

Figure 4.2 Accumulator (Broken) Flow Rate during Refill and Reflood Periods, 0.8 DECLG Break.

Time trEB c

40

S20

4-)

-O

- 0

U-

V)

0

0 100 200 3 0cc2-

CD -i

Time after EOBY (sec) 0 '

Figure 4.3 HPSI (Intact) Flow Rate during Refill and Reflood Periods, 0.8 DECLG Break

I *913 A 8*0 spOL.A~d POOIP~j PuP [1t4ja 6uLtfp alub moI] (ualoig) dH V anL

eE 1Pas) A903 .JaI4P aUJl1

u I- C 001 0

0

-n

0

01 (

0

oz~

17 ANF-87- 159 4 . Supplement 1

0(

00

0 0

00

4

oe N0

(z (A

00

En

(oasqL) 4:)~uI)34e mOL IS

18 ANF-87- 159 Supplement 1 0

00

MC

0

0 0

4

(3)

4- 4-j 4- C )

0 0 0

0- 0

(oas/qL)~U. (ul.g)ae LL S

c4

(1

.C

Wa

.(D

c2

TB w

FCl

C n w S XI

H

C,

U

,a -. j

co 40 s0 120 160 200 240 280 320 360 -400 3n

TIME AFTER BREAK (SEC)

Figure 4.7 Reflood Core Mixture Level, 0.8 DECLG Break, Cosine Power Shape

mI I I I I I I I

Cu

O

Ii

LJo

T A B( w

c:

mi:

z

CDC

v

(D

CD co 40 80 120 160 200 240 280 320 360 400 =L

TIME AFTER BREAK (SEC)

Figure 4.8 Reflood Downcomer Mixture Level, 0.8 DECLG Break, Cosine Power S

a I I ~ I I I IIII In*

m In

CL,

CO a In

cr c'C U,

LEJ

I II II I.

z -'In

cc

a n

(D : 40 80 120 160 200 240 280 320 360 400 TIME AFTER BREAK (SEC)

Figure 4.9 Reflood Upper Plenum Pressure, 0.8 DECLG Break, Cosine Power Shape

mII I I I I I

Cu

cu LL

-j

cc

cc WLIcu

0 z X 0

(D --J

co 40 80 120 160 200 240 280 320 360 400 TIME AFTER BREAK (SEC)

Figure 4.12 Reflood Downcomer Mixture Level, 0.8 DECLG Break, Top-Skewed Power Shape

In I I I I I

In

CO CL

en En

CL

-l

cc

:

C1

Ln

CO

CD .j 40 80 120 160 200 240 280 320 360 400

= 01 TIME AFTER BREAK (SEC)

4.13 Reflood Upper Plenum Pressure, 0.8 DECLG Break, Top-Skewed Power Shape

FG=2.21.FDH=1.70 C

(I I I III III cu

1. PCT NODE

(NODE 23 AT 11.00 FT.) O

LLcu 2. RUPTURED NODE

mn (NODE 18 AT 9.75 FT.) L g LUI Eo

LU CL

a:

Do

ao

I..J

CD z

0

02

(D -r

5.0 20.0 40.0 60.0 80.0 100.0 120.0 140.0 160.0 180.0 200.0 m TIME - SECONDS

Figure 4.16 Cladding Temperature During Refill and Reflood Periods, 0.8 DECLG Break, Top-Skewed Power Shape *


5.0 CONCLUSIONS

For break sizes up to and including the double-ended severance of a reactor

primary coolant pipe, the Emergency Core Cooling System for H. B. Robinson

Unit 2 will meet the Acceptance Criteria as specified in 10 CFR 50.46, with

the 1.70 (FAH) limit and the axially dependent power peaking limit for 2.32

(FqT) shown in Figure 2.1. The criteria are as follows:

(1) The calculated peak fuel element clad temperature does not exceed the

2200'F limit.

(2) The amount of fuel element cladding that reacts chemically with water

or steam does not exceed 1% of the total amount of zircaloy in the

reactor.

(3) The cladding temperature transient is terminated at a time when the core

geometry is still amenable to cooling. The local cladding oxidation

limit of 17% is not exceeded during or after quenching.

(4) The core temperature is reduced and decay heat is removed for an

extended period of time, as required by the long-lived radioactivity remaining in the core.

0II


6.0 REFERENCES

(1) "H.B. Robinson Unit 2 Large Break LOCA/ECCS Analysis with an Increased Enthalpy Rise Factor," ANF-87-159, Advanced Nuclear Fuels Corp., November 1987.

(2) Dennis M. Crutchfield (USNRC Asst. Director Division of PWR Licensing-B) to Gary M. Ward (ENC Manager, Reload Licensing), "Safety Evaluation of Exxon Nuclear Company's Large Break ECCS Evaluation Model EXEM/PWR and Acceptance for Referencing of Related Licensing Topical Reports," dated July 8, 1986.

ANF-87-159 Supplement 1

Issue Date: 4/28/88

H. B. ROBINSON UNIT 2 LARGE BREAK LOCA/ECCS

ANALYSIS WITH AN INCREASED ENTHALPY RISE FACTOR

Distribution

TH Chen

NF Fausz

RC Gottula

JS Holm

LA Nielsen

GL Ritter

BD Stitt

IZ Stone

BD Webb

HE Williamson

CP&L/HG Shaw (12)

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