neacrp 3-d lwr core transient bencrmark

. NEACRP 3-D LWR CORE TRANSIENT BENCRMARK Draft Specifications

Herbert Finnemann ' SIEMENS AGIKWU

Erlaigen, Gqmariy

and

Aldo Galati ENEA, CRE Casaccia

Romc, 1t;ly

Paper to be presented at the 3411 NEACRP Mcciing PSI, Wucrcnlingcn, Septcnlbcr 3-4, 1991

NEACRP 3-D LWR Core Trnnsicrlt Bcnclmark .

Draft Spcciricat ions

TABLE OF CONTENTS

1. Introduction 2. Rctcrcnce Pressurized Watcr Rcactor

2.0. General 2.1. Core Geometry 2.2. Neutron Modcling 2.3. Macroscopic Cross Scctions and Derivatives 2.4. Composition Map 2.5. Doppler Temperature 2.6. Fuel Assembly Geomctry 2.7. Thennophysical Properties

2.8. Neutronics-Thennohydraulics Coupling 2.9. Operation Data

2.10. Heat Exchange Correlations 2.1 1. Pressure Drops

3. PWR Problcms 3.0. Nature of the Problcms 3.1. Calculations of the Inidal Steady Statc 3.2. Transient Calculations

4. PWR Problems: Output Rcqucstcd 4.1. Steady Statc Results 4.2. Transient Results

5. Rckrencc Boiling Watcr Rcactor

5.0. General 5.1. Core Geometry 5.2. Neutron Modeling 5.3. Macroscopic Cross Sections and Derivatives

Page j

3 i

.-

5.4. Composition Map 5.5. Doppler Tcmpcnture 5.6. Macroelement Geometry 5.7. Thennophysical Properties 5.8. Neutronics-Therrnohydraulics Coupling 5.9. Operation Data 5.10. Heat Exchange Correlations 5.1 1. Pressure Drops

,6. DWR Problems

6.0. Nature of the Problems 6.1. Calculations of the Initial Steady State 6.2. Transient Calculations

7. BVVR Problems: Output Requested 7.1. Steady State Results 7.2. Transient Results

List of Figures List of Tables

Page 15

15 16 16 16 17 17 17 IS 15 18 18

20 20 20

condcnsation "recipes") together with the refercnce 3-D solutions. tin extensive assessment

hlathematical benchmarks, based on well defined problems with a complete set of

input data and a unique solution, are widely used and accepted means of verifying the

reliability of numcrical simulations, i.e. the accuracy, stability and efficiency of numerical

met!iods. ~roblems are often very testing, but tend to be somewhat simplified - in order to

make the analysis manageable - when the purpose is the intercomparison of several different

models.

This is the case for the present benchmark, which is aimed at assess in^ the

discrepancies between three-dimensional codes for transient calculations in Light Water

Reactor cores. Thanks to the development of advanced numerical methods for larger and

more efficient computers and to the increased interest in accurate core dynamics

simulations, quite a few such codes are now available in various OECD countries. So,

NEACRP's initiative to promote this international benchmark, appears timely, and likely to

ztcrzct a Large participation.

The reference problem chosen for simulation in a PIVR is the ejection of a control assembly from the core, which may occur as a consequence of the rupture of the drive

. mechsnism casing located on the reactor pressure vessel top. This event can produce

significant, well localized perturbations of the neutronic and thermohydraulic core

pmmcten, without exceeding the safety margins. Hence, a rather realistic standard reactor

~ituation is defined, that efficiently utilizes the neutronics and thermohydraulics submodels

of thz reactor dynamics code.

The RWR problems consist in sub-promptcritical reactivity excursions generated by

rapid cold water injection or core pressurization events. This set of problems was chosen

because it is felt that the analysis of two such cases - in which the interplays of the

ncutronic and thermohydraulic effects are markedly different - represents a direct and

effective way to fulfil the objcctives of the benchmark.

Most problems in this benchmark are also suitable for one-dimensional calculations.

It is strongly recommended to participants to supply 1-D solutions (using their own

of the accuracies of I-D against 3-D approximation schemes could prove a very useful

exercise.

In the following sections the complete set of input data is given for both PWR and

BWR problems. This must be considered a draft, as the "final" data will be provided on

request to participants in diskette form.

2.2, Neutron hlodeling

2. Rcrcrcncc Prcssurized Watcr Rcactor Two prompt neutron groups, i.e, thermalized and fast neutrons, and six delayed

neutron groups are used for neutron modeling. The boundary condition for the solution of - the neutron diffusion equation & flux vanishing at the outer reflector surface.

Velocities and the energy release per fission for the two prompt ncutron groups a x given in Table 2.1, and are considered to be independent in time and space. Table 2.2

shows the time constants and fractions of delayed neutrons. No delayed energy release is

considered.

The reference scheme for the Prcssurizcd Water Reactor (PWR) is derived from real reactor geometry and operation data. One modification was introduced, consisting in the addition of a central control assembly (CA), which allowed us to sct up the problem of a

single rod ejection from a core octant with full rotational symmetry.

The set of data givcn in the following paragraphs and in the pertinent tables and

frgurcs, completely defincs the lhrcc pairs of PWR benchmark exercises. 2.3. Macroscopic Cross Sections and Derivatives

A complete set of macroscopic cross sections for transport, scattering, absorption and fission and their derivatives with respect to boron density, moderator tcmpcraturc, moderator density, and fuel temperature is defined for each composition. Table 2.4 shows the definition of all cross sections, derivatives, and reference valucs associatcd with a - composition.

The cross section at a numerical node with CA is determined by adding the cross +

2.1. Core Geometry

The radial geometry of the reactor core is shown in Figure 2.1. Radially, the core is divided into cclls of 21.606 cm width, each corresponding to one fuel assembly (FA), plus a radial reflector cell (shaded area) of the same width. Axially, the reactor core is divided

into 16 layers with heights of 7.7, 11.0, 15.0,30.0 (10 layers), 12.8 (2 layers) and 8.0 cm (from bottom to top), adding up to a total height of the active core of 367.3 cm. Uppcr and lower axial reflector have d~ichesses of 30.0 cm.

Fuel assemblies with differcnt U-235 enrichments and differcnt numbers of rods of

burnable absorbers are present in the core.The axial and radial distributions of the

section AXcA conwibuted from the CA to the cross section without CA:

enrichments and absorbers can be found in chaptcr 2.4. The radial arrangement of control assemblies is shown in Figure 2.2. The total CA

Icnght, which coincides with the absorber Icnght, is 362.159 cm. The driver device section

where p is the relative insertion in the node, i.e. OIpSl. The contribution of the CA drivcr is treated in an analogous way. The incremental cross section for the CAs (absorber type 1 :

in 2.1 % enriched FAs; absorber type 2: in 3.1 % FAs) and for their drivers, are given in

Table 2.5 with the same key as in Table 2.4. following thc top of the absorbers is distinguished from the absorber via a different cross

scction data set. No tip of control rods is defined. The position of the lower CA absorber

cdgc from the bottom of the lowcr rcflcctors is 37.7 cm for a completely inserted CA, and 301.183 crn for a completely witlidra\vn CA. Measured in units of steps, complete insertion or withdra\val of a CA corresponds to 0 and 228 steps, respectively.

.- .

kg,, is in units of W/mZ0K and Pmd is in unils of Wlcm, varying betwcen 0 and 500 Wlcm.

2.1 1 .Prcssurc Drops

A homogeneous core pressure of 155 bar is assumed.

3. PWR Problems

3.0. Nature of the Problems

The transient to bc analyzed as a function of time in three space dimensions arc

generated by the rapid ejections of a control assembly (CA) from an initially dclayed c r i h l

core at hot zero power (IIZP) or at full power (FP). The set of realistic problems offers a

variety of reactivity excursions -from about 0.1s to about 1.1s- that arc expectcd to

efficiently test both the thermal-hydraulic and neutronic models of reactor dynamics codes. With respect to axially one-dimensional solutions, which are also of great interest in this

bcnchmark, the cases are meant to present increasing levcls of difficulty for sucii approximation.

3.1. Calculations of the Initial Steady State

In order to achieve an effective multiplication factor of one, the critical steady sta:c

parameters of the reactor core have to be found from a search calculation of the critical

boron concentration for the given thermal power and CA configuration, and for \lie

parameters defined in Tables 2.7 and 2.8.

3.2.Transient Calculations

Six cases (or, better, three pairs of cases) are submitted for thc benchmark calculations. They are as follows:

Case A1 : (Figure 3.1). Core octant with rotational symmetry. Ejection of a central CA

(circled) at HZP. Case A2 : (Figure 3.2). Same as above at FP. Case B1 : (Figure 3.3). Core octant with rotational symmetry. Ejection of a peripheral

CA (circled) at HZ. Case 0 2 : (Figure 3.4). Same as above at FP.

Case C1 : (Figure 3.5). Full core. Ejection of a peripheral CA (circled) at IIZP. Case C2 : (Figure 3.6). Same as above at FP.

In all cases, the CA is ejected in 100 ms. During the whole calculation, the initial critical boron concentration and the positions of the (other) CAs are kept constant.

4. PWR Problems: Output Requested

r The presentation of results should contain steady state and transient output data as

follows. 00 7

4.1. Steady State Results

- critical boron concentration;

- radial power distribution at axial layers number 6 and 13; - maximum power value and position;

- axial power distribution (core averaged).

4.2. Transient Results

- Power history;

- core averaged fuel temperature history;

- maximum fuel temperature history:

- coolant outlet temperature at time t = 0.0, 0.1, 0.2, 0.5, I., 2., 5. , lo., 20. s.

5 . Rcfcrcncc Boiling Watcr Reactor

The reference scheme for the BWR problems is dircctly derived from existing reactors. Some minor changes have been introduced, with the purpose of making life easier

for most 3-D core dynamics codes. This is the case for the definition of a macroclement -a homogeneous average of four real BWR'fuel elements with the pertincnt control rod in the

middle- wich senlplifies the core configuration and reduces thc minimum number of nodes

in the X-Y plane by a factor of four.

5.1. Core Gcomctry

The X-Y reactor geometry is shown in Figure 5.1. The side of the square fucl

macroelement, as well as of the analogous macrocell in the radial reflector, is 30.48 cm. Axially, the reactor is subdivided into 14 layers, each 30.48 cm high, as shown in Figures 5.2 through 5.10. The largest acceptable mesh in the solution schemes will therefore be

30.48 x 30.48 x 30.48 cm.

5.2. Neutron Modeling

Two prompt neutron energy groups and six delayed neutron groups are used for neutron modeling.

Table 5.1 givcs the mean prompt neutron inverse velocities vl-1 and v2-1 and the

prompt energy release per fission, E,. These values are to be considered space and titnc

indcpendcnt. No delayed cncrgy release is considercd.

'I'able 5.2 gives the delayed ncutron fractions Pi and the time constants hi (i=1, ..., 6).

The neutron fluxes arc assumed to vanish on the reactor boundaries.

5.3. h?acroscopic Cross Sections and Derivatives

Let p and T be the water density and the Doppler temperature respectively. A n y -

n~acroscopic cross section I: (transport, absorption, fission, scattcring) must be calculated - following the formula

= I:, + (P- P,) + ( 4 ~ - 4 ~ ~ )

where:

- po and To are reference values of water density and Doppler temperatux

respectively;

- I:, is the value of I: at the rcfercnce point P,~(p,,T,);

- Z is the derivative of C with respect to water density, at the reference point P

Po=(~o,To!;

- 2,. is the derivative of I: with respect to the square root of the Doppler temperaturc,

at the reference point P,=(p,,T,).

The above equation leads to the definition not only of a nuclear composition as a

cornplete set of two-group macroscopic cross sections (LWn1 , Zag, ,vLfSI , XfJ1 . ZrB2. V I : ~ , ~ Zf,2) at the reference point Po, but also of a generalized nuclear

composition as a complete set of macroscopic cross sections and of their dcrivativcs at thc

rcference point P,.

Table 5.3 shows the list of data characterizing a generalized nuclcar composition,

including the reference values p, and To and an integer to identify the composition. Tablc

5.3 also shows the key to Tables 5.4 tllrough 5.22. Tables 5.4 through 5.22 give the data of the generalized nuclear conlpositio~is

included in our BWR problems. These data take into account a11 the rnatcrials that arc

present in the core (fuel, coolant, structures, control and burnable absorbers).

5.4. Composition Map

To relate the core geometry with the generalized nuclear compositions, 10 types of

nlacroelements (including the radial reflector macroelement) were distinguished in the map

of Figure 5.1. For each macroelement type, a number (Composition Identifier) is

associated to each layer in Figures 5.4 through 5.22. Number 19 is associated to all radial

rcflcctor layers.

So, the 3-D nuclear composition map is completely defined, in the sense that a

Composition Identifier is associated to all 3-D meshes of our reactor. That allows the

participants to calculate the actual macroscopic cross sections in all spatial meshes both in

the steady state and in the transient calculations, by knowing the local water density and

Doppler temperature.

Obviously, the macroscopic cross sections and their derivatives, as given in the

generalized nuclear compositions, are homogenized over the mesh volume, so that the

intcmal structure of a macroelement is useless from the neunonics point of view.

5 .5 . Dopp!er Temperature

The Doppler tempenture to be used in the macroscopic cross section calculations is

related to the actual temperature by the formula

T = (1-a) TFBC + a T F , ~

whrre TFIC and TFVS are the fuel temperatures at the fuel rod center and surface respectively

nnd a = 0.7.

5.6. Macroelement Geometry

The geometrical data of the macrcelement are given-in Table 5.23. PC) 00

The water flow cross-section of the macroelement docs not include the wafer gaps

between the four real-BWR elements of the m~croelement

5.7. Thennophysical Properties

The U02 density is 10.42 glcm'. The cladding material is Zircnloy-4 with a dcnsity of

6.6 glcm3.

5.8. Neutronics-Themohydraulics Coupling

The thermohydraulics affects the neutronics through the water density and thc

Doppler tempenture in each neutronic mesh.

The neutronics affects the thermohydraulics through the heat sources in the fuel and in

the coolant. It is assumed that the volumetric power density qU'(X,Y,Z,r,t) inside n fuel

rod can be approximated as follows:

qU'(X,Y,Z,r,t) = q0'(X,Y,Z,O,t) [I + y (r/rF)2] O<rlrF

where:

- r is die radius inside the fuel pellet;

- rF is the outer pellet radius;

- t is the time;

- Z is the mial position in the fuel element;

- X,Y identify a macroclement;

- the parameter y, constant in time and space, was chosen equal to 0.2.

5.9. Operation Data

At the beginning of transient, the rcactor is in equilibrium. The steady state operation data arc given in Table 5.24. The inlet mass flow through the core is properly distributed to obtain the same pressure drop across the whole core.

5.10.1-Ieat Exchange Correlations

Samc as in section 2.10.

5.1 1.Pressure Drops

The inlet orifice diameter is reduced in the peripheral core macroelements shown in Figure 5.1 1. The pertinent inlet pressure drops vs. flow rate for standard and peripheral n~acrcelcments arc shown in the Figures 5.12 and 5.13 respectively.

The frictional pressure drop inside a channel is given by the formula

\vhc:c: - ap13.2 is the frictional pressure gradient (barlcm); - G is the macroelement mass flow rate (Kgls);

- x is the steam quality; -f(x) is the frictional factor given in Table 5.25.

Tllc total pressure drop in a channel can be obtained by adding the inlct pressure drop to the

frictional one.

6 . U W R Problems

6.0. Nature of the problems

Two classes of problems (cold water injection into the core; core pressurization) arc

submitted for 3-D and 1-D solutions. In both cases, the initiating event is generated out of

the core and involves the whole core: The problems are to some extent complementary, as each type tcnds to emphasize

neutronics or thermohydraulics aspects of core dynamics, mainly due to the different timc scales. The core pressurization, which may be due to blockage of main steam isolation valve, induces istantaneous void collapsing with the pertinent reactivity effect, while the thermal feedback is slower. Oq the contrary, the cold water injection, which may be due to increase of cold feedwater flow rate or to failure of preheaters, induces void collapsing during a relatively long time (some seconds), due to the thermal inertia and to the effective

water flow rate. As a conseqyence, neutronic and thermal responses are practically simultaneous.

6.1. Calculation of the Initial Steady State

The calculated effective multiplication factor, kCrr, will be uscd to dividc the numbcr v

of neutron produced per fission, in order to obtain a critical steady state. As a consequence,

the macroscopic cross sections and their derivatives given in Tables 5.4 through 5.22 will be used only during the steady state calculations.

6.2. Transient Calculation

The cases proposed in the frame of the benchmark are:

Case Dl: Inlet cold water transient. The inlet water enthalpy vs.timc is given in

Fig.6.1.

Case El: Core pressurization. The system pressure vs.time is gi\,cn in Fig.6.2.

. - Due to the procedure adopted to establish the initial criticality, the values of v q , and

"In (and of their derivatives) to be used in the transient calculations will be those of Tables

5.4 through 5.22, divided by the k,rrcalculated for the pertinent steday-state core.

7. DWR Problems: O u t p u t Rcqucstcd

7.1. Steady State Results

- kerf;

- X-Y power density map at axial levels Z = 91.44, 182.88,274.32, 365.76 cm;

- coolant outlet density map;

- axial Dower distribution (core avenged);

- maximum power density value and position.

7.2. Transient Results

- Power history;

- core averaged fuel temperature history;

- history of core averaged coolant outlet density;

- maximum fuel temperature history;

- coolant outlet density map at time t = 0.0,0.5, I. , 2., S . , lo., 20., 60 s

List of figures

,\) Prcss~~rixetl IVntcr Reactor

Pig.2.1. - Cross section of the reactor core Fig.2.2. - Arrangement of control assemblies Fig.2.3. - Composition numbers in axial layers 1 and 18 (bottom and top rcflector) Fig.2.4. - Composition numbcrs in axial layer 2 (bottom layer of active core)

Fig.2.5. - Composition numbers in axial layers 3 through 17 (active core) Fig.3.1. - Case A 1: Initial configuration of control assemblies Fig.3.2. - Case A2: Initial config~mtion of control assemblies

Fig.3.3. - Case Bl: Initial configuration of control assemblies

Fig.3.4. - Case B2: Initial configuration of control assen~blies

Fig.3.5. - Case C1: Initial configuration of control assemblies

Fig.3.6. - Case C2: Initial configuration of control assemblies

13) Boiling Water Reactor

Fig.5.1. - BWR initial map Fig.5.2. - B I R macroelement type 1

Fig.5.3. - BWR macroclement type 2

Fig.5.4. - B \ R macroelement type 3

Fig.5.5. - BWR macroelcment type 4 Fig.5.6. - BIVR macroclement type 5

Fig.5.7. - BWR macroelement type 6 Fig.5.S. - BWR macroelement type 7 Fig.5.9. - BIVR macroclement type 8 Fig.5.10. - BWR macroelctnent type 9

Fig.S.11. - Map of macroclement inlet orifices

Fig.5.12. - Inlet pressurc drop vs. flow rate (standard macroele~nent)

2 Fig.5.13. - Inlet pressure drop vs. flow rate (peripheral macroelemcnt)

00 Fig.6.1. - Case D: inlet water subcooling vs. time

w Fig.6.2. - Case E: core pressure vs. time

List of tables

A) Pressurized Water Reactor

Table 2.1. - Velocity and energy release of prompt neutrons Table 2.2. - Decay constant and fractions of delayed neutrons Table 2.3. - Definition of ompositions Table 2.4. - Key to macroscopic cross sections tables

Table 2.5. - Cross scctions A&-- of control assemblies

Table 2.6.1. - Cross sections and their derivatives for composition number 1


Table 2.6.3. - Cross sections and their derivatives for composition numbcr 3

Table 2.6.4. - Cross sections and their derivatives for composition numbcr 4

Table 2.6.5. - Cross sections and their derivatives for composition number 5 Table 2.6.6. - Cross sections and their derivatives for composition number 6 Table 2.6.7. - Cross sections and their derivatives for composition number 7

Table 2.6.8. - Cross sections and their derivatives for composition number 8 Table 2.6.9. - Cross sections and their derivatives for composition number 9


Table 2.6.1 1. - Cross sections and their derivativcs for composition number 1 1

Table 2.7. - Data of the subassembly (FA) geometry

Table 2:8. - Steady state operation data

B) Boiling Water Reactor

Table 5.1. - Prompt neutron general data

Table 5.2. - Delayed neutron parameters

Table 5.3. - Key to macroscopic cross section tables

Table 5.4. - BWR: generalized nuclear composition numbcr 1

Tablc 5.5. - BWR: generalized nuclear composition number 2

Table 5.6. - BWR: generalized nuclear composition numbcr 3

Table 5.7. - BWR: generalized nuclear composition number 4

Table 5.8. - BWR: gencralized nuclear composition number 5

Table 5.9. - BWR: generalized nuclcar composition number 6 Table 5.10. - BWR: generalized nuclear composition number 7 Tablc 5.11. - BWR: generalized nuclcar composition number 8

Table 5.12. - BWR: generalized nuclear composition number 9 Table 5.13. - BWR: generalized nuclear composition number 10




Table 5.17. - BWR: generalized nuclcar composition number 14

Table 5.18. - BWR: generalized nuclcar composition number 15

Tablc 5.19. - BWR: generalized nuclear composition number 16

Table 5.20. - BWR: gencralized nuclear composition number 17



Table 5.23. - Data of macroclement geometry

Table 5.24. - Steady state operation data

Table 5.25. - Friction factor vs. steam quality

NEACRP

3-D LWR CORE TRANSIENT BENCHMARK fQ v-

Pressuixed Watcr Reactor 0 0 _I , . - .

A B C D E P G H I L M N O P Q R S Cy)

I 00 1

-

NEACRP

3-D LWR CORE TRANSIENT BENCHMARK

Pressurized Watcr Reactor

Pig.2.4 - Composition numben In axial layer 2 (bommlaycr of active corn)

NEACRP

3-D LWR CORE TRANSIENT BENCHMARK P

PnssPrired Water Reactor-

Fig.2.5 - Compositionnumben in uhllaycrs 3 h u g h 17 (urivc con)

NEACRP

,-- 3-D LWR CORE TRANSIENT BENCHMARK

p n s s a e d Water Reacmr

@ - CA e be ejected

X s t e p 0 228 - I

Pig.3.1 - Case Al: Initid confSgmtlon of control as3emblles

@ = CA m be ejected

CA type A C

Position in steps 100 200

2 Pig.3.2 - Case A2: Initial conlieuration of control assemblies 00 0 4

NEACRP


Pnssurixed Water Reactor

@ = CA be ejected

- 228

Fig.3.3 - Case B 1: Inidal conliguramn of control assemblies

@ = C n ' e be ejected

-

Fig.3.4 - Case B2: Initial configuration of control assemblies

NEACRP

3-D LWR CORE TRbLNSIENT BENCHMARK

Boiling Wattr Reactor + .

NEACRP

3-0 LWR CORE TRANSIENT BENCHMARK

Boiling Water Reactor

A B C D E P G H I J K L M N O P Q I

g Macmrlemenf type 1 0 Mumelement type 6

m m e k m n t type 2 ~ u m e l e m n t type 7

~ x m e k m e n t type 3 Mumelement type 8

Mumelement type 4 Marmelement type 9

,, I&umekmen~ type 5 ..:. Redid r e f i c ~ ) ~ macmelement

Fig.5.1 - BWR initialmap

C ] .... Composidonnumbcr 1 0 Composition number 3 .... Cornpo~idonnumber 2 Compositionnumber 4

Pig.5.2 - BVTR ;Ilsrroelement type 1

NEACRP


Boiling W a k r Reactor

botiom

Composition number 1 Compositionnumber 6 a Compositionnumber 5 Compositionnumber 4

Pig.5.3 - BWR macroelement type 2

NEACRP

3 - D LWR CORE TRANSIENT BENCIlMARK

Boiling W a k r Reactor

1 botiom

Compositionnumber 1 a Composition number 6

Composidonnumber 5 Composibn number 4

Compositionnumber 7

Fig.5.4 - B WR macroelement type 3

O V W

G f i f i

:2 2 :2 n n n 0 0 0 a a a

g E E U U U

Layer number

- - - m a n

G G G

n n n 0 0 0

f i c f i :2 :2 :z n n n 0 0 0

a

U

NEACRP

3 -D LWR CORE TRANSIENT BENCHMARK

Boiling Water Reacmr

U macroelement viih 38ndud inlet orifice (see Figure 5.12)

macroelement xith reduced inlet orifice (see Figure 5.13)

Fig5 1 1 - Map 01 the macroelement inlet orifices

bar

3

2

1

NEACRP


Boiling Water Rcocmr

F i g . 5.12 - Inletpnssun drop vs.flow rate ( s * d macroelemnr)

Fig. 5.13 - Inletpressun drop v3.flcw rate (peripheral mroelernen~)

NEACRP


Pressuxi~ed Watcr Rcactor

composition number charac'irbtics

1 aid rrflector 2 radial reflector 3 radial reflector n - e n m t comer 4 2.1 wlo 5 2.6 wlo 6 3.1 wlo 7 2.6 wlo, 12 burnable absorben rods (BA) 8 2.6 vlo, 16 BA 9 2.6 wlo, 20 BA 10 3.1 wlo, 12 BA 11 3.1 wlo, 16 BA

NEACRP 3-D LWR CORE TRANSIENT BENCHMARK

Pressurized Water Rcactor

ax,, , I J C

a ~ z ~ , ~ / a c

JC, , , /~TM

av=f,21aThl

a=,,/ap

aVzf.243

a ~ , , / a d ~ F

~ v z & ~ T ~

I where:

I - Comp.Nr. is the composition number, ranging from 1 to 1 1 I - c is the boron concentration @pm)

! - p is the water density (gIcm3); - TbI is the moderator temperatun CC); - TF is the Doppler temperature (OC); Table 2.3 - Defiiobn of compositioru

Reference values are labeled with subscript 0. Macroscopic cross sections are in units of crn.l.'l'l;c meanings of the indiccs of cross sections are:

1,2 fast or thermal neutron group tr transport 1-2 scattering from group 1 into group 2 a absorption f fission v number of neutrons per fission

Thc transport cross section is related to the diffusion constant D by D=1/(3 X, )

Table 2.4 - Key to macroscopic cross section ~ b l c s

NEACRP


Prcssmixed Water Reactor

0.373220E-02 -0.3 19253E-02 0.247770E-02 -0.102786E-03 -0.377989E-04 -0.213926E-01 0.255875E-01 -0.282319E-02 -0.1 15483E-02 absorber type 1

0.374092E-02 -0.3142398-02 0.242926E-02 -0.122634E-03 -0.459250E-04

-O.l67503E-Ol 0.256478E-01 -0.328086~-02 -0.134262B-02 absorber type 2

NEACRP


P r e s s d e d Water Reacmr

Table 2.5 - Cross sections A&.* of control assemblies

Table 2.6.1 - Cross sections and their derivatives for composition number 1

0.697 102E-02 -0.11 9034E-02 0.879034E-04 -0.655496E-04 -0.377989E-04

-0.113 10SE-01 0.170013E-02 0.1 16252E-02 0.500154E-03 driver

Table 2/52 - Cross sections and their derivatives for composition number 2


NEACIZP

3-D LWR CORE TRANSIENT BENCHMARK * I

Pressurized Water Reactor

HEACRP



Table 2.6.4 - Cross sectiom and their derivatives for composition number 4


Table 2.6.6 - Cross scctions and their derivatives for composition numbcr G


Table 2.6.8 - Cross sections and their derivati~ res for composition number 8


NEACRP




NEACRP



Pellet diameter C M diameter (outride) CLad wall Wkness FR pieh Guide tube diameter (ouaide) Guide tube diameter (inside)

Geometry Number of fuel p!m Number of guile tubes

Heated perirnetcr (PA) Flow cross-section (FA) Hydra& diameter

Table 2.7 - Data of the subwsembly (FA) georneoy


Con inlet tempemtut Con p n s s u 155 bar Net mass flow rhrough core 12893 Kgls

Table 2.8 - Steady sat? operation datn

NEACRP

3-0 LWR CORE TRANSIENT BENCHMARK


Table 5.1 - Prompt neumn general data

NEACIlP 3-D LWR CORE TRANSIENT BENCHRIARK

Boiling Wnler Reactor

where:

- p is the water density (glcm3); - T is the Doppler temperature (OK);

- p, is the reference water density (glcm3);

- To is the reference Doppler temperature (OK);

- C.Id. is the Composition Identifier, ranging from 1 to 19

- thc cross sections arc expressed in cm-1

Table 5.3 -Key to macroscopic cross section tables Table 5.2 - Delayed mubun paremeten

w u r


BOILIIIG WATER REACTOR

Tab.5.4 - BUR : g e n c r a l i a e d n u c l e a r compos i t i on number 1

Tab.5.5 - BWR : g e n e r a l i z e d n u c l e a r compos i t i on number 2


NEACRP

3-D LWR CORE TRANSIENT BEllCHllARK

BOILIIIG WATER REACTOR




IjEACKP

3 - 0 LWR CORE TRAllSIEllT BEIlCHMARK

BOILING WATER REACTOR

- . 8 6 7 1 1 4 E - 0 4 - .172597E-04 - .277381E-04 - .112749E-04 3 .000000E+02

T a b . 5 . 1 0 - BWR : g e n c r a l i z e d n u c l e a r c o m p o s i t i o n number 7

T a b . 5 . 1 1 - BWR : g e n e r a l i z e d n u c l e a r c o m p o s i t i o n number 8

T a b . 5 . 1 2 - BWR : g e n e r a l i z e d n u c l e a r c o m p o s i t i m b c r 9 e

IiEACRP

3-D LWR CORE TRAtlSIENT BEtlCttMARK


Tab.5 .13 - BWR : g e n e r a l i z e d n u c l e a r c o m p o s i t i o n number 1 0

Tab.5 .14 - BWR : g e n e r a l i z e d n u c l e a r c o m ~ o s i t i o n number 11

0 . 1 8 8 5 7 5 E t 0 0 0 .143771E-01 0 .101653E-01 0 . 4 1 2 8 8 8 E - 0 2 0 .167544E-02

0 . 6 9 3 6 3 6 E t 0 0 0 .694981E-01 0 .728618E-01 0 .295674E-01 12

0 . 1 2 9 1 5 8 E t 0 0 0 .198915E-01 0 .231898E-02 0 .963117E-03 0 . 4 0 3 9 6 7 E - 0 3

0 . 7 6 5 3 7 8 E t 0 0 0 .110057E-01 0 .208959E-01 0 .870788E-02 0 . 5 5 0 0 0 0 E t 0 0

0 . 0 0 0 0 0 0 E t 0 0 - .163375E-04 0 .204384E-04 0.000000E+00 0 .000000E+00

- . 8 2 5 2 8 9 E - 0 4 - .244344E-04 - .309227E-04 - .125519E-04 3.000000E+OZ

e 5 . 1 5 - BUR : g e n e r a l i z e d n u c l e a r c o m p o s i t i o n number I2

3-D LWR CORE TRAIISIENT BEIKHMARK

BOILI14G WATER REACTOR

Tab.5.16 - BWR : generalized nuclear composition number 13

Tab.5.17 - BUR : generalized nuclear composition number 14


rn

3-D LWR CORE TRAI4SIENT BEIICHMARK



Tab.5.20 - BWR : generalized nuclear composition number 17


IIEACRP

3-D LWR CORE TRANSIEIIT BEIICHMARK

B O I L l l i G WATER REACTOR NEACRP



O.OOOOOOE+OO 0 . 0 0 0 0 0 0 E - 0 0 0 . 0 0 0 0 0 0 E - 0 0 0 . 0 0 0 0 0 0 E + 0 0 3 . 0 0 0 0 0 0 E + 0 2

T a b . 5 . 2 2 - BWR : generalized nuclear composition number 19

Number of fuel rods 196

Outer clad diameter 1.430 cm

Inner clad diameter 1.267 cm

Pellet diameter 1.237 cm

Fuel rod pith 1.875 cm

Flov cross-section 400.78 cm2

Heated perimeter 880.5256 cm

Hydraulic d h t e r 1.4730 cm

Tab. 5.23 - Data of macroelement geomety

Con thermal power 1800 MW

Totalinletmslrs flow rare 13000 Kgls

Con pnssun 67.0 bar

Coolant inlet subcoohg 46.52 KJlKg

Tab. 5.24 - Steady s ts t operatj.cn da?a

. NEACRP



Table 5.25 - Friction factor vs. sbam quality

QUESTIONNAIRE 8 :

to bc returned by October 15, 1991, to:

Dr. Enrico Sartori OECD/hTEA Data Bank 91 191 Gif-sur-Yvette

FRANCE (Fax: +33-1-6941-3965)

Name.. ....................................................... Title.. .................................. Organization (full address) .................................................................................... ..................................................................................................................... ..................................................................................................................... Phone ................................. Fax ................................. Telex ..............................

1) I amlam not interested to participate in the NEACRP's 3-D LWR Core Transient Benchmark. I

have the following commentslsuggestions on the Draft Specifications:. ............................. ............................................................................................................... ...............................................................................................................

2) Pleasc mail me the final input data in a 5 114" MS-DOS HD 1.2 hlD diskette lother (specify) ... ...............................................................................................................

3) I intend to contribute lD13D13D+lD solutions (circle one) by 03.15.1932 to the following set(s) of problems (circle):

PWR : A1-A2 ; B1-B2 ; C1-C2 ; ALL ;

BWR: Dl ; El : ALL ;

by using the following codes/models: .................................................................. ...............................................................................................................

4) I unlam not planning to attend the Specialist Meeting to bc organized in June 1992 for concluding thc benchmark with the presentation and comparisons of all the results

neacrp 3-d lwr core transient bencrmark

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