ML20043C621

Simulate-3 Validation & Verification

ML20043C621
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	09/30/1989
From:	Digiovine A, Gorski J, Tremblay L YANKEE ATOMIC ELECTRIC CO.
To:
Shared Package
ML20043C618	List: BVY-90-063, Forwards YAEC-1659-A, Simulate-3 Validation & Verification ML20043C621
References
YAEC-1659-A, NUDOCS 9006050392
Download: ML20043C621 (129)

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1 SIMULATE-3 VALIDATION and VERIFICATION i

September 1988

\\

by.

A. S. DiGiovine J. P. Gorski M. A, Wemblay l

Ed 9/9/88 Prepared By:

/

~

A. S. DiGio' vine, !fe'nior Engineer (Date) l Nucle r nginee Department Prepared By:

Jj#Gorski, Nuclear Engineer (Date)

Nuclear Engineering Department Prepared By:

AAh

/

III M. A. Tremblay, Engineer [

(Date)

Nuclear Engineering Department

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. Approved By:

/

R[. Cacclapog, Manager

'(Ddte)

Reactor PhysiM Group Nuclear Engineering Department Approved By:

A B. C. Slifek Director (Date)

Nuclear Engineering Department Yankee Atomic Electric Company Nuclear Services Division 580 Main Street Bolton, Massachusetts 01740-1398 I

l Esoruary 20, 1990 l'

NW 90-032 Mr. G. Papanic, Jr.

g Senior Project Engineer t

E Yankee Atomic Electric Company 1671 Wercester Road Framingham, MA 01701

Dear Mr. Papanic:

I

SUBJECT:

ACCEPTANCE FOR REFERENCING OF TCPICAL REPORT YAEC-1659,*

" SIMULATE-3, VALIDATION AND VERIFICATION" The staff has completed its review of the Topical Peport YAEC-1659, " SIMULATE-3 I

Validation and Verification" submitted by the Yankee Atomic Electric Company bv letter dated September 12, 1988. Additional information was submitted on September 1, 1989.

YAEC-1659 presents the validation of the SIMULATE-3 code for use as an incere reactor physics analysis model. The code is to be used in the generation of spatial reactor physics calculations required in the reload licensing analysis.

l These analyses encompass whole core reactivity salculations such as boron letdewn, startup test predictions and temperature coefficient calculations.

In addition, the code will be used in detailed power distribution analysis, I

including pin-by-pin power distributions, as well as incere detector reaction rate calculations, YAEC-1f.E9 focuses upon three major applications of the SIMULATE-3 code.

The I

first is application to cperating Pressurized Vater Reactors (PWRs) and I

l includes comparison of SIMULATE-3 generated data to actual measured data, as

- g well as to the BNL PWR Core Standard Problem.

The second application is to g

operating Boiling Water Reactors (BWRs) and again includes comparisons to actual measured data, The final application focuses on the pin-by-pin power distribution capabilities of SIMULATE-3. This application compares multi-assembly SIMULATE-3 pin-by-pin power distributions to higher order transport theory solutions.

In adoition, pin-by-pin power distributions for an operating PWR are compared between SIMULATE-3 and the currently accepted method of pin 1

power distribution calculations, PDQ-7.

We find the aoplication of the SIMULATE-3 code accectable for use in reload analyses unoer the limitations celineated in the associated NRC technical evaluation.

The evaluation defines the basis for acceptance of this topical recort.

Pe do not intend to repeat our review of the matters found acceptable as I

described in YAEC-1659 when the report appears as a reference in license appitcations, except to assure that the material presented is applicable to the specific plant involved. Our acceptance applies only to the matters described in the application of YAEC-If>E9.

I

Contact:

L. Lois, SPXB/ DST J

Ext 20890 I

I

s i

r G. Papanic

-2 February 20, 1990 l

l

, I In accordance with proc +dures established in NUREG-0?90, it is requested that the Yankee Atomic Electric Comeany publish an accepted version of this topical report within three months of receipt of this letter.

The accepted version shall include an.A (desionating accepted) following the report identification symbol.

Should our criteria or regulations change so that our conclusions as to the acceptability of the report are invalidated. Yankee Atomic Electric Company and/or the applicants referencing the topical report will be expected to revise and resubmit their respective documentation, or submit,iustification

' I for the continued effective applicability of the topical report without revision of their respective documentation.

Sincerely, DRIGltL2 SI""'.n v, A. C. THW<p Ashok C. Thadani, Director Division of Systems Technology Office of Nuclear Reactor Reculation l

YAEC.1659 Evaluation l

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E A'CL OSURE SAFETY EVALUATION FOR THE TOPICAL REPORT YAEC 1659

" SIMULATE-3, VALIDATION AND VERIFICATION"

1.0 INTRODUCTION

By letter dated September IP,1988, the Yankee Atonic Electric Company submitted the Topical Report YAEC-1659 for NRC review (Ref. 1). Additional information was submitted on September 1, 1989 in response to a staff reouest.

i.

YAEC-1659 presents the validation of the SIMULATE-3 code for use as an incore reactor physics analysis model.

The code is to be used in the generation of spatial reactor physics calculations reouired in the reload licensing analysis.

These analyses encompass whole core reactivity calculations such as boron letdown, startup test predictions and temperature coefficient calculations.

In addition, the code will be used in detailed power distribution analysis, including pin-by-pin power distributions, as well as incore detector reaction rate calculations.

YAEC-1659 focuses.upon three major applications of the SIMULATE-3 code. The first is application to operating Pressurized Water Reactors (PWRs) and includes comparison of SIMULATE-3 generated data to actual measured data, as well as to the BNI PWR Core Standard Problem. The second application is to g

operating Boiling Water Reactors (BWRs) and again includes comparisons to 5

actual measured data. The final application focuses on the pin-by-pin power distribution capabilities of SIMULATE-3. This application compares multi-assembly SIMULATE-3 pin-by-pin power distributions to higher order transport theory solutions.

In addition, pin-by-pin pcwer distributions for an operating-PVR are compared between SIMULATE-3 and 'the currently accepted method of pin power distrib.ition calculations PDQ-7, Restrictions to be observed in the application of this topical report are listed in Section 3.2.6.

I

2

.j 2.0 SUMMAPY OF THE TOPICAL REPORT The repert describes the SIMULATE-3 code, which is to be used for reload physics design analyses, including generation of startup predictions, core follow calculations and physics data for safety analyses. The report demonstrates the validity for the intended application by comparing the results of SIMULATE-3 analyses to measured data from operating PWRs and BWRs.

In aedition the report demonstrates the SIMULATE-3 pin power reconstruction capability (which is required for safety an:. As) by comparing to:

(a) data from critical experiments, (b) higher order.wmerical solutions, and (c) PDO-7 generated distribution.

Finally the SIML' LATE-3 results were compared to the BNL PWP Core Standard Problem which is an NRC sponsored problem designed to test the validity of physics codes to typical reload reovirements.

I In particular, Chapter 2 gives a descriptive overview of the code followed by the PVR and BWR validation calculations in Chapters 3 and 4 respectively.

Chapter 5 discusses comparisons for the PWR pin power validation.

Finally, Appendix A outlines the comparison to the BNL standard problem.

3.0 EVALUATION I

3.1 SIMUL ATE-3, Description SIMULATE-3 has been acquired by Yankee Atomic Power Company from Studsvik of

/cerica. Detailed description is given in two Studsvik documents References 3 and 4 The code is based on the nodal technique-with coupled thermal

, g hydraulics and Doppler feedback.

The code includes the following models:

the I E two group diffusion equation, fuel assembly homogenization, the baffle /

reflector model, cross section depletion and pin power reconstruction.

The two-group diffusion model solves the diffusion equation in three dimensions, while the assembly homogenization employs flux discontinuity factors to treat the global flux shape and the assembly heterogeneous flux distribution. The flux discontinuity concept is also applied in the baffle /

i reflector region in the radial and axial directions to eliminate the need for I

I

3 L

user albedos or other adjustments at the core reflector interface.

The fuel depletion model uses functionalized microscopic cross sections to account for 4el exposure without detail tracking of nuclide concentrations.

Finally, SIMULATE-3 is used to predict the three-dimensional pin-by-pin power L

distribution in a manner which accounts for pin burnup and the neutron spectral effects.

SIML' LATE-3 uses the ENDF/B-IV cross section library with some g

' =W ENDF/E-V fission spectra. Depletion history effects are produced and stored in the TABLES-3 code for use in 51MUL ATE-3. The TABLES-3 code is part of the CASMO-3G package of codes which is presently under NRC review.

In addition to

[

pin power distribution, 51MULATE-3 calculates control rod worths and moderator, Doppler and xenon feedbacks.

3.2 SIMULATE-3 Validation l g The validation effort is divided into four parts by comparison to PDQ-7, PWR u

measurements, BWR measurements and pin power reconstruction. We shall examine each one of these cases.

1 3.2.1 SIMUL ATE-3 Versus Higher Numerical Order Calculations

' I This benchmarking has been performed by the code vendor (Studsvik) and includes 20 PWR and BWR calculations, and 3D PWR for the IAEA benchmark and fer PDQ-7 calculations.

These calculations quantified the predictive capability of the SIMt'L ATE-3 components. PWR assembly power distributions had

! g.

l 3'

a maximum rms error of 1.5 percent and up to 3 percent for BWR.

The overall accuracy is comparable to the PDQ-7 calculations which is acceptable to the staff.

In addition this benchmarking provided verification of the depletion capability of the code. The capabilities reported in this section have been incorporated in YAEC-1659 via References 3 and 4.

3.2.2 PWR Validation This validation includes comparisons with actual operating plant data from McGuire Unit 2 Cycles 1 through 3 and from Farley Unit 2, Cycles 1 through 4 (Refs. 5 and 6).

Both plants are Westinghouse, 4 loop PWRs and represent I

4 conditiers similar to those which the applicant intends to use the code.

Enrichments ranged from 2.1 to 3.2 percent U-235 with standard and optimized fuel assemblies.

The loading patterns were out-in and transition to low leakage.

Cycle burnups were from 9.9 to 14.6 Gwd/Mt. There were various numbers of I

burnable borosilicate glass absorber pins.

The three dimensional model includes four radial and twelve axial nodes.

Thermal hydraulic and Doppler feedbacks were explicitly modeled as was fuel temperature as a function of t

exposure.

I The results demonstrated that the hot full power boron depletion for all three cycles (50 points in all) was predicted to 10 ppm at 16.

The end of cycle I'

boron concentration was predicted within an absolute value of 7 pprr, and maxirron difference at the beginning of cycle of 30 ppm. These results are I

well within the ANS-19.6.1 standard and industry practice and are acceptable.

The hot full power reaction rate distribution for the in-core movable detectors and the axial offset, were 1.5 percent on the overall rms and 1.0 percent average delta-offset. The reaction rate calculation was also used for the prediction of the assembly power. The average difference in the predicted l

and measured assembly powers for the three cycles was 1.0 percent. These results are well within the limits of ANS-19.6.1 and usual industry practice and are acceptable.

There were two induced xenon transients one at middle cycle and the other at the end of Cycle 3 for which the SIMULATE-3 predicted well the axial offset.

u The ' average absolute difference in axial offset was 0.78 within i 16, for the first transient and 1.f4 within i is for the other. Due to the difficulty of xenon transient predictions and the total length of these tests the results are very good and acceptable.

Similarly the moderator temperature coefficient was predicted to within an average of 1.6 pcm/'F of the measured value toward the end of both cycles at or near the critical boron concentration.

Finally the code was used for the prediction of the startup parameters for Cycles 1, ?

and 3.

Control rod worths were estimated to an average absolute difference of 3.5 percent within 16 and the corresponding boron concentrations to an I

I s

t averace absolute difference of P1 ppm within : 16.

These values are well within the Afd-19.6.1 criteria of 2 10 percent for the control rods and e 50 ppm for the boron concentration and thus are acceptable.

As with the McGuire-2 calculations SIMULATE-3 was used by Studsvik for the Farley-2 Cycles 1 throuch 4 to compare measured to calculated results.

Farley-2 is a three-loop 157 assembly, 2652 Wth Westinghouse plant.

In all four cycles standard 17xl? L'estinchouse fuel was used with enrichments from 2.1 to 3.45 w/o U-235.

The fuel rods contained a varying nurber of burnable

- I:'

poison pins of borosilicate glass.

l The following results are reported from Reference 7 l

5 Boron Letdown:

For a total of 48 measurement to calculation comparisons the average absolute dif ference is 12 ppm within 1s of 9 ppm, The end l

of cycle prediction average for the four cycles is 4 pom which is eFcellent agreement for the end of Cycle.

i In Core Hot-Full-Power Detector Response:

For a total of 15 flux maps-the rms differences of measured to calculated reaction rates is near 1 percent for all cycles.

l Hot-Full-Power Distributions: The results are similar to the detector L-response, i.e., within 1 percent.

l.

Hot-Zero-Power Power Distributions: ' The results are similar to the hot-full-power distributions but the uncertainties are larger as j

expected, thus, the averace absolute difference.is 2.7 percent within j

1 3 G.

Hot-Zero-Power Start-Up Parameters; Con' trol rod worths measured with the rod-swap and the conventional boration dilution method showed an average absolute measured versus calculated difference of 3.2 percent with a t 16.

I

in sumary it has been demonstrated that the code can predict PWR plant operatino parameters within acceptable limits.

I; 3.2.? FKP Validation I

As in the PWR validetion, here also the validation consists of modeling several cycles of two operating PWRs in a manner similar with the intended applications.

The first application is an analysis of the Quad Cities Unit 1 Cycles 1 and ?,

performed by the applicant.

Calculated results were compared to the corresponding measured values for:

eigenvalues as a function of expnsure, cold criticels at the beginning of Cycle 1 cold criticals during Cycles I and

!~

2 and traveling-in-core probe traces during both cycles. The individual assembly enrichments ranged from 1.7 to 2.73 percent U-235.

Gadolinium loadings ranged from 0.5 to 3.0 percent weight in Gd 0. PICBURN-3, CASMO-3G, p3 TABLES-3 and the EPRI void correlation were used with SIMULATE-3.

Twenty-four d "' "'

d'

" d' P ""b ""'

"d- '"' '** '""

E iW data were taken from EPRI reports in References 7 and 8.

The overall hot eigenvalue is 1.00415 with : 16. The ecuilibrium xenon cases have very consistent distribution'about the average.

For the cold criticals during l

Cycles 1 and ?'the average eigenvalue is 0.99586 within t 17.

Finally the traveling-in-core probe differences of the measured versus the calculated absolute overall average is 5.46 percent with a standard deviation of 5.52 percent.

In sumary the Quad Cities Unit I eigenvalues and power distributions lI demonstrated consistent calculational capability well-within the ANS-16.9-1 standard values, thus, they are acceptable.

l The second validation was performed by the code vendor Studsvik of America and involved measurement data from the Forsmark Unit 1 (Pef. 9).

Forsmark Unit 1, is a 676 fuel assembly plant, 2700 MWth with 36 in-core-traveling probes.

Gama sensitive power measuring devices were installed at the end of Cycle 5, which eliminate a large part of the uncertainty due to local asyrrnetry from lI

7 L

i-L neutron activated in-core probes. The fuel designs of Cycle 1 through 6 which were analyzed included 8x8 and 9x9 configurations with enrichments from 2,08 percent to 3.00 percent by weight U-235 and Gadolinium loading from 0 to 3.95 percent by weight in Gd 0. The hot eigenvalue calculations indicated an p3 average Cycle 1 through 6 value of 1.00429 within a standard deviation of 0.00114 The individual cycle eioenvalues (except for Cycle 1A which did not have a smooth power history) range from 1.00270 to 1.00725 Similarly the cold criticals resulted in an averace eigenvalue of 0.99857 with a standard deviation I-of 0.00108.

Finally the in-core-traveling probe measurenents, in a similar manner as for Quad Cities, produced computed to measured average absolute differences of 0.017 within a standard deviation of 0.0027 We conclude that-I SIMULATF-3 demonstrated the capebility to predict power distributions and critical states within acceptable accuracy limits.

l 3.2.4 PWR Pin Power Reconstruction Validation l I This is the final step in the validetion to demonstrate the capability to calculate individual pin power. One such validation by comparison was I

performed with the measured critical experiment pin power distribution performed by B&W (Ref.10). The results demonstrated that SIMULATE-3 can LI calculate pin powers to an res difference of 0.8 percent in lattices without cadolinia and 1.3 percent with gadolinia. However, this verification did l

not include depletion.

Further validation was provided from higher order numerical calculations using CASMO-3G and pin-by-pin power distribution using diffusion theory via PDQ-07 These results show the SIMULATE-3 to be within 1 percent of the CASMO-3G, PDO-07 solution even in redded configurations.

For control rod insertions this was also the case except that the worst difference increased to about 3 percent. The SIMULATE-3 calculations are more realistic than diffusion theory based peak power estimates.

Another benchmark was based on a two cycle depletion of a quarter core l

including the core-reflector interface which has a large flux gradient and spectral history different than the interia infinite medium spectrum. A McGuire quarter core was selected with the Cycle 1 and Cycle 2 loading patterns. Under these conditions pin power estimation is most difficult, yet the SIMULATE-3 I

I

I e

peak-pin power per assembly for all assemblies is within 2.5 percent of the PDO-07 value. The difference in the peak pin for the entire core exposure showed an average agreement within 1.1 percent.

The above comparisons demonstrated excellent capability by SIMULATE-3 to calculate pin power distributions. Although no fontal criteria or standards for pin power distribution exist, its favorable comparison to the PDQ-07 results make it acceptable, 3.2.5 Corroarison to PNL. Standard PWR Problem As part of the SIMULATE-3 validation, the Brookhaven National Laboratory (BNL)

PWR Core Standard problem was analyzed. The problem was commissioned by the NRC and was designed to test the validity of a particular code applied to typical reload physics calculations.

The areas compared were:

l

• core average radial power distribution

' core average radial exposure distribution

E

' ' '"' ' ** " **' d'" "

W

• the power defect

• the suberitical boron concentration

' the Doppler c'oefficient

' the moderator temperature coefficient

• differential rod worth, and

' the integral rod worth The results of the comparison were reviewed by BNL and "... found to be in generally good agreement...."

!I 3.2.6 Limitations The following limitations are imposed in the use of the SIMULATE-3 code:

'I 1.

SIMULATE-3 is to be used for PWR and BWR steady-state physics analyses, i

and

.I 1

I l

~..

9 2.

$!MULATE-3 is to be used with the approved versions of CASMO-3G and TABLES-3 codes, i.e., is subject to the approval of YAEC-1363.

a.0 FUMMARY AND CONCLUSIONS I

The staff reviewed the Topical Report YAEC-1659 "$1MULATE-3 Valfdation and Verification" and found that:

LE th' ' d' ' "*'" " " " " 6'""h**d '" """ '"' """ """'d

!E parameters the BNL standard PWR problem and for its ability to estir. ate PWR pin power distribution, LI l2 the results showed that SIMULATE-3 is able to estimate operating parameters well within the American National Standard requirements and the accepted industry practice, I

L the Applicant demonstrated the code's ability to predict pin power distribution, and the Applicant demonstrated the code's ability to accurately reproduce the results of the NRC sponsored standard PWR problem.

Considering the above the staff finds the application of the SIMULATE-3 code acceptable for referencing in future reload submittals subiect to the limitations noted in Section 3.P.6 of this evaluation.

5.0 REFERENCES

I, 1.

Letter from G. Papanic, Jr., Yankee Atomic Electric Company to USNRC,

" SIMULATE-3, Validation and Verification " September 12, 1988.

2.

Letter from G Papanic, Jr., Yankee Atomic Electric Company to USNRC,

" Additional Infomation on YAEC-1659 SIMULATE-3, Verification and Validation," September 1, 1989.

?-I.

I g

10 I_:

3.

Studsvil/NFA-8617. "CASMO 3, A Fuel Assembly Rurnup Program" by M.

Edenius, A. Ahlin and H. Haggblom, dated November 1986, 4

Studsvik/50A-88/03, ' SIMULATE 3:

The Studsvik Steady State Nodal Peactor Analysis Code" by K. S. Smith et al., dated Aueust 1988 5.

YAEC-1608, "McGuire Unit ? $1MULAir-3 Benchmark Analysis Cycle 1 Through 3" by A. S. DiGiovine et al., October 1987.

6 Studsvik/SOA-87/10,

• SIMUL ATE-3 Benchmark Report Furley linit II" by K. S.

Smith et al., April 1987 l

7.

EPRI NP-?40, " Core Design and Operating Data for Cycles 1 and 2 of Quad Cities 1" by N. H. Larsen et al., dated November 1976.

P.

EPRI NP-552, " Core Design and Operating Data for Quad cities 1 Cycle 3" by N. H. Larsen et al., dated March 1983, l

9.

Studsvik/NFA-88/44, " SIMULATE-3 BWR Benchmark, Forsmark Unit 1, Cycles 1 Through 6" by H. Hakansson et al., dated July PS, 1988.

, I 10.

" SIMULATE-3 Pir Power Reconstruction:

Benchmarking Against the B&W Critical Experiments," Trans. Am. Nucl. Society, Vol. 56 pg 531, San Diego, Cal., by K. S. Smith, June 1988.

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I DISCLAIMER I

This document was prepared by Yankee Atomic Electric Company for its own use.

The use of information contained in this document by anyone other than Yankee Atomic Electric Company is not authorized, and in regard to unauthortv;f use neither Yankee Atomic Electric Company or any of its officers, directors, agents or employees assumes any obligation, responsibility or liability, or makes any warranty or representation, with respect to the contents of this document, or its accuracy or completeness, I

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ABSTRACT I

This report presents the validation of the SIMULATE 3 computer code for use as an j

incore reactor physics analysis model. The intended use of the code is in the generation I

of spatial reactor physics calculations typically required as part of reload licensing analysts.

These analyses encompass global reactivity calculations such as boron j

letdown. startup test predictions, temperature coefficient calculations, etc. In addition, the code will be used in detailed power distributton analysts, including pin by pin power distributtons, as well as incore detector reaction rate calculations, in a previous document. YAEC 1363. CASMO 3G VALIDATION, Yankee Atomic Electric presented the lattice physics computer code that will be used to supply required nuclear data constants to the SIMULATE 3 code. The validation presented in this report j

I uses only data supplied by the CASMO 30 computer code, l

This report focuses upon three mqfor applications of the SIMULATE 3 code. Thefirst is application to operating Pressurtzed Water Reactors (PWRs) and includes comparison of SEMUIA7E-3 generated data to actual measured data. as well as to t!w BNL PWR Core Standard Problem. The second application is to operating Bolling Water Reactors (BWRs) arnt again includes comparLsons to actual measured data. The final application focuses on the pin by pin power distribution capabilities of STMULATE-3. This application compares multi assembly SIMUIATE-3 pin-by-pin power distributtons to h@her order transport tlwory solutions. In addition, pin by pin pouer n

distributionsfor an opemttng PWR am compared between SIMULATE 3 and the currently accepted method ofpin power distributton calculations, PDQ-7.

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I TABLE OF CON'IENTS I

DISC 1 AIMER,............................................

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ABS'IRACT,.............................................

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UST OF TABLES vil UST OF FIGURCS x

ACKNOWLEDGEMENTS xil 1.0 IN1RODUC110N/ SCOPE OF APPUCATION..,,.,,,,.......,,,....

1 2.0 SIMUIEIE 3 OVERVIEW 5

2.1 Description of SIMUIATE-3...,,,,.........,,...........

5 2.1.1 'Iwo Group Diffusion Model...,,..................,,.

5 2,1,2 Assembly Homogenization Model....,..................

6 2.1.3 Bafile/ Reflector Model 6

2.1,4 Fuel Depletion Model.......,,.....................

6 2.1.5 Pin Power Reconstruction Model............,,,,,......

7 2.2 CASMO 3G Overview...................................

8 2.3 TABLES 3 Description..................................

9 2.4 Calculational Capabilities........,,,..........,,,,,,,..

10 I

- 2.5 Validation 10-3.0 PWR VALIDATION CA14U1A110N 15 3.1 McGuire Unit 2. Cycle 1 Through 3 Validation...................

15 3.1.1 Model Description..................................

15 3.1.2 Boron IAtdown Results 16 3.1.3 HFP Detector Calculations 16 3.1.4 HFP Power Distributions.............................

17 3.1.5 HFP Xenon Transient 17 I.

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3.1.6 HFP EOC Moderator Temperature Coemclents 19 3.1.7 HZP Startup Test Predictions..................,,.......

19 3.1.8 McGuire Unit 2 Summary................,,,,.,,,,..

20 3.2 Farley Unit 2. Cycles 1 through 4 Validation....................

20 3.2.1 Model Descdption...........,...............,,,.

21 3.2.2 Boron Letdown Results 21 I

3.2.3 HFP Detector Calculations 22 3.2.4 HFP Power Distributions.........................,,.

22-3.2.5 HZP Power Distributions,,,,........................

22 I

3.2.6 HZP Startup Test Predictions.....,,,..........,,....

23 3.2.7 Farley Unit 2 Summary.............................

23 4.0 BWR VALIDADON CALCUIADON 41 4.1 Quad Cities 1 Cycles 1 and 2 Validaiton 41 4.1.1 Quad Cities 1 Model Description 41 4.1.2 Hot Eigenvalue Calculation. Cycles 1 and 2......,,,.......

42 4.1.3 Cold in. Sequence Criticals. Cycles 1 and 2 42 4.1.4 Cold local Criticals. Beginning of Cycle 1 43 I

4.1.5 'nP Trace Evaluation. Cycles 1 and 2 43 j

l 4.1.6 Quad Cities 1 Summary................,,........

4.

44 4.2 Forsmark Unit 1 Cycles 1 Through 6 Validation..................

44 4.2.1 Forsmark 1 Model Description....................

45 4.2.2 Hot Eigenvalue Calculations. Cycles 1 Through 6.............

45 4.2.3 Cold Critical Calculations. Cycles 1 Through 6..............

46 4.2.4 'nP Trace Evaluation. Cycles 1 Through 6.,,...............

46 I

4.2.5 Forsmark 1 Summary....................,,........

47 5.0 PWR PIN POWER RECONSIRUCDON VALIDA"uON 65 i

1 5.1 Validation versus Measured Critical Experiments 65 5.2 Validation versus Measured Reaction Rates.....................

66 5.3 Validation versus Higher Order Numerical Calculations 66 5.4 PWR Quarter Core SIMULAIE 3 Reconstruction Compared to PDQ 7.....

68 l

5.5 Pin Power Reconstruction Summary 69 6.0

SUMMARY

AND CONCLUSIONS 83

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7.0. REFERENCES 89-APPENDIX A.......................................... -

A1 APPENDIX B.......................................................

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I LIST OF TABLES I

Number 110c Eagt 1.1 last of Key PWR Physics Parameters 3

I 1.2 List of Key DWR Physics Parameters 4

2.1 PWR Functional Dependencies 13 2.2 BWR Functional Dependencies 14 r

I 3.1 McGuire Unit 2 Operating Characteristics 24 I

3.2 McGutm Unit 2 Cycle 1 Boron letdown 24 3.3 McGuire Unit 2 Cycle 2 Boron Letdown 25 3.4 McGuire Unit 2 Cycle 3 Boron Letdown 25 1

3.5 McGuire Unit 2 Cycles 1 Through 3 Detector Comparison Case Description ' 26 3.6 McGuire Unit 2 Cycles 1 Through 3 Comparison of Predicted and Measured 27 Axially Integrated Average Reaction Rates and Axial Offsets 3.7 McGuire Unit 2 Cycles 1 Through 3 Comparison of Predicted and 28 Measured Assembly Powers 3.8 -

McGuire Unit 2 MOC 3 Xenon Transient 29 I

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3.9 McGuire Unit 2 EOC 3 Xenon Transient 30 l

3.10 McGuire Unit 2 EOC HFP Moderator Temperature CoefIlcient 31 lI

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I 3.11 McGuire Unit 2 Cycle 1 HZP Control Rod Worths 31 3.12 McGuire Unit 2 Cycle 2 HZP Control Rod Worths 31 3.13 McGuire Unit 2 Cycle 3 HZP Control Rod Worths 31 3.14 McGuire Unit 2 Cycle 1 HZP Boron Endpoints 32 3.15 McGuire Unit 2 Cycle 2 HZP Boron Endpoints 32 I

3.16 McGuire Unit 2 Cycle 3 HZP Boron Endpoints 32 I

3.17 McGuire Unit 2 HZP BOC isothermal Temperature Coeflicients 32 3.18 Farley Unit 2 Operating Characteristics 33 3.19-Farley Unit 2 Cycles 1 Thru 4 Comparison of Predicted and Measured 33 Reaction Rates. Axial Offsets, and Assembly Powers 3.20 Farley Unit 2 HZP Core Average Axial Detector Reaction Rates 34 I

3.21 Farley Unit 2 Cycles 1 and 2 Comparison of Predicted and Measured 34 Reaction Rates, Axial Offsets, and Assembly Powers at HZP BOC 3.22 Farley Unit 2 Cycle 1 HZP Control Rod Worths 35 l

. 3.23 Farley Unit 2 Cycle 2 HZP Control Rod Worths 35 3.24 Farley Unit 2 Cycle 3 HZP Control Rod Worths Using Rod Swap Technique 35 3.25 Farley Unit 2 Cycle 4 HZP Control Rod Worths Using Rod Swap Technique 35 4.1-Summary of Quad Cilles Cycles 1 and 2 Hot Depletions 48 I

4.2 Quad Cities Cycles 1 and 2 Cold Critical Cases 49 I

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E 4.3 Summary of Quad Clues Cycles 1 and 2 'I1P Data 50 4.4 Characteristics of Forwnark Unit 1 Fuel Designs 51 4.5 Forsmark Unit 1 Hot Eigenvalue Summary 52 4.6 Forsmark Unit 1 Cold Critical Summary 52 I

5,1 SIMUIATE 3 Pin Power Validation for B&W Critical Experiments 71 5.2 Colorset Analysis Case Listing and Identifiers 72 I

5.3 Colorset Verification Results 73 5.4 McGuirr Unit 2 Cycle 1 Comparison Between PDQ 7 and SIMUIATE 3 of 74 Peak Pin 6.1 Summary of SIMULNIE 3 Accuracy for PWR Application 85 l

6.2 Summary of SIMUIA'IE 3 Accuracy for BWR Application 87 6.3 Summary of SIMUIATE 3 Accuracy for PWR Pin Power Reconstruction 88 A.1 eNL PWR co,e Staneare e,oslem calculatee ea,amete,.

A.2 A.2 BNL PWR Core Standard Pmblem BNL/YAEC Comparison Results A3 LI I

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LIST OF FIGURES I

Number Dguc:

Eagt I-1 3.1 McGuire Unit 2 Cycle 1 Summary of Reaction Rate Comparisons 36 3.2 McGuire Unit 2 bycle 2 Summary of Reaction Rate Comparisons 37 3.3 -

McGuire Unit 2 Cycle 3 Summary of Reaction Rate Comparisons 38 3.4 McGuire Unit 2 Cycles 1 Through 3 Summary of Reaction Rate 39 Comparisons 3.5 McGuire Unit 2 Cycle 3 Xenon Transient 40 i

4.1 Quad Cities Cycles 1 and 2 Hot Eigenvalues 53 4.2 Quad Cities Cycles 1 and 2 Cold Eigenvalues Results 54 I

4.3 Quad Cities Unit 1 Cycle 1 BP Trace Summary Predicted Measund 55 Average MP Integrals 4.4 Quad Cities Unit 1 Cycle 2 UP Trace Surrunary Pmdicted - Measund 56 Average EP Integrals 4.5 Quad Cities Unit 1 Cycle 1 BP Trace Comparison Summary 57 4.6 Quad Cities Unit 1 Cycle 2 RP Trace Comparison Summary 59 Forsmark Unit 1 BP Trace Summary Cycles 1A Rtrough 6 61

. 4.7 l

Predicted - Measured Avenge TIP Integrals 4.8 Forsmark Unit 1 TIP Trace Summary Cycle LA Predicted - Measured 62 Average TIP Integrals

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,,. 4.9 Forsmark Unit 1 MP Trace Summary Cycle 6 Predicted Measured 63 Average UP Integral s

4.10 Forsmark 1 Cycle O Core Average Axial EP Comparisons 64 5.1 McGuire Unit 2 Cycle 1 loading Pattern and Naming Convention 75 I

5.2 McGuire Unit 2 Cycle 2 loading Pattern and Naming Convention 76'

' I-5.3 McGuire Unit 2 Cycle 1 BOC and EOC Comparison Between PDQ-7 77 and SIMUIA'IE 3 Peak Power Pin by Assembly I

5.4 McGuire Unit 2 Cycle 1 BOC.15 Gwd/Mt Comparison Between PDQ 7 78 and SIMULATE 3 of Assembly Power 5.5 McGuire Unit 2 Cycle 1 EOC 14.0 Gwd/Mt Comparison Between PDQ 7 79 and SIMUlA*lE 3 of Assembly Power I

5.6 McGuire Unit 2 Cycle 2 BOC.15 Gwd/Mt Comparison Between PDQ 7 80 and SIMULATE-3 of Assembly Power 5.7 McGuire Unit 2 Cycle 2 BOC.15 Gwd/Mt Comparison Between PDQ 7 81 and SIMUIA'IE 3 Peak Pin by Assembly 5.8 McGuire Unit 2 Cycle 2 BOC (.15 Gwd/Mt) Comparison Between PDQ 7 82 I

and SIMUIATE 3 Pin Distribution of Upper Id Quadrant of Assembly 28 Which Contains Peak Pin E

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ACKNOWLEDGEMEh%

l The authors wish to acknowledge all the indMduals who assisted in the development of 1

the CASMO SO/SIMUIEIE 3 methods program. Many different people within Yankee Atomic assisted with their ideas and support, in particular, we would like to acknowledge R A.

Wochlke K. J, Morrissey, and G. M. Solan for their assistance. In addition, many indMduals outside of Yankee Atomie also helped us. We thank M. Edenius, K. S. Smith, and D. M. Ver Planck of Studsvik of America for their assistance. Finally, thanks to N. Barbetta for helping us put the reports together.

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' I 1.0 IhTRODUCnON/ SCOPE OF APPLICA*nON I

his mport describes the SIMUIAE 3 computer code, which is intended to be used by 8

Yankee Atomic Electric Company for reload physics design analysis The code is capable of perfortning all calculations currently performed with codes such as SIMULATE 2 and PDQ 7."

his includes generation of startup test predictions, core follow calculations, and physics data I

for safety analysis.

We intent of this report is to describe the computer code and demonstrate the validity of applying the code to reload physics design analysis.

An overview of the code and the associated spectrum lattice code, CASMO-3G, is included." In addition, the report is divided into three major categories; application to Pressurized Water Reactors (PWRs), application to Boiling Water Reactors (BWRs), and application to generating detailed pin by pin power distribuuons.

For the PWR application, several key physics parameters are calculated using SIMUIA'IE 3 and compared to measured data from two operating PWRs. These parameters are typical parametem associated with reload physics core design and are listed in Table 1.1.

For the BWR appucation, several key parameters were also generated and compared to operating BWR data. Rese parameters are hsted in Table 1.2.

The final application is the demonstration of the SIMULATE-3 pin power reconstruction capability. 21s capability includes pin by pin power distribution generation, which is required for safety analysis evaluations. 21s data is generated using SIMULATE-3 and compared to I

measured data from critical experiments, and comparisons to higher order numerical benclunark solutions, in addition, a comparisen is made between SIMUIATE-3 pin power distributions and those generated using a PDQ 7 model of an operating PWR. The final section of this report contains a summary of the typical accuracy associated when using SIMUIATE-3 to calculate key reactor physics parameters.

As further verification of the SIMUIATE-3 code, the code is applied to the BNL PWR Core l

his problem was designed by Brookhaven National laboratory for the Standard Problem.e l.

purposes of testing the validity si a particular physics code applied to a typical reload tpe I

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- calculation, as well as to test the competence of the engineem using the code in performing reload analysis.

Since the intent of this report is consistent with these purposes, the evaluation that Yankee Atomic itceived imm Brookhaven is included in Appendix A of.this I

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TABLE 1.1 LIST.OEEN PWR PlWSICS PARAMMERS Cntical Boron Concentration versus Cycle Exposure Cycle length Detector Fission Reaction Rates Radial Assembly Power Distribution Axial Power Distribution Control Rod Worths I

Critical Boron Concentrations at flot Zero Power versus Control Rod Insertion Ternperature Coefficients Axial O!Tset During Xenon Transients at Power I

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I TABLE 1.2 j

g UST OF KEY BWR PHYSICS PAFAMETERS W

i K efective versus cycle Exposurt local Cold criticals g

in. sequence cold cr$ticals

-,mc.s.,,usc,cl.m -.

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I 2.0 $1MUIAE-3 OVERVIEW his section provides a brief overview of the SIMUIAE 3 code used for PWR and BWR steady state phyaks analysis. An overview of the lattice physics code CASM'O 30, as it i

pertains to SIMUIA'IE 3, is also included, his overview is intended to summarize the characterisucs of the codes, the modehng appmximations employed, and some of the I

benchmarking provided by the code vendor, Detailed descriptions of the theory of each code are contained in separate documents provided by the code vendor, Studsvik of America."

i 2.1 Desertotion of SIMUIATP 3 l

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The SIMULATE 3 code is an advanced nodal code for performing steady state incore physics calculations, with coupled thermal hydraube and Doppler feedback.

We physics models employed are quite different than those of corntnuonal nodal codes. Rese models have chminated the need for user adjustable parameters (e.g., albedos, thennal leakage corrections, etc.), and all of the physics data required for SIMULATE 3 is obtained directly from CASMO 30 assembly spectrum / depletion calculations. We SIMULATE 3 code consists of five physics models: the two group diffusion equation model, the fuel assembly I

homogenization model, the baflie/ reflector model, the cross section/ depletion model, and the pin power reconstruction model.

2.1.1 Two-Groun Diffuminn Model The spatial neutronics model used in SIMULATE 3 is called the QPANDA model which solves the three dimensional, two-group neutron diffusion equation: using forth-order polynomials, to represent intm nodal, flux distributions, in both fast and thermal groups.

QPANDA explicitly treats group to group coupling effects on the intra nodal flux distributions, an important phenomenon which is ignored in conventional nodal models.

We spatial distribution of exposure within each node is also used to impmve the accuracy of the nodal couphng coefficients and to provide information required for calculation of pin by pin power distributions. We nodal couphng equations are solved by a nonhnear iterathc technique in which polynomial couphng equations are sohrd analytically, using sources obtained from the global flux iteration.

We flux iteration is performed using Wielandt's fractional iteration (eigenvalue shifting) and the Cyclic Chebyshev SemiIterative method.

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2.1.2 mmbly llomncentation Model I

De use of a nodal model huplies that the fuel assemblies are treated as homogenous. It is weD known that conventional flux volume weighting of cross sections leads to mispredictions in the couphng between assemblies, and correspondingly, to mispredicuons in the reactor power distributions.

In order to avoid introducing such ermrs, SIMUIAE 3 I

employs the use of flux discontinuity factors to treat the spatial homogenization of fuel assembhes, ne disconttnuity factors are derived from the assumption that the flux distribution is comprised of two pieces: a global shape (homogeneous smooth flux distribution) and a local shape (heterogeneous assembly flux clistribution).

This assurnpuon allows discontinuity factors (ADFs) to be edited from the same assembly calculations performed to compute two group cross sections.

When used in the QPANDA model, ADFs alter the prediction of neutron currents between assemblies and effectively chminate homogenizauon ermrs.

I 2.1.3 Baffle / Reflector Modei The discontinuity factor concept is also applied to the modehng of homogenized baffle / reflector nodes in SIMUIATE 3. The CASMO 30 reflector opuon is used to sohe a fuel assembly /bafDe/ reflector region. This data is used in the QPANDA model to represent all radial reflector nodes, and detailed tests have demostrated that this model eliminates the need for any user adjustments in the baffle / reflector modehng. A similar model is also used for upper and lower axial reflectors.

I 2.1.4 Fuel Deoletion Model The fuel assembly depletion model used in SIMUIAE-3 uses macroscopic cross section data functionahzed versus exposure and " history" variables.

We use of history variables (covered in more detail in Section 2.3) allows SIMULAE 3 to accurately model the effects of local conditions (e.g., moderator density, inserted control rods, etc.) on the depleuon-induced changes in nuclide concentrations. 21s macroscopic depletion model permits SIMUIATE 3 to account for microscopic depletion effects, without tracking nuchde concentrations in each node of the reactor model.

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I 2.1.5 Pin Power Reconstruction Model SIMUlA1E 3 is also capable of accurately predicting three dimensional pin by pin power distributions. IndMdual pin powers are computed by assuming spatial separability of the smooth intra nodal power distributions and the local pin by pin power distributions.

I Calculauon of the smooth intra nodal power distribuum requires evaluauon of the intra nodal flux and fission cross secuon distribution. Intranodal flux distribuuons are computed using bi quadratic flux expansions with constraint terms taken from the QPANDA surface averaged fluxes, currents and nodal corner point fluxes. The corner point fluxes are evaluated by using the QPANDA fourth ortler flux distributions and continuity conditions at corner points which assure corner point continuity of reconstmeted fluxes.

I Calculation of the intra nodal cross section distribution requires treatment of two distinct phenomena. First, the effects of depleting nodes with asymmetric (tilted) flux distributions are I

evaluated by expanding the intra nodal exposure distribution in bi quadratic polynomials and calculating pin by pin homogenized fission cross sections as functions of the pin by ptn exposures.

Secondly, the effects of spectrum interaction from neighboring assemblies are treated. These spectral history" effects develop because the spectrum which exists on the surface of an assembly is affected by neighboring assemblies, and this spectrum is different from that assumed in the infinite medium CASMO 30 assembly depletion calculauons.

Consequen0y, the actual cross sections for the surface of an assembly depend not only on the exposure, but also on the spectrum which existed as the assembly was depleted. These spectral interacuon effects are modeled in SIMULATE 3 by continuously integrating the I

spectra for each assembly surface and by evaluating the surface cross sections based on the exposure integrated spectra.

Once the intra-nodal flux and homogentzed fission cross section distdbutions are known, the intra nodal power distribution can be computed. Pin by pin power distributions are then computed by multiplying the intra nodal power distribution with the CASMO 30 pin by pin power distributions.

The CASMO-3G power distributions account for all of the assembly heterogene1 ties (water holes, burnable absorber pins, etc.), and the intra nodal power distribution accounts for all of the gross power tilts and assembly interaction effects. The 7

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spectral history treatment in SIMUIAE 3 allows all pin power distnbutions to be compute from single assembly CASMO-30 calculations. All CASMO-3G pin power distnbutions are evaluated such that the distrfbutions reflect local conditions (e.g., exposure, moderator density, etc.) which exist in each node.

I 2.2 CASMO-30 Oven tew I

CASMO-3G is a two-dimensional, multi-group transport theory code for the calculation of eigenvalue, spatial reaction rate distnbutions, nuchde depletion of pin cells, and depleuon of I

BWR and PWR fuel latuces. The code is an improved venton of the CASMO and CASMO 2 codes.

It is capable of modeling crucifonn control rods containing cyhndrical absorber elements, cluster control rods, water gaps, incore instrumentation channels, boron curtains, burnable absorber rods, bumable absorbers within the fuel rods, and fuel rods loaded with gadohntum. CASMO 3G includes a bafDc/ reflector cross section generation model, and the abihty to generate assembly flux discontinuity factom for use in SIMULAE 3, I

The nuclear data library is based on ENDF/B IV with some fission spectra data taken from ENDF/B V, The data are collected in a library containing cross sections in 40 energy I

groups, for neutron energies from 0 to 10 Mev.

The data required from CASMO 30 for SIMUIAE-3 consists of two group cell average macmscopic cross sections, and microscopic cross secuona for xenon, samar.um, and soluble boron. Also required are discontinuity factors, pin by pin distributions, coraer point Duxes, and detector reacuon rates.

For a variety of individual reactor statepoint conditions, the CASMO 'JG code is used to calculate spectrum weighted information for each unique fuel type to be analyzed with I

SIMULAE 3. Wese conditions include variation of moderator temperature, soluble bomn (for PWRs), insertion of control rods, fuel temperature and exposure. History effects based on depletion at off nominal conditions, are also included.

his data is then processed into tabular form by the TAB 12S 3 code for use in SIMUIAE 3. Yankee Atomic has submitted a topical report on the CASMO 30 code for application to SIMUIAE 3.* That report desenbes the theory and presents vahdation of CASMO 30.

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I 2.3 TABIES-3 Description The TABLES-3 computer code processes the informauon imm single assembly CASMO 30 lattice calculadons into a tabular fonn for use in the SIMULATE 3 code.' ne methodology used by TABLES 3 is similar to that used by the TABLES 2 computer code.

We methodology is based upon defining node average cross sections which are a summation of

" partial cross sections", ne partial cross secuons are funcuons of the different statepoints I

analyzed at the CASMO 30 level.

The TABLES 3 code funcuonahzes the " partial cross sections" by subtracting the cross sections generated from each of the statepoint calculations from some base condiuon cross section. The functionahzauons are presented in Tables 2.1 and 2.2 for PWR and DWR apphcations, respectively. Dese functionalizauons are typical of those used for many years with nodal codes for reload analysis. History effects are included in the cross section model in order to improve the accuracy of the cross section 4

representauon. The intent of these histories is to provide accuracy comparable to a detailed microscopic analysis.

I in a PWR, the history effects that are most prominent are soluble boron history and moderator history, in a BWR, the dominant history effects are void history and control rod I

history.

Consider, for example, requirements on the cross section model of including 1

moderator history efects. At the CASMO 30 level, a base depletion is performed at a fixed moderator condition.

Staue statepoint calculations, or branch calculations, are perfomied j

fmm this depletion to other moderator conditions, ne modemtor partial cross sections that will be calculated by TABLES 3 will have been produced using consistent nuclide concentrauons (that of the base case with the excepuon of hydrogen and oxygen concentrations).

For instantaneous moderator changes this representation is adequate.

I However, since a typical PWR will deplete, with varying moderator conditions axially in the core, the instantaneous effects do not entirely address the spectmm effects of depletion at I

moderator conditions that are different than the base case moderator condluon. Dese effects result in different nuclide concentration changes and different reactivity effects with depletion.

Werefore, the need arises to include history (depletion) effects in the cross section representation associated with the SIMU1 ATE-3 model. We history efect is incorporated in the SIMULATE-3 model by performing an additional depletion at some other moderator condition.

This data is combined with the base moderator depletion and instantaneous moderator branch cases to produce history parual cross sections. A similar approach is used

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for the remaining history effects. Exposure weighted average soluble boron, moderator, void and control history variables att accumulated widi depleuon on a nodal basis to quantify the history component of the cross sections.

In addition to macroscopic cross secuan handling, the TABLES 3 code also tabularizes CASMO 30 generated microscopic cross sections for xenon, samartum and soluble boron.

I hese nuchdes are treated exphcitly in SIMULATE-3, as are the precursor nuclides, iodine and promethium.

Finally, discontinuity factors, corner point fluxes CASMO-30 pin power distribuuons, and detector reacuan rates are also functionahzed by the TABLES 3 code for I

appbcation at the SIMU1XIE 3 level.

2.4 Calculational Canabilities ne SIMULAIE 3 code is capable of calculating physics parametem typically associated with reload physics analysis for PWRs and BWRs. De code can calculate core reactMty and j

power distributions in two or three dimensions.

Modemtor Doppler, and xenon feedback I

effects can be utilized or isolated for sensitMty studies. Incore computer constants such as predicted detector signals can be generated. Previously Yankee Atomic has perfonned such calculations using the SIMU1 ATE and PDQ 7 computer codes."

2.5 Validation The amiracy of each of the five models which comprise the physics portion of SIMUIA*IE 3 have been evaluated by the code vendor though direct comparison to higher-order numerical calculations." Wese tests provide verification that each model functions properly, and equally as important, these tests provide quantification of the accuracy of each model. De benchmark tests include the following cases:

I 2 D PWR Benchmark (with and without control rods) 2 D BWR Benchmark (with and without Discontinutty Factors) 2 D BWR Benchmark versus PDQ 7 (with and without Discontinuity Factors) 3 D 1AFA PWR Benchmark 3 D PWR Benchmark versus PDQ 7 I I I

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nene benchmarks demonstrate the following versus the reference soluuon:

the two group neutmnic model, QPANDA, used in SIMULAE3 prwides assembly power distributions ranging from 1.0% maximum RMS errors for NR applicauons to 1.5% to 3.0% RMS errors for BWR applications.

I the use of discontinuity factom corrects most of die spatial homogenization errors which are traditionally introduced by the use of flux volume weighted

- I cross sections.

the CASMO-30/ SIMULATE 3 baflie/ reflector modeling techniques yield very accurate power distributions, including inside or outside baflie/ reflector corners.

the overall accuracy of the SIMULATE 3 code is comparable to the accuracy of dirret two-group PDQ 7 calculations.

The second benchmark report prwides further verification by application to a N'R for I

two cycles of depletion." The problem tests the accuracy of the nodal method in terms of depletion. his is also an integral test of the partial cross section functionalization in that it tests the validity of the partial cross section" methodology to accurately reconstruct macroscopic cross sections for a given statepoint at the SIMULA'IE 3 level.

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• lhis benchmark demonstrates; I

the two group neutronic model, Ql'ANDA, used in SIMULA'IE 3 provides accumte assembly power distributions for PWR problems.

I the macroscopic depletion model (macroscopic cross sections functionalized versus assembly exposure, including history effects) and the SIMULA*IE 3 code is comparable in accuracy to microscopic depletion models, I

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l These benchmarks estabush the vertheation of the code system. CASMO SO statepoint calculations, TABLES 3 cross secuon funcuonalizauon, and SIMULATE-3 reactMty and gross I

power distribuuon calculauons are all verthed.

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E TABIE 2.1 PWR FUNCTIONAL DEPENDENCIES I

Iggg CASMO-30 Datn 'INoe Punction I

t Fuel Cross Section Exposure Moderator History Boron History Instantaneous Moderator Instantaneous Doppler instantaneous Boron I

Control Rod Insertion Fission Product Data, Exposure l

Discontinuity Factors, Instantaneous Moderator Pin Power Distributions, instantaneous Doppler Corner Point Fluxes.

Instantaneous Boron Detector Reaction Rates Control Rod Insertion I

I Radial and Axial Reflectors Cross Section, Instantaneous Boron Discontinuity Factors Instantaneous Moderator.

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TABIE 2.2 BWR PUNCT10NAI, DEPENDENCIES l

h CASMO-30 Datn Woe Function Fuel Cross Section Exposure Vold History Control Rod History Instantaneous Void I

Instantaneous Doppler Instantaneous Moderator Control Rod Insertion I

Fission Product Data, Exposure Discontinuity Factors Void History Instantaneous Void

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l Instantaneous Doppler Control Rod Insertion I

Radial and Axial Reflectors Cross Section, Instantaneous Vold Discontinuity Factors Instantaneous Moderator I

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I 3.0 PWR VAUDNnON CAlfUIENON Yankee Atumic intends to use the SIMULATE-3 code to l'erform reactor physics analyses for operating PWR and BWRs. Bis section provides vahdation calculauons performed on two operating PWRs.

De next section will provide simila verification on BWRs.

We PWR verification consists of modehng several operating cycles of the PWR reactors using SIMU1AIE 3 in a manner similar to the way Yankee Atomic intends to model reactors for I

licensing applications.

Data gw.erated using these models are compared to actual plant measured data in order to vahdate the code for tractor physics appucations.

3.1 McGuire Unit 2. cveles 1 thmunh 3 VaMation I

The SIMULAIE 3 code was appbed to the analysis of McGuire Unit 2, a Westinghouse 4 loop. 3411 Mwth Pressurized water reactor, owned and operated by Duke Power Company."

ne analysis consisted of performing typical reactor physics calculations used in reload licensing. Dese calculations were compared to measured data encompassing three cycles of I

operation. Comparisons were made for entical boron concentrations as a funcuon of cycle exposure, detector reaction rates, assembly power distributions and axial offset as a function of exposure, axial offset dunng xenon transients, and startup test predictions consisting of reactivity coefIlcients, boron endpoints and control rod worths.

I Table 3.1 provides a brief summary of the cycle design of the three cytles analywd.

Rese cycles include the Standard and Optimized Westinghouse fuel design using dry pyrex burnable absorbers which are removed after one cycle of operation, as well as transition from a conventional out in loading pattern to a low leakage pattern featuring loading fresh fuel I

inboard.

3.1.1 Model Descriotion Single assembly CASMO-3G lattice calculations were performed for each of the fuel types resident in the core in Cycles 1 through 3. De fuel designs ranged in enrichment from 2.1 to 3.2 w/o U 235, and contained various numbers of burnable absorber pins. ne burnable poison material was borosilicate glass. CASMO 3G cross section data was processed with the TABIES 3 code for use in the SIMULAIE-3 model. I I

I ill The SIMUIAE3 model was three dimensional with four nodes per assembly radially and twelve nodes axially.

In the fuel region, while the reflector is modelled explicitly.

hermal hydrauhe (moderator) feedback and Doppler feedback were used. Fuel temperature, as a funcuon of exposure, was also used.

3.1.2 Bomn litdown Resuhn 8

Hot Full Power (HFP) depletions were performed to calculate critical baron as a funcuon of exposure for each of the three cycles. 21s data is compared to measured data in Tables 3.2 through 3.4.

As the results demonstrate, the SIMUIAE3 model achieves acceptable agreement versus the measured data. The average absolute difference in the SIMUIAE3 calculated critical boron and the plant measured data was 13 ppm with a 10 of 10 ppm for I

50 data points over the three cycles that were compared.

We end of cycle bomn concentrations were predicted to within an average absolute difference of 7 ppm, indicating accurate prediction of reacuvity depletion rate to end of cycle.

I 3.1.3 HFP Detector Calculations McGuire Unit 2, contains the standard Westinghouse flux mapping system comprised of movable fission chambers. In the core, 58 of the 193 assemblies are measured directly with these devices which traverse the central instrument tube of the assembly. Flux maps are taken appmximately every thirty days of plant operation. We HFP SIMULATE-3 model was used to calculate detector reaction rates (fission rates) to compare to the plant measured data.

A sample set of flux maps from Cycles 1 through 3 were used for comparison purposes, I

hble 3.5 presents the list of cases analyzed. Table 3.6 presents a summary of the results of reaction rate comparisons that were made.

As the table demonstrates, SIMULAE3 accurately predicts the detector reaction rates and axial offsets versus the measured data, ne average overall RMS difference over the three cycles in the predicted axially integrated reaction rates versus the plant data was 1.5% and the average absolute delta offset was 1%.

De axial offset is defined as the difference of the flux in the top half of the core and the bottom half of the core divided by the total flux. It is assumed that the detector reaction rates are proporuonal to the fluxes. Herefore, the axial offset is calculated using the detector reaction rates hitegrated axially. Figures 3.1 through 3.3 present radial distributions for each 16 -

I I

E of the three cycles analyzed, ne value at each location is the percent difference and the standard deviauon of the percent difference between the SIMU!AIE 3 predicted axially 1

integrated reacuon rate and the plant measurement integrated and averaged over the entire cycle. Sqns are included in order to determine if any radially trends exist, such as in out tilts. Figure 3.4 presents the data for all three cycles combined together, As the figures show, there are no significant radial trends with each cycle.

3.1.4 HFP Power Distributions In the previous section, the SIMUIAIE 3 capability to accurately calculate detector reaction rates was demonstrated.

In this section, the inferred measured assembly power distributions, are compared to SIMUIBIE 3 predicted assembly powers.

Assembly power l

distributions are not measured directly by the Westinghouse flux mapping system, but are j

l inferred from measured detector reaction rates.

Multiplication of the predicted assembly powers by the ratio of the measured detector reaction rate versus the predicted reaction rate l

1s performed for the instrumented assemblies. The non instrumented assembly powers are j

derted using distance weighting techniques to couple them to the instrumented locations and i

to the analytically predicted assembly power distribuuon.

De SIMUIAIE 3 values are produced directly by the model discussed previously. Table 3.7 presents the summary of the average difference between the measured and assembly powers for each of the cases presented in Table 3.5.

The overall average absolute difference in predicted and measured assembly powers over the three cycles was 1%. Again these results are quite good, and j

should be; considering the detector reaction rate comparison results.

As the results

)

1 demonstrate, there is no nouceable trend versus cycle depletion.

3.1.5 HFP Xenon Transient As part of plant operating procedures, mild xenon transients are induced at or near HFP in order to perfonn a calibration of the incore and excore detection systems. We transients are induced by the insertion and withdniwal of the argulating control bank in order to produce a range of axial offsets that span the allowable operating space, his is usually in the +5% to 15% range in axial offset.

I 17 -

I I

I De HFP SIMU1AE 3 model previously presented was used to simulate a xenon transient in order to compare predicted core average axial offsets to those measured at the plant. 21s is a diEicult calculation due to the uncertainty in the time control rode nwve to a certain position, and when the actual axial offreet is measurrd. In addition. It is essential that the plant has been operating at an equihbrium condition (no sgnificant change in offset over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

I ho plant measured xenon transients were modeled for McGuire Cycle 3.

The first transient occurred at a core exposure of 6.245 Owd/Mt or approximately at middle of cycle (MOC).

De second transient occuntd near end of cycle (EOC) at an exposure of 9.330 Owd/Mt. Each of the transients lasted appmximately 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> with the regulating bank ranging in insertion from 214 steps to 179 steps withdrawn (228 steps is fully withdrawn).

Dese contml rod maneuvers produced measured axial offsets that ranged from +3.5% to -

11.6%.

The SIMUIAE 3 model predicted axial offset quite well venus the plant measured data I

for the MOC transient. This appears to be a good benchmark case since the measured axial offset remains quite flat (indicating plant equihbrium) Just prior to the regulating control bank being moved (at 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> into the transient). Table 3.8 presents the comparison of the plant measured c. ore average axial offset for this transient and the SIMULAE 3 value, ne average absolute difference in axial offset was 0.78 with a o of 0.59. Figure 3.5 presents a plot of the plant measured core average axial offset and the SIMUIAE-3 predicted core average axial offset versus time during the tmnsient. The figure also includes the regulating control bank position during the transient, and further demonstrates the accuracy of SIMULAE 3.

The second xenon transient evaluated with SIMULATE 3 did not achieve as good lg agreement as the MOC transient, nis is due to the fact that the plant was not in true g

equihbrium when tue transient began (indicated by the measured core average axial offset fluctuating prior to the actual induced xenon transient). These results are presented in Table 3.9.

De results appear very good with the exception of the end of the transient. We average absolute difference in offset was 1,54 with a o of 1.14. Figure 3.5 also presents a plot of the plant measured core average axial offset and the SIMUIAE 3 predicted core I

average axial offset versus time during this tmnstent, ne figure also includes the regulating control bank position during the transient. As the figure shows the plant measured axial lI 18 -

I I

I ofeet is not stable at the beginning of the transient, demonstrating that the plant is not in equilibrium before the actual transient began. Ovemil, SIMLRATL3 predicts the axial offset quite well versus the plant measured data during the transient.

The average absolute delta in core average axial c" set for both transients combined was I

1.2% with a to of 1% nts is alightly higher than the results of the steady state reaction rate compartsons. However, due to the nature of conducting this measurement, the results I

are quite good and acceptable.

3.1.6 HF'P EOC Moderntor Temnetuture Coemetents Moderator temperature coemclents (MTCs) were measured near the end of each cycle at HFP. nese measurements were conducted at or near a critical boron concentration of 300 ppm by varying moderator temperature of the plant by appmximately 3*F, Dese conditions were modeled with the HFP SIMUIATE 3 model and comparisons made to the plant measured results, Table 3.10 presents the SIMUIATE 3 calculated hf!Cs versus the plant measured.

I De average absolute dtIIerence for the three measurements was 1.6 pcm/'F ne results are quite good; showing the SIMUIATE 3 models ability to predict temperature coef!)clents at HFP conditions, especially in light of the dimculty associated with taking such a measureinent.

3.1.7 HZP Startun Test Twdictions I

Startup test predictions were generated using SIMUIA'ILS at Hot Zem Power (HZP) at the beginning of each cycle.

The tests consisted of conducting control rod worth measurements via the conventional boration/dtlution technique. The control mde are B.C I.

with Ag In Cd tips. The control rods were measured in a sequential, non overlap fonnat.

During the testing rod worths are measured as well as the critical boron concentration with each bank fully inserted (also refened to as boron endpoints).

Isothermal temperature coefficients (TICS) are also measured with contml mds inserted in sevemi different configurations.

Startup Physics Tests for Cycles 2 and 3 employed a rod swap technique of measuring control rods worths, in these cycles only two contml banks were measured with the I

19 -

I I

I comentional boration/dtlution; consequently, only these banks are predleted with SIMUIA'IE-3. De rod swap technique was not modeled for this SIMULATE 3 analysis.

Tables 3.11 through 3.13 present the predicted control rod worths and compares them to measured data. The agreement is very good in Cycle I with mixed results in Cycles 2 and 3.

ne average absolute difference was 3.5% with a to of 2.7%. However, all predicuons remain I

within the 10% acceptance criteria versus the measured data. Tables 3.14 through 3.16 presents boron endpoint results for the control rod measurements conducted in Cycles 1, 2 and 3.

De average absolute difference was 21 ppm with a to of 17 ppm, and include all I

rods out measurements for each of the three cycles. Again overall agreement is quite good in Cycle 1 with acceptable results in Cycles 2 and 3.

Table 3.17 presents isothermal temperature coefilcients that were measured during Startup Testing of Cycles 1, 2 and 3.

nese coefficients were predicted quite well by the SIMUIATE 3 code with the average absolute difference being 1.2 pcm/'F with a to of 0.3 pcm/'F.

3.1.8 McGuire Unit 2 Summarv In summary, the results of the SIMUIA'IE 3 analysis demonstrate the code coupled with I

data from CASMO 30 can effectively predict the reactor physics characteristics of an operating PWR. 'Ihis translates into confldence in pedonning reload physics calculations and generating physics data for safety analysis applications ustng the SIMULATE-3 code. A varied set of validation cases were conducted that attest to this.

I 3.2 Earlev Unit 2. Oveles 1 through 4 VnMntion I

A similar validation of the SIMULATE 3 code, as was performed by Yankee on the McGuire unit, was perfonned by Studsvik of America on the W.stinghouse Farley Unit 2." A I

summary of the Studsvik results are presented in this report since Farley provides several additional validations, beyond those presented by the McGuire analysis.

Farley is a three loop,157 assembly, 2652 Mwth, Westinghouse unit. The core design through four cycles of operation consists of standard Westinghouse 17x17 fuel rod array assemblies (same as McGuire) spanning in enrichment from 2.1 to 3.45 w/o U 235. Table 3.18 presents a brief summary of the four cycles analyzed.

As with McGuire, the fuel I I

E contains a varying number of burnable poison pins. The burnable poison is bomatiscate glass, ne analysis of Parley adds to the SIMULA1E 3 PWR validation by including several characteristics which are different from McGutre. Dese additions are:

Slightly higher enrichtnents Control rod material is full length Ag In Cd Cycles 3 and 4 were 18 month Cycles with extremely high soluble boron I

concentrations at beginning of cycle Burnable poison was insened into depleted fuel HZP power distributions were predicted and compared to measured data Rod Swap Startup Test Predictions were performed with SIMULATE 3 I

3.2.1 Model DescriotioD I

The SIMU1 ATE 3 model constructed for Farley was virtually identical to the model setup j

i for McGuire.

"u only distinguishable difference in the model was the treatment of inserting l

burnable ;r. t ains into depleted fuel. nas was required since fresh burnable poisons were l

inserted iro a pleted fuel assemblies in Cycle 3.

Ilowever, no additional CASMO 30 boron cases were required to deal with the high boron concentrauons at beginning of Cples 3 and 4.

3.2.2 Bomn letdown Results Reference 16 contains detailed tables of data for the critical boron concentrations versus plant measured data for four cycles of operation. For the total of 48 comparison points, the

]

average absolute delta in ppm of SIMULATE 3 venus the plant measurement was 12 ppm I

with a to of 9 ppm. No trends with tycle or cycle exposure were evident. We end of cytle

"**"""""*'"'"**""*'""**"""d"*"'""'"**"'"'"'

E t

y being 8 ppm. Bis indicates an accurate prediction of the reactivity depletion rate to end of cycle.

We results are quite similar in accuracy to those witnessed in the McGuire SIMUIATE 3 validation analysis.

l ilI lII 21 -

'I LI

1 I

3.2.3 HFP Detector Cakuhunna Comparisons between SIMULAE 3 generated incore detector reaction rates and plant meaured data was performed on four cycles of Farley, Farley contains the standard Wesunghouse flux mapping system. As was done in the McGuire analysis, SIMU1EIE 3 predicted reaction rates were compared against plant measured results.

Table 3.22 sununarizes the results of this analysis. A total of 15 flux rnaps were used for comparison I

purposes. De average RMS differences in reaction rates was near 1% in at cycles. Ne noticeable trends exist with cycle or cycle exposure, ne detailed asse.noly to assembly results also Irveal no noticeable trend with assembly type or location m the core (periphery vs. inboard).n The average absolute difference in axial offset was 1%.

I 3.2.4 HFP Power Distribull2Il3 I

The assembly power distribution comparisons for Farley yield similar results to the McGuire analysis.

No discemable pattem in the results is seen, with the exception of I

beginning of each reload cycle which exhibits slightly higher difierences than the rest of the cycle. De results overall are very good and, again, quite similar to the McGuire results. De average di!Terence was 1%. The statistical sumtnary versus the measured data can also be found in Table 3.22.

I 3.2.5 HZP Power Distribution 3 I

As part of the Farley benchmark, comparisons were made between SIMUIATE 3 and measured data for power distributions at Hot Zero Power (HZP), beginning of cycle. D ese comparisons were conducted for Cycles 1 and 2.

SIMUIATE 3 calculated detector reaction rates, axial offset and assembly relative powers were compared to plant measured data.

Tables 3.20 and 3.21 summartze the results. The percent differences are larger than those witnessed at HFP but are expected since the uncertainty of the measurements is higher. De average absolute difference was 2.7% with a o of 2%. However, the SIMULWIE 3 results are very good, particularly for Cycle 2 which exhibits such a large positive axial offset.

I I

I 22 -

I I

..I 3.2.6 HZP Startuo Test Predictions The Startup Test Predictions for Farley are interesting in that the rod swap technique was used in Cycles 3 and 4 (boration/ dilution was used for Cycles 1 and 2) and the control rod material is Ag In Cd. De results for control rod worths are presented in Tables 3.22 through 3.25 for each cycle, respectively. Cycles 1 and 2 used the com'entional boration/ dilution measurement technique and Cycles 3 and 4 used a rod swap technique of I

measurement. ne results for all cycles are very good with the average absolute percent difference versus measured data being 3.2% with a o of 2.5%. Again no discernable trends I

with cycle were apparent.

Referrnce 16 also includes Isothermal Temperature Coefficient comparisons conducted during HZP testing.

Over the four cycles analyzed the average absolute difTerence between the plant measurrd and SIMULAIE 3 predicted was 0.9 pcm/'F with a to of 0.3 pcm/'F.

I 3.2.7 Parlev Unit 2 Summary i

in summary, the results of the SIMUIATE 3 analysis demonstrate that the code can accurately predict plant operating characteristics.

This Farley analysis coupled with the I

McGuire results show the code to be quite accurate for performing reactor physics calculations. The analysis conducted spanned a large array of calculations and demonstrated the code capabilities versus measured plant data for typical operating PWRs.

1 I

' I E

I 23 -

I I

I I

i TABLE 3.1 McGUIRE UNIT 2 OPERATING CHARACTERISTICS CHARACTERIST1C CYCLE 1 CYCLE 2 CYCLE 3 ENRICHMENTS 2.1/2.6/3.1 2.6/3.1/3.2 3.1/3.2/3.2 i

FUEL DESIGN STANDARD STANDARD /OFA STANDARD /OFA

1. OF BP'S 1520 64 352 BOC HZP BORON 1295 PPM 1413 PPM 1379 PPM LOADING SCHEME OUT/IN OUT/IN TRANSITION i

l CYCLE LENGTil 14.6 Gwd/Mt 9.0 Gwd/Mt 10.8 Gwd/Mt l

I TABLE 3.2 McGUIRE UNIT 2 CYCLE 1 BORON LETDOWN CYCLE EXPOSURE MEASURED SIMULATE 3 DIFFERENCE (Gwd/ML)

(PPM)

(PPM)

(PPM)

I 1.070 846 816 30 1.865 836 808 26 2.119 837 803 34 3.142 794 763 31 I

3.872 758 729 29 4.929 704 674 30 5.963 652 616 36 I

7.075 582 550 32 7.962 520 495 25 8.737 461 445 16 9.257 420 410 19

, I 9.818 389 371 18 l

10.517 337 321 16 11.248 284 260 15 I

13.327 125 111 14 14.261 38 38 0

14.582 11 13 2

l l

I 24 -

I I

1 I

i TABLE 3.3 McGUIRE UNIT 2 CYCLE 2 BORON LETDOWN I

CYCLE EXPOSURE MEASURED SIMULATE 3 DIFTERENCE (Owd/Mt)

(PPM)

(PPM)

(PPM) 0.975 857 865 8

0.131 851 849 2

1.250 840 838 2

I l.407 835 823 12 2.112 700 755 5

i 2.559 723 713 10 t

I 2.810 695 600 5

3.270 054 647 7

4.215 571 558 13 4.759 522 507 15 I

5.263 482 461 21 6.086 389 385 4

6.717 334 329 5

I 7.410 261 267 0

8.438 100 177 8

i 9.827 48 59 11 I

TABLE 3.4 McOUIRE UNIT 2 CYCLE 3 BORON LETDOWN CYCLE EXPOSURE MEASURED SIMULATE 3 DIFFERENCE (Gwd/M()

(PPM)

(PPM)

(PPM)

.232 928 953 25 l

.478 900 926 26

.718 880 907 27 I

.915 881 891 10 1.386 846 854 8

1.887 814 816 2

2.007 808 806 2

I 2.046 799 803 4

2.208 794 790 4

2.475 772 768 4

I 3.003 726 725 1

4.I32 641 630

.I1 4.890 570 566

-13 5.743 503 496 7

I 6,914 410 400

-10 7.740 338 333 5

7.943' 318 317 1

I

• Cycle 3 had not completed operation wheii this analysis was conducted I

25 -

I I

J I

TABLE 3.5 MeOUIRE UNIT 2 CYCLES 1 THROUGH 3 I

DETEC1DR COMPAR1 SON CASE DESQRil'110N EXPOSURE POWER LEVEL BANK D STEPS CYCLE (Gwd/Mt)

(% PULL POWER)

WITHDRAWN

• 1 1.221 88.48 185 I

3.417 89.74 203 I

5.470 100.00 214 7.577 100.00 218 I

9.258 100.00 215 12.370 100.00 216 14.270 100.00 220 1

2 0.981 100.00 214 1.448 100.00 214 2.737 100.00 210 I

4.218 100.00 214 0.723 100.00 211 9.171 100.00 214 3

1.052 100.00 214 3.020 100.00 211 4.909 100.00 212 I

5.787 100.00 217 7.951 100.00 214 9.029 100.00 212 I

• 228 steps is fully withdrawn li La

'I LI I

26 -

I

t I

TABLE 3.6 I

McCUIRE UNIT 2 CYCLES 1 THROUGH 3 COMPARISON OF PREDICTED AND MEASURED AXIALLY INTEGRATED AVERAGE REACTION RATES AND AX1AL OPPSETS REACT 10N RATE EXPOSURE RMS OF AVERAGE AX1AL CYCI.E (Owd/Mt)

DIFFERENCES (%)

DELTA OFTSET (%)

- I 1

1.221 1.181 1.5 I

3.417 1.401 1.5 5.470 1.I80 1.2 7.577 1.768 1.6 9.258 1.476 1.0 12.370 1.268 0.7 14.270 1.(34 0.8 2

0.981 2.239 1.2 1.448 2.186 1.7 2.737 1.602 0.6 4.218 1.446 0.5 I

6.723 1.182 0.6 0.171 1.173 0.5 t

3 1.052 1.842 1.1 3.026 1.598 1.2 4.909 1.768 0.9 5.787 1.328 1.2 7.951 1.547 1.0 9.029 1.445 0.7 I

I

! I I

I 27 -

I

-w

,,__..__m

I TABLE 3.7 I

MeOUIRE UNIT 2 CYCLES 1 THROUGH 3 l

COMPARISON OF PREDICTED AND MEASUREQ

[

ASSEMBLY POWERS I

ASSEMBLY POWER EXPOSURE AVERAGE CYCLE (Owd/Mt)

DIFFERENCES (%)

i 1

1.221 0.790 3.417 0.870 I

5.470 0.000 7.577 1.042 9.258 0.910 12.379 0.767 14.270 0.770 i

2 0.981 1.598 1.448 1.525 2.737 1.016 4.218 0.985 3

0.723 0.793

,. g 9.171 0.752 l

3 1.052 1.136 I

3.020 0.998 4.909 0.970 5.787 0.843 7.951 0.981 9.029 0.951 I

I 28 -

i j

I TABLE 3.8 i

McGUIRE UNIT 2 MOC 3 XENON TRANSJEh*I AXIA1, OPPSET (%)

l

'n M E BANK D SIMUI. ATE 3 MEASURED S3 - MEAS (HOURS)

POSITION' O

211 4.0 4.2 0.2 0

211 4.0 4.1 0.1 12 211 4.1 4.3 0.2 1

19 211 4.1 4.4 0.3 20 207 5.3 5.2 0.1 i

21 201 7.5 7.3 0.2 I

22 197 10.3 9.4 0.9 23 190

-11.8 11.2 0.0 i

24 199 12.9 11.4 1.5 25 202 12.8 11.0 1.2 20 203 12.7 11.5 1.2 27 203 12.2

-11.2 1.0 28 201

-12.1 11.0 1.1 29 204 10.0 0.4 0.0 30 209

-7.0 0.1 0.9 31 211 3.0 3.5 0.1 32 205 1.7 2.1 0.4 33 211 2.2 1.4 0.8 34 200 4.7 3.1 1.0 35 194 2.0 0.8 1.8 I

30 182 1.0 3.4 1.9 37 181 4.5 0.4 19 38 181 7.9 9.0 1.1 39 184 10.2 11.1 0.9 I

40 214 4.7 5.0 0.3 41 213 3.9 3.8 0.1 42 210 3.8 4.0

-0.2 AVERAGE DELTA AO IS 0.78 to 0.59

' I

= 2 - > ~~*-

I I

29 -

I

I TABLE 3.9 I

McGUIRE UNIT 2 EOC 3 XENON TRANSIENT AX1AL OFFSET (%)

TIME BANK D SIMULATE 3 MEASURED S3 MEAS (HOURS)

POSIT 10N' I

O 211

-4.6

-3.9 0.7 6

211 4.7 4.3 0.4 12 212 4.0 4.5 0.5 19 212

-4.0

-4.0 0.0 20 208 5.3 4.7 0.6 21 205 5.6 6.3 0.7 22 200 8,5 8.7 0.2 I

23 202 9.5 9.6 0.1 24 200 11.6 11.3 0.3 1

25 204

-11.5 11.2 0.3 26 205 11.5 11.2 0.3 27 205 11.4 11.0 0.4 28 203 11.0 10.8 0.8 29 211 0.0

-8.0 1.9 I

30 211

-8.2

- 6. 0 2.2 31 211 0.0 3.7 2.3 32 211 3.4 1.2 2.2 33 211 0.5 1.3 1.8 34 211 2.4 3.5 1,1 35 187 3.0 1.7 1.9 36 179 7.9 6.2 1.7 i

I 37 179 11.5 9.0 2.5 38 208 0.3 3.7 2.6 39 205 5.0 3.5 1.5 i

I 40 204

~4.2

-4.5 0.3 41 205 2.6 4.1 1.5 42 206 2.4 4.0 1.6 43 206 1.8 4.0 2,2 I

44 200 1.3 4.1 2.8 45 204 1.0 4.3 3.3 46 206

-0.9 4.5 3.6 I

47 207 0.7

-4.3

-3.6 48 208 0.7 4.3 3.6 49 208 1.2 4.2 3.0 AVERAGE DELTA AO IS 1,54 to 1.14

• 228 steps is fully withdrawn I --

.-~.~

E TABLE 3.10 McGUIRE UNIT 2 EOC HFP MODERATOR TEMPERATURE COEF.

CYCLE EXPOSURE MEASURED PREDICTED DIFFERENCE (Gwd/Mt)

(PCM/'F)

(PCM/'F)

(M P)

.s.

I 11.14

-15.5 19.4 3.9 2

7.21 23.8 24.0 0.2 5

7.78

-22.3 21.7

-0.6 l

TABLE 3.11 McGUIRE UNIT 2 CYCLE 1 HZP CONTROL ROD WORTHS I

BANK MEASURED PREDICTED DIFFERENCE (PCM)

(PCM)

(%)

I, D

664 671 1.0 DC 1283 1296 1.0 DCB 1105 1085 1.8 DCBA 678 697 2.7 I

D-A SE 853 881 3.2 D-A SE SD 771 788 2.2 D A SE SD SC 1026 1077 4.7 I

TABLE 3.12 McGUIRE UNIT 2 CYCLE 2 HZP CONTROL ROD WORTHS BANK MEASURED PREDICTED DIFFERENCE (PCM)

(PCM)

(%)

D 665 658

- 1.1 C

871 932 6.6 I

TABLE 3.13 McGUIRE UN(T 2 CYCLE 3 HZP CONTROL ROD WORTHS BANK MEASURED PREDICTED DIFFERENCE (PCM)

(PCM)

(%)

D 556 580 4.1 C-787 873 9.8 i

LI L

31 -

g

I TABLE 3.14 MeOUIRE UNIT 2 CYCLE 1 HZP BORON ENDPOINTS BANK MEASURED PREDICTED DIFFERENCE I

(PPM)

(PPM)

(Pred - Meas)

ARO 1295 1202 3

l' D

1217 1227 10 DC 1097 1099 2

DCB 997 995 2

DCBA 938 928 10 I

D A SE 800 841 19 D A SE SD 791 764 27 D A SE SD SC 094 601 33 l'

TABLE 3.15 McOUIRE UNIT 2 CYCLE 2 HZP BORON ENDPOINTS BANK MEASURED PREDICTED DIFFERENCE (PPM)

(PPM)

(Pred - Meas)

I ARO 1413 1432 19 D

1333 1359 26

-I C

1318 1329 11 TABLE 3.16

_ McOUIRE UNIT 2 CYCLE 3 HZP BORON ENDPOINTS BANK MEASURED PREDICTED DIFFERENCE (PPM)

(PPM)

(Pred Meas)

I

/.3 1379 1409 30 1302 1347 45 t

-1250 1315 50 I

TABLE 3.17 I

McOUIRE UNIT 2 ligEJOC ISOTHERMAL TEMPERATURE COEFFICIENTS BANK CYCLE MEASURED PREDICTED DIFFERENCE-I (PCM/'F)

(PCM/'F)

(Pred - Meas)

I..

ARO 1

-1.41

-2.48 1.07 D

2.73

-3.78 1.05 DC

-0.07

-7.54 1.47 ARO 2

-1.73 3.12 1.39 ARO 3

0.55

-1.32 0.77 I I I;

~

I I

1-(:i l TABLE 3.18 FARLEY UNIT 2 OPERATINO CHARACTERISTICS W

cnunertaisnc CYCM 1 WCE.2 CYcu 3 CYCt.E 4 l

ENRICIIMttG Ll/2.0/3.1 2.6/3.1/3.1 3.1/3.1/3.4 3.1/3.4/3.45

1. of Wra 1074 0

704 432 BOC IIZP DORON 1313 PPM 1387 PPM 1605 PPM 1910 PPM i-l-g thADING SCllEME OUT/IN OW/lN TRANSmON IN/OLTT l3 CYel.E t.ENGT11 15.4 Owo/MT 10.4 OwD/MT 14.6 OWD/MT 15.2 OwD/Mr I

TABLE 3,19 II FARLEY UNIT 2 CYCLES 1 THRU 4 COMPARISON OF PREDICTED AND MEASURED REACT 10N RATES.

l' AX1AL OFFSETS. AND ASSEMBLY POWERS l

l REACTION RATE AX1AL ASSEMBLY EXPOSURE RMS OF OFFSET POWERS CYCLE (Gwd/Mt)

DIFFERENCES (%)

DELTA (%)

RMS (%)

1 1.522 1.04 0.5 1.01 l

5.556 0.98 1.1 0.97 10.071 0.77 0.9 0.81 14.278 0.74 0.7 0.78 2

1.117 1.21

-1.3 1.22 6.511 0.94

-1.5 0.93 9.519 1.13

-1.9 1.13 i'

3 2.343 0.94

-0.1 0fi4 6.335 0.81

-1.7 0.80 10.425 1.01

-0.7 1.00 13.854 0.98

-0.7 0.98 4

1.048 1.40 0.1 1.38

3 6.111 0.80

-1.4 0.80

--g 10.113 0.89

-1.0 0.87 i

12.123 0.97

-1.8 0.96 I I

I I

TABLE 3.20 FARLEY UNrr 2 HZP CORE AVERAGE AX1AL DETECIDR REACI1ON RATES CYCLE 1 CYCLE 2 I

NODE MEAS SIM 3 DIFF NODE MEAS SIM 3 DIFF 12 0.241 0.291 0.050 12 1.166 1.120 -0.044 I

11 0.593 0.636 0.043 11 1.812 1.832 0.018 10 0.976 0.962 0.014 10 1.875 1.907 0.033 9

1.241 1.218 0.023 9

1.667 1.673 0.007 8

1.425 1.381 0.044 8

1.368 1.365 0.005 I.

7 1.570 1.513 -0.057 7

1.139 1.127 -0.010 6

1.516 1.472 -0.045 6

0.868 0.856 -0.014 5

1.442 1,415 0.027 5

0.689 0.680 -0.009 I

4 1.235 1,209 0.026 4

0.522 0.513 -0.009 3

0.938 0.954 0.016 3

0.405 0.405 0.001 2

0.608 0.654 0.046 2

0.316 0.327 0.012 1

0.214 0.295 0.081 1

0.173 0.194 0.021 AO 0.8 0.0 RMS=.04 A0 49.7 49.7 RMS=0.02 I

I TABLE 3.21 FARLEY UNrr 2 CYCLES 1 AND 2 I

COMPARISON OF PREDICIED AND MEASURED RFACTION RA1TS.

AXIAL OFFSETS. AND ASSEMBLY POWERS AT HZP. BOC REACITON RATE AX1AL

' ASSEMBLY BANK D RMS OF OFFSET POWERS CYCLE STEPS WDRWN DIFFERENCES (%)

DELTA (%)

RMS (%)

I.

1 224 2.64

-0.8 2.68 2

213 2.84 0.0 2.86 I

I I

34 -

L I

I s

TABLE 3,22 j

FARLEY UNrr 2 CYCLE 1 HZP COh' TROL ROD WORMS BANK MFASURED PREDICTED DIFFERENCE (PCM)

(PCM)

(%)

D 1430 1406 1.71 DC 1224 1191 2.77

- I.

DCB 1967 1961 0.31 DCBA 1288 1348 4,45 TABLE 3.23 FARLEY UNTP 2 CYCLE 2 HZP COMOL ROD WORTHS j

~

BANK MEASURED PREDICTED DIFFERENCE (PCM)

(PCM)

(%)

D 1048 1114 5,92 DC 1076 1106 2.71 DCD 1506 1543 2.40 DCBA 1553 1600 2,94 i

TABLE 3.24 FARLEY UNTP 2 CYCLE 3 HZP COMOL ROD WORMS USING ROD SWAP 'IECHN10UE L

BANK.

MEASURED PREDICTED DIFFERENCE (PCM)

(PCM)

(%)

I D

1035 1105 6.33 C

635 704 9.80 B

1403 1444 2,84 i

A-759 769 1,30 TABLE 3,25

,I FARLEY UNTP 2 CYCLE 4 HZP COMOL ROD WORIES USING ROD SWAP ' TECHNIQUE l

BANK MEASURED PREDICTED DIFFERENCE I'

(PCM)

(PCM)

(%)

D 900 993 3.32 C

886 910 2,64 B

1224 1237 1.05 A

608 608 0.00 I:

I

I F10URE 3.1 McGUIRE UNTP 2 CYCLE 1

SUMMARY

OF REAC'I1ON RATE COMPARISONS 1

2 3

4 5

6 7

8 9

10 11 12 13 14 15 0.53 ~~ --- - 1.50 ----

I 1.17 --- --- 0.49 ---

1

~ ~ - --

2 3.15 ~ ~ --- --- ----- 0.2 9 --- --- ~ ~ ---- ---

0.67 ----

1.65 ---

~ ~ - - ~ ~

~ ~ - - - -

0.15 ----- 0.04 -- --

1.04 ---- 2.64 3

-- - ~ ~ -

-- - - - - ~ ~

0,41...-

1.03 ---

1,51 ---- 0.90

-0.06 -----

4

-0.73 0.36 ----

0.1 5 ---- -- - ---

0.72 1.53 -----

I 0.10 - -- ---

--- 0.39 - -

0.4 5 --- 0.39 5

0.52 - - 0.52 --- 0.37 -----

0.18 ~~

6 2.82 -- -0.51 -- --- 0.17 --- 0.69 - - --- ---- - -- -

1.11 -----

I 0.80 --- 0.66 -- ---- 0.86 - -

0. 93 --- -

---

0.82 - ---

0.10 --- ----- 0.4 5 ---- -----

0.2 4 -----

-0.27 ---

7 0.90 --- - -- 0.40 --

-- - 0.56 --- ---

0.46 ~~

8 0.94 -- 0.10 ---- 0.56 ---- 0.35 - - -- 0.86 ----- 0.20 0.11 0.52 ---

0.85 ---- 0.59 0.93 0.34 -----

1.28 -- 0.71 -- - 0.43 ---- 0.67 - -

0. 66 - -- -- -

- - ----- ----- -- 0.46 - - - 0.56 --- -- ----- 0.05 g.

1.29 --- 0.99 -- - ---- ----- 0.83 0.48 -- --- - - -- -

l

.. ~

.E.

10 0.37----

-0.38 -- - - --- ----- -0.13 ---- --- ---

- - - - - - - - - ~ ~ - - -

1 3 0.49--------------

0.53 ---- 0.54 ----

i 11 1.01 -- -- --- 0.03 ---- --- --- -2,21

- ~ ~ ~ ~ - - -

0.64 ----- --- 0.44 ---- ----- -----

1.22 l

12 1.05 -- ----- 0.34

-- -- - - 1.17 --

1,19 ---.-....- 0.39 ---- -- -- 0.66 --- ---

1 L

0.27 ---- -- - -- -- - ---- ---

-3. 4 2 IS

-- -0.07 ---- 0.07 - --

1.83

0. 63 ----- ----- -----

0.53 --- 0.51 - --

L 14

2. 50 ----- ---- ---- 0.73 ----- ---- -0.64 --- 0.49 -- - Average DE.(%)

1.56 -

1.19 --- - ----- 0.69 ----- 0.95 ~~ Stand. Dev.

l 15

-0. 61 -- -- ----- -2.17 ----- --- ----- -----

0. 68 -----

0.67 --

l l

% Average DEerence = (SIMULATE 3 - Measured)/(SIMUIATE-3)

• 100%

1I I I

I I-FIGURE 3.2 McOUIRE UhTP 2 CYCLE 2 1

SUMMARY

OF REACIlON RNIE COMPARISONS I

1 2

3 4

5 6

7 8

9 10 11 12 13 14 15

--. 0.47 -.

0.12 - -

0. 56 ---- -- --

1.25 --

1 I'

r 2

3.32 ---- ----

1.03 ----- 0.72 ~ ~ -- - - ----- ---

2.17 -- - ----- 0.81 -- --

0.4 3 ~ ~ ----- --- --- ---

3 2.08 - ---

0.29---

1.07-- - 2.20

- - - ~ ~ - - - - - ~ ~ - -

1,44 -.-

0,36..---

0.38 - ---

1.50 I

4 1.46 -0.17 -- --

0.53 --- -- --

j

--- ---- ~ ~

1.92 1.08 --- --- --- ---- 0.87 ---

5 0.08 ~ ~ --- - --

1.05 ~~

1.22 --- -0.04 -- '-

l 0.4 2 ~ ~ - - ---

0.69 -- -

1.22 - - 0.06 ---- -----

l 6

0.68 ---- 0.61 --- - -- 0.27 -- - 0.04 -----

0.54----.

1.15 - --

1.24 - -- 0.40 ---- ---- 0.82 --- 0.82 - ---

0.1 1 ---

~ ~ 0.90 --- ---- 2.86 - -

1.00 -- - -- --

7 0.61 -- ---- 0.97 -----

1.02 --- -- -- 0.65 ---- -----

8 0.21 ---

1.83 -----

1.17 --- - 0.00 - - -- --

1.09 - - -----

1,46 -0.14 --- -

0.78 -- -

1.09 - - 0.97 ~~

1.12 ----- ----- 0.92 - -- ---

1.27 0.23 ----

l 3 9

-.... 0.37 - --- - - -

0.43 ---- 0.88 ---

---

- 1.68

.. 3 1.07 ----

0.78 ---- 0.61 --- -

1.26 i

10 1,77 --- 0.32 --- - ----- ----- - -

-0. 59 ---- - --

L I-0.82 -----

1.01 ----- --- -

0.2 7 ----- -- -- ----

l 11

-2. 55 - --- --- --- -0.19 --

--- - 0.95 ----- ----- 0.19 -----...-- ----

2.78

, =

1.2 5 -- --- ---- 0.54 ----- ----- 0.78 --- ----

0.2 4 -----

2.20 l

l 12 0.37-----

0.9 5 ----- ----- - 1.36 --- - - ---

- - - - - ~ ~ - - -

0.52 - ---

0.80 ----- - -- 0.68 --- - -- -

o 13

-0.01 -- - 0.65 --

---

2.16 ---- ---- - --- - -- - --- 3.03 i

0.07..... 0,41 --.-

0. 4 9 ---.. --- --.. --..-. --

1.83 I

14

-0. 48 ---- ----- ----- 0.49 - -- - --- - 1,38 --

3.98 ----- Average DifT.(%)

0.99 --- ----- -- -- 0.62 ----- ----- 0.94 ----

1.36 ----- Stand. Dev.

15

- 1.60 ---- - --- 0.62 -----

1'I-1,07..-.- -.-

1.13 ----- ----- ---- ----

% Average Difference = (SIMULATE Measured)/(SIMULA'IE-3)

• 100%

,I

. LI 1

1 1

m.

a m

m

I FIGUR10 3.3 McGUIRE UNrr 2 CYCLE 3

SUMMARY

OF REACHON RATE COMPARISONS I

1 2

3 4

5 6

7 8

9 10 11 12 13 14 15 1.90 -

1.18 2.38 --

---

0.36 ---

1 2

- 1.06 -- ---

1.05 - - 0.65 - - --- --- -- ----

0.34 - - 0.82 -

2,04 -.-

0.87 --

-0.70 - --- 0.36 - ---

1.39 3

0.44 -- -- 0.86 -- -- 0.51 ----

1.97 I

1,61-----------------------

1.08 1.24 ----- ----

4 0.91 0.68 --- ~~ --

1. 62 -- -- - - -- ---- ----

0.17 - - --- ---

1.85 --- 0.11 ---

1.44 -- ----

5 I

-- --- ~~

0.66 -- ~ ~ ----

1.30 --- 0.55 --- 0.48 --- - -- -

6 1.19 ----- - 1.4 5 ----- -----

0. 81 ---

0. 5 8 -- - --- ----

1.07 -----

0.91 ---

0.30 ---- -- -- 0.83 - -

0.4 2 --- - ---- -- ---- ----- 0.40 ----

-0.28 -

- - 0.32 -- - ----- 2.79 --- ----- 0.39 --- ---

7 0.73...~..-.. 0.56 ~~ - - -- 0.77 ---- --- 0.25 -- -----

.l 5

8 2.17 ----

-0.44 - -

1.33 - - 0.68 --- ~~ --

---

-0.87 0.60 0.81 --

2.23 - -- 0.51 - -

1.04 ---- 0.73 -- ---- --- - -- 0.30 0.68 1,40- ---

I 1.50

0. 50 -....~.. -......... ~..... -

.0. 62 ----- ---- - -- -- --

g 0.59 0.53 ----- - -- ---

0,44.-- -..

0.47 ----- ---- ---- -- -- - 1.0 5 ---- ----- ----

am 10 g

0.83 -----

0.68 ---

u 0.12 ----- --- -.-- - 0.29 -.

- -- 0.33 - - -- - 1.16 -.--- ---- -- --

1.55 1.50 -- --- --- 0.60 -- ----- 0.76 ----- -----

0.7 5 ----- ----- ----- 2.37 12

--- - ----- -- - ---- 0.16 ---- -- -- 2,2 8 ---- ----- -0.12 -----

0.49 ---- -----

0.60 -- - -- -- 0.82 -----

0.15 ----- -- -- --- -- - ----

-4.21

-0.11 - --

1.46 -----

13 1.20 0.88 ----- 0.43 --- --- 0.54 --

8 14 0.61 -- -

--- --- 0.48 ---- --- -0.58 ----- -1.22 --- - Average Diff.(%)

1.37 -----

1.74 ----- Stand. Dev.

1.4 3 ----- - - ----- 0.37 -- -

15

- 1.16 --- - ---

1. 83 ----- - --- ---- -----

I 0.68 --

---

0.56 --- ----- -- -- --

% Average Difference = (SIMULATE Measured)/(SIMU1AIE-3)

• 100%

I I I I

I I.

FIGURE 3.4 McGUIRE UhTP 2 CYCLES 1 'IMROUGH 3

SUMMARY

OF REACmON RATE COMPARISONS B

1 2

3 4

5 6

7 8

9 10 11 12 13 14 15' 0.85.~.

.o.22.-...

1.73 --- - ---

1.35 ---

. I.

1 0.48 ~~

0. 5 7 -- -- - -- -- -- - --

2 2.54. --

0.78 --- 0.64 ~ ~ ----

2.11 - -

1.03 ----- -0.06 - - 0.83 - 2,42 3

1,19.-.- 0,79 -.- 0,99 -..

1,08 I

0.3 7 - --- - ---

0.39 0.47 --

4

- -- ~ ~ - - - - - - -

1.60 1.25 -- -- ----- - --

1.11 ---

- - - - - - - - ~ ~ - -

0.23 - -- -

1.06 - --- 0.68 ---- 0.32 ----

5 I

0.39 ---- - -

1.04 --- 0.85 - --- 0.60 --- -- -

-0.25 - -

0.4 5 ---

- - - - --- 0.92 ---

6

-0.99 - --

-0.87 ~~

1.92 ---- 0.66 --

---

0.89 -----

0.7 8 ----- --- -- - -- ~ ~ 0.84 - -

I.

0.06 - -- --- - 0.79 -- ---

-0.18 ~~--

0.27----

7 0.93 -- --- - 2.41 -- --

0.72 -----

0.57 ---

8 0.41 - -- 0.48 - - 0.89 ----- -0.50 -- - --- 0.81 ----- 0.04 0.47 0.32----

1.96 -----

1.19 --- 0.84 ----- 0.95 -- --- 0.97 ---- 0.46 1.03 0.46----

.I 0.08 --- 0.62 ----- --- ----- 0.90 0.34 ----

9 1.19 0.80 ----- -

--- --- ~ ~ --- -

1.04 --- 0.84 ----- -----

0.88 ---- 0.1 1 --- ---- ----- -- ---

0.57-----

10 I

0.93 -..-

0. 81 - - ----- -- - ----- 0.67 ---

11 1.23 ---- ---- --- -0.04 -- ---- 0.72 ---- ----

-0.2 8 ----- ---- ----. 2.38 '

0.80 - --- ---- 0.77 --- - ----

1.57 0.34 ----

1.60 ---

0.53 ---- -- -- 0.53 - - - -- -0.99 --- - ---

12 1.40 - -- --- 0.7 9 ----- -----

0.92 -- --

13

- ----- -0.08 ----- 0.69 ----- -- -- 0.63 - --- ---

-3. 53 o,51--.-.

0,73...-.....

1,23...-...................-

1.65 I.

14

-0. 88 --- - ----- ----- 0.71 - --- ---- - 1.02 ---- - 1.82 ---- Average Diff.(%)

1.86 -- - ----- ---- 0.84 ----- ----- 0.91 --- -

2.00----

Stand. Dev.

0.03 ---- ----- ----- ---

15

- 1.04 -----

I 0.87 - --- -----

1.2 2 ---- --- - -- -----

% Average Difference = (SIMU1 ATE 3 - Measured)/(SIMUIATE 3)

• 100%

I I I

I a-FIGURE 3.5 g

McGUIRE UNIT 2 CYCLE 3 XENON TRANSIENT I

a..

MOC TRANSIEh7

.43.

i..

,,oo I.

A is.

$ao x

j lb-B

^

LEGEND

...o I

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N F

• = BANK D POSmON L

M = MEASURED DATA D

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X

.. iso p

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\\

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F a

n c.

c.

I I

r T

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40 I.-

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-la-

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16 h k b b io $3 is is 1% $0 $3 $4 de Ya do 33

$+

de de 40 43 44 TIME (HOURS)

EOC TRANSIENT 16

.42o I

16-

.aco i..

A 15-I

-y

. 1e0 g

to-1.EGEND

"' A B'

A L

I

{..

S = SIMULATE.3 DATA 6-

• = BANK D POSmON N

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.i o M = MEASMD DATA 0

U 3.

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O i -*-

.i oo s F

. frann sa n n n e c c e e.e. e e.. c e c e 1

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('o (3 i+ is is t'o iz 1'4 fe f'e d'o 32 34 d'e s'e do 42 44 4e t'o so' a

-r TIME (HOURS)

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_ ~ _ _... _

4.0 SWR VAIJDA'nON CALCULA' HON nis section pmvides validation of the SIMULATE-3 code for application on Boihng Water Reactors (BWRs). De vahdation consists of modehng several cycles of two operating BWRs in a manner consistent with Yankee's intended application. De models are used to generate l ~

eigenvalue and Traversing incom Pmbe (HP) trace data as a function of cycle exposure his

g data is then compared to plant measured data, g

i 4.1 Quad Cities 1 Oveles 1 and 2 Validation The SIMULA*IE 3 code was applied to the analysis of the Quad Cities 1 reactor.

Quad Cities 1 is considered a typical BWR of GE design and has been pnviously benchmarked at YAEC using SIMUIATE 2.

For the validation of SIMULATE-3, the first two cycles of operation were modelled. The analysis consisted of comparing SIMUIATE 3 results to plant measured data for the following:

hot eigenvalues as a function of exposure, I

l local cold criticals at the beginning of Cycle 1, cold in sequence criticals during Cycles I and 2, and TIP traces during Cycles 1 and 2.

I 4.1.1 Quad Cities 1 Model Descriotion l I Single assembly lattice calculations wen performed for each of the fuel designs resident j

in Cycles 1 or 2 of Quad Cities. De fuel designs contained individual pins that ranged in

, g 3

U 235 enrichment fmm 1,7 w/o to 2.73 w/o. Some of the fuel designs contained Gd,0,3 loadings ranging from 0.5 to 3.0 w/o. Cycle 2 contained several MO, assemblies, which were also modeled. Both 7x7 and 8x8 lattices wem employed. De pellets were both dished and undished, resulting in diffennt fuel densities. ne gadounia pin spectrum calculations were l

performed using the MICBURN 3 code, for use in CASMO-3G.*' he CASMO-3G cross section data was processed with the TABLES-3 code for use in the SIMUIA'IE-3 model.

The SIMUIATE-3 hot model consisted of a quarter core representation using one node

., assemhl,,ae1a11y, ane 24 ex1a11, The cold critical caicuiations were conducted in fuli

, L I

I core geometry.- For the thermal hydrauhes peruon of the code, the EPRI void correlation is assumed.

4.1.2 Hot Elaerwalue Calculations. Oveles 1 and 2 Table 4.1 presents the eigenvalues obtained from SIMUIRE 3 depletions for Cycles 1 and 2.

De critical control md positions and core conditions as a function of cycle exposure I

were taken from the standard EPRI data set found in Refemnces 18 and 19.

De core condition histories provided in Reference 18 show that certain '11P data sets were measured when the plant was not in equilibrium xenon conditions. However, the SIMUIRE 3 model was run assuming equihbrium xenon.

The equihbrium xenon cases yield an average eigenvalue of 1.00472 with a o of 0.00078. while the ent.re data set yielded an average elgerwalue of 1.00415 with a o of 0.00214.

I ne results of the entire data set are graphically presented in Figure 4.1.

As the figum demonstrates the equihbrium xenon eigemalues versus exposure and cycle are quite I

consistent. The sohd une is the average k effective and the dashed knes are the plus and minus la calculated using the equilibrium cases only.

4.1.3 Cold In-S auence criticals. Oveles 1 and 2 I

A series of cold in sequence criticals were run during Cycles 1 and 2. ne SIMULNIE-3 results, corrected for the period, are presented in Table 4.2. We average predicted eigenvalue is 0.99674 with a to of 0.00258. In reviewing the data, the first eigemahle at the beginning of Cycle 1 appears suspect. If this value is omitted, the average eigemalue is 0.99586 with a o of 0.00071.

In any case, with the exception of the first critical, the values are very consistent and show no trend with cycle number or cycle exposure. Figure 4.2 shows the results versus core exposure and moderator temperature. No noticeable trends are evident in the results versus these parameters.

In Figure 4.2 the average k effective of 0.99586 is portrayed as a solid hne with the o's portrayed as dashed lines. Notice that the first in-sequence eigenvalue hes far from the norm. This value is more than two standard deviations fmm the average and therefore, may be ehminated fmm the statistical summary.

I I I

I 4,1,4 cold incal crinnale. mainnina of evele 1 At the beginning of Cycle 1 a series of local criticals were conducted at cold conditions.

Table 4.2 also presents the SIMUIAE 3 calculated values, corrected for the I

period. The consistency is excellent. De avemge eigerwalue was 0.99829 with a o of O.00053.

I 4.1.5 TfP Trace Evaluation. Oveles 1 and 2 TIP traces were analyzed for Cycles 1 and 2 of the Quad Cities, Unit 1, Figures 4.3 and 4.4 present the di!Terence, in percent, for each integrated TIP tmce, calculated by

-l SIMUIAE-3, versus the plant measured integrated UP trace. The data in the figures is the average percent difference over the entire cycle in each instrument location.

Figure 4.3 presents this data for Cycle 1, and Figure 4.4 presents this data for Cycle 2. Also included.

in each of the figures, is the standard deviation, relative to the average, of each location.

These figures give an indication of the accuracy of the SIMULAE 3 model in predicting the I

radial power distribution in the core. The average percent difierences, standarti deviations of the averages, and the RMS for all TIP locations, as a function of exposure, for each statepoint analyzed, are sununarized in Table 4.3.

No exposure or cycle dependence is apparent. De overall average absolute percent difference was 5.46% and the average standard deviation was 5.52%.-

Figures 4.5 and 4.6 present the core average axial TIP trace comparisons for l

SIMUIAE 3 and the plant measured data. The comparison of prv11cted data and measured l

• • • • • • • "'"'"*'S'""'*"

EW the two cycles analyzed. These two cycles include a range of exposure, control rod insertion, power level, and flow rate.

Part of the difference between the SIMUIAE 3 and the plant measured data arises from the '11P asymmetry. That is, an instrumentation error believed to be caused by bowing of the instrumentation tube. 21s causes it to move out of the center of the narrow / narrow water gap. Neutron TIPS, such as those of Quad Cities, are significantly affected by such mis-De analytical model, of course, assumes that all instrumentation tubes are orientation.

centrally located. Approximately half of the differences between SIMUIAE-3 and the plant I I

I

.g MP data is ascribed to BP asymmetry, nis is the amount of knprovement observed in the current model when Vermont Yankee replaced neutron sensitive devices with garmna sensitive devices."

4.1.6 Quad Cities 1 Summary I

The Quad Cities Unit I validation study using SIMULA~IE 3 yielded very good results.

The hot eigenvalues are very consistent and within an acceptable range.

De power distributions are also good considering the limitauons of the measured data set.

his benchmark demonstrates the code to perfonn as good or better than previous studies on Quad Cities.

I The cold eigenvalues am also well behaved versus varying conditions and do not demonstrate any significant biases, here is however, a difference in the average eigenvalues I

from hot to cold conditions. Since these two conditions are ahvays analyzed separately, this does not pose a problem. The hot and cold cross section TABLES 3 libraries have unique I

applications which do not overlap.

4.2 Forsmark Unit 1 Oveles 1 throuch 6 Validation A similar validation of the SIMULKIE-3 code, as was performed by Yankee on the Quad Cities 1 reactor, was performed by Studsvik of America on the Forsmark Unit 1 BWR" A summary of the Studsvik results is presented in this report since Forsmark provides several I

different validations of SIMULATE 3 beyond those presented by the Quad Cities analysis.

Fonunark 1 is a ASEA-A'IOM BWR-2700 reactor located in Sweden. We core consists of 676 fuel channels. Control is maintained by cruciform control rods and the cort circulation flow system. Dere are 36 VP detector strings. Gamma sensitive UP devices were installed near the end of Cycle 5.

Rus, Fommark demonstrates the improvement in the plant measured to SIMUIA'IE 3 predicted TIP data comparisons when the effects of UP asymmetry are reduced.

I The analysis consisted of six cycles of operauon. These cycles included a variety of fuel designs. Table 4.4 presents a summary of the different fuel designs that were employed. The I B

I analysis of Forsmark-1 adds to the SIMUIAIE 3 BWR validation by including several characteristics of the reactor which are different from Quad Cities, nese additions include :

Axially graduated Od (with black and grey Od zones) 8x8 and 9x9 fuel rod array bundles Higher Od loadings Reconsututed depleted assemblies I

load following and long coast dowTi periods low enriched and natural U ends For each cycle analyzed, hot eigenvalues as a function of exposure, and associated 'I1P trace measurements were evaluated. A series of cold critical eigenvalues wem also analyzed.

4.2.1 Forsmark 1 Model Descrintion The Forsmark CASMO-3G statepoint calculations were constructed similar to those of I

Quad Cities.

Gadolinta cross sections were generated using the MICBURN 3 code.

'111e SIMUIAIE 3 model was similar to the Quad Cities model. The EPRI void correlation was also used in the SIMULATE-3 analysis.

4.2.2 Hot Einenvalue Calculations Cveles 1 through 6 Table 4.5 presents the eigenvalues obtained from SIMUIAIE-3 depletions for Cycles 1 thmugh 6.

The average eigenvalue for all cples is 1.00429 with a nominal standard deviauon of 0.00114. (Reference 24 did not contain enough detail to calculate the actual I

standard deviation of the hot eigenvalue over all cycles. Standard deviations were reported for each cycle individually. The number reported here is simply an average of those values found in Table 4.5.) ~ 'Itie eigenvalues for individual cycles exhibit somewhat smaller deviauons indicating consistency during the cycle, however the average eigenvalue from cycle to cycle changes.

• 111e largest single change in the average eigenvalue, from one cycle to the next, is 0.31%ak (Cycle 3 to Cycle 4), well within the reactivity anomaly criteria of 1.0Mk.

With the exception of Cycle 1A (which Reference 24 indicates did not have a smooth power history), the Fommark 1 average eigenvalues lie in a mnge from 1.00270 to 1.00725 I I

with standard deviations of 0.00049 to 0.00158. Notice that the Quad Citics equilibrium hot eigenvalue results, reported in Section 4.1, fall in the middle of this range (1.00472 to 0.00078) demonstrating consistency between the SIMUIATE 3 models for two different reactors.

E 4.2.3 Cold Critical Calculations Oveles 1 through 6 I

A number of cold critical calculations were conducted at the Forsmark reactor at the beginning of each cycle. Table 4.6 presents the SIMUIAIE 3 results for the cold criticals. A total of thirty seven criticals were calculated. For all criticals the average eigenvalue was 0.99857 with a nominal standard deviation of 0.00108. The results are not as good as the Quad Cities nsults.

However, the results are still acceptable for predicting cold criticals because the reactivity anomaly criteria is 1%,ik. Again, the hot to cold eigetwalue bias observed in the Forsmark 1 is similar to that leported in Section 4.1 for Quad Cities.

4.2,4 UP Trace Evaluation. Oveles 1 through 6

3 TIP traces were analyzed over the six cycles of operation at Forsmark 1.

As was done for the Quad Cities benchmark, both axial and radial behavior were examined.

As an indication of the radial performance of the SIMULATE 3 model, axially integrated EP traces l

were analyzed.

The petrent difference between the calculated and measured data was averaged at each UP location in the core. Figures 4.7 p!rsents this data averaged at each location over the six cycles analyzed. Notice that, when all cycles are considered. there is no apparent radial tilt in the results (that is, the + and differences att random'f distributed).

ne average absolute percent difference for all locations shown in Figure 4.7.s 1.7% and the I

average of the standard deviations shown in the figure is 2.7%.

In examining the results for each cycle individually, Cycle 1A exhibited the worst agreement, nese results are presented in Figure 4.8.

For this cycle, the absolute percent difference for all locations is 2.8% and the average of the standard deviations is 1.6%. Cycle 6, which had ' gamma sensitive TIP devices, demonstrated the best agreement. De average absolute percent difference is 1.6% and the average of the standard deviations is 0.6%.

Reference 24 also reports results of comparing axial point by point EP data at each location l

between plant measured and SIMUIAIE-3 calculated data, at each TIP location.

These l E I

. ~.

I results are also avemged over the six cycles, and each cycle individually. nese resulta yield an average point by point difference of 4.8% over the six cycles.

Cycle 1A exhibited an average difference of 6.9% and Cycle 6 yielded an average difference of 3.9%.

The Forunark report did not include core average axial MP trace comparisons between the plant measured and S!MUlA1E 3 in a manner consistent with that shown for the Quad Cities analysis (the Quad Cities analysis included axial EP trace comparisons for each statepoint analyzed). Rather a sample of trace comparisons were presented for the beginning, middle, and end of each cycle. Because the gamma sensitive DP deWces are the least prone

~

to instrument error, only the axial TJP trace comparisons for Cycle 6 are presented here, Figure 4.10 presents the axial core average TIP trace comparisons at the beginning, middle, I

and end of cycle, ne axial agreement is very good considering that Cycle 6 contains the accumulated void and exposure history effects of all the previous cycles.

l Both sets of data demonstrate that, overall, SIMUIATE 3 predicts the power l

distribution in the core quite well, and to a degree of accuracy comparable to the Quad Citics analysis. As was discussed previously, the installation of gamma sensing devices reduces asymmetry in the plant measured data, which results in improved comparisons to the Sm!UIAIE 3 model.

4.2.5 Forsmark 1 Summary The Forsmark 1 validation using SBtUIATE 3 demonstrate the capabilities of the code to I

model real plant operation. No special treatment of these designs were required in terms of l

either the CASMO-3G calculations performed or for the SIMUIAIE 3 model.

lI I

I l I I

I TABIE 4.1

SUMMARY

OF QUAD CTImR CW'tRR 1 AND 2 HCYl' DEPLETIONS Cycle

'HP Data -- Exposure Power Flow Rods Xenon K-ef Set (Owd/'O Icvel (%) Rate (%) Inserted Eq.

(Pred.)

1 1

0.247 87.0 86.1 1936 No 0.99909 I

2 0.646 89.7 101.6 2032 Yes 1.00353 3

0.800 89.2 99.6 2056 Yes 1.00420 4

1.334 87.5 99.6 2192 Yes 1.00409 5

2.031 97.6 100.0 2056 Yes 1.00515 6

2.894 96.1 97.2 2156 Yes 1.00593 7

3.480 87.5 96.8 2436 Yes 1.00494 I

8 3.696 92.4 96.7 2316 Yes 1.00650 9

4.297 94.7 94.8 2290 Yes 1.00544 10 4.809 93.1 92.8 2256 No 1.00389 3

11 5.471 80.2 75.0 2104 No 1.00154 12 5.949 88.6 99.9 2064 No 1.00488 13 6,175 88.0 96.1 1892 unknown 1.00583 14 6.710 90.3 97.6 1924 No 1.00357 15 6.948 87.8 97.9 1688 No 1.00713 16 7.239 87.8 97.9 1688 Yes 1.00438 I

2 17 6.625 56.7 50.3 1788 No 0.99770 18 6,833 86.5 85.5 1408 No 1.00289 I

19 7.225 91.0 89.8 1204 Yes 1.00467 20 7.641 83.5 83.7 1832 No 1.00082 21 7.973 96.0 99.4 1160 Yes 1.00411 l

22 8.293 99.6 98.7 956 No 1.00617 23 9.229 98.1-99.2 744 No 1.00703 24 10.195 98.5 98.0 368 No 1.00321 25 10.827 85.8 100.3 336 No 1.00574 26 11.699 72.8 95.8 32 Yes 1.00488 27 11.973 68.2 96.0 0

Yes 1.00440 l

28 12.348 69.6 97.7 0

Yes 1.00439 29 12.466 59.2 96.8 0

Yes 1.00418

""~

I LI

I I

TABLE 4.2 g

QUAD Cri1ES CYCLES 1 AND 2 g

COLD CRmCAL CASES In Sequence Criticals Date Cycle Exposure Sequence Temperature Period K effective L

(Owd/'ll (T)

(Sec)

I 4/5/72 1

0.0 A

147 230 1.00293 2/8/73 1

2.6 A

160 300 0.99615 5/7/73 1

3.4 A

120 120 0.99573 8/7/73 1

4.48 B

120 45 0.99479 I

1/6/74 1

6.27 A

180 300 0.99642 10/6/74 2

7.51 A

185 100 0.99514 12/16/74 2

8.29 D

160 45 0.99683 5/4/75 2

9.619 B

190 130 0.99596 I

E Local Criticals I

Local Period K efective (sec) 1 160.

0.99845-2 45.

0.99763 3

60.

0.99774 4

148.

0.99835 5

238.

0.99753 6

25.

0.99913 I

7 260.

0.99816 8

350.

0.99852 9

30.

0.99886 10 175.

0.99854

'I I I

I TABLE 4.3

SUMMARY

OF QUAD CmES CYCLES 1 AND 2 W DATA I

Cycle Exposure Average Stanaan! Deviation RMS (Owd/O Dif!'erence 1

0.247 4.45 5.60 5.53 "I

0.646 4.84 5.87 5.80 O.800 4.87 5.92 5.92 1.334 5.08 6.14 6.07

,I 2.031 4.87 6.06 5.99 2.894 5.15 6.35 6.27 3.696 5.37 6.47 6.39 4.297 5.05 6.13 6.05 I

4.809 4.88 5.99 5.91 5.471 4.49 5.63 5.56 5.949 5.09 6.21 6.13 I

6.175 4.95 6.02 5.94 6.710 4.61 5.51 5.44 6.920 4.92 5.96 5.89 6.948 4.91 5.96 5.89

,I i

2 6.610 4.12 5.00 4.94 6.625 4.12 4.99 4.93 lI 6.833 4.45 5.36 5.29 7.225 3.97 4.78 4.72 7.641 4.24 5.16 5.10 7.973 4.06 4.91 4.85 I-8.293 4.08 5.04 4.98 9.229 4.14 5.13 5.07 10.195 3.74 4.62 4.56 10.827 4.12 5.05 4.99 I

11.973 4.07 4.80 4.75 12.348 3.85 4.50 4.44 I

I I

II

,3 1

LI I I t

I

I

-l TABLE 4.4 E

CHARAC'IERIS11CS OF FORSMARK UhTI' 1 FUEL DESIGNS load Geometry Enrichment Gd leading Gd Pins Segments Cycle Init a 8x8 2.08 3.95 3

3 1 A.1B,2 Init b 8x8 2.08 0.0

.0 1

1A Rel 1 8x8 2.80 2.00 4

3 2,3 Rel2 8x8 2.82 2.00 4

3 3,4 Rel3 8x8 2.94 2.50 4,5 3

4,5,6 t

Rel4 9x9 2.91 2.50 5,6 4

5 I

Rel5 9x9 2.94 2.50 6,7 3

6 em i 8x8 2.82 3.00 o

3 E

Demo 2 9x9 3,00 3.00 6

2 4

y 1

4.5,6 cn 8

I I

I LI 1 s a

'I I

LI lm 51 -

!g IE

I I-'

TABW 4.5 FORSMARK UhTI' 1 HOT EIGENVALUE

SUMMARY

Cycle Average Standard Dedation Number of Points 1A 1.00063 0.00228 9

1B 1.00270 0.00082 11 2

1.00380 0.00158 23 3

1.00418 0.00104 15 4

1.00725 0.00049 22 5

1.00450 0.00093 16 6

1.00697 0.00084 13 l:

u TABW 4.6 FORSMARK UNTI' 1 COLD CRmCAL

SUMMARY

3, Cycle Average Standard Deviation Number of Points I:

1A 0.99682 0.00262 7

IB 0.99260 0.00219 6

2 0.99920 0.00134 5

3 1.00050 0.00085 4

i 4

1.00327 0.002.13 5

I 5

0.99929 0.00234 5

S 0.99830 0.00112 5

I I

I I

I I

.. - -. - ~. - - - - - - -.

- - ~ ~. - ~... -.. - ~. -

E FIGURE 4.1 I

QUAD CmES CYCLES 1 AND 2 HOT EIGENVALUES I

I QUAD OTES CYCLES 1 AND 2 HOT EIGENVALUES VS. CORE AVERAGE EXPOSURE t010 I

O O

I O

............g.....I........y.......... 0,,,,,,,,,,,,,,,,,,,,,,,,,,,g,,,,,,,,,,,,

t005-a o

a

... 4. 9....................c.......... 6** ***A**************************9'**

O O

t000-O-

I 0

I 0_

E. _ _ _

I O NON-EQU!UDIUUM 3o

- AVERADE VALUE '

I i-0.990 0

1 2

3 4

5 6

7 8

9 to 11 12 13 EXPOSURE (GWCVT)

I l

I I;

I

I

.__...__.__._...___.....__..--~_.___.__....._.-_.._...-.__m.-

~ _. _. _

E

.3 FIGURE '4.2

,E-QUAD CmES CYCLES 1 AND 2 COLD EIGENVALUES RESULTS I

1 Apc 2 cnmCALS DromstE I

AVERAGE VALUE LOODa l,

i 1000-l-

M

..................................................g.............. 9.............

, I O

0.05-..................................u '************a******** A *******************

i LI 0.000 e

i-

,2 3

4 0

0 7.-

0 e

4 I

DFOSURE (OWyQ QUAD 1 AND 2 COLD N-CRmCALS VS.WODE TDdPUtATURE tem

.... ao I-AVERAGE VALUE t006-I O

1000-a

...............................................g..............g...............

..I 0.00s...............;..................................................a...........

v v

v 10 0 as 30 00 300 TDAPERATLNE MN FMfEPeCT)

I 54 -

I

=--

I

-I E

F10URE 4.3 I

QUAD CTI1M UhTP 1 CYPm 1 TIP TRACE

SUMMARY

PREDICED MEASURED AVERAGE '11P LVIEGRAIS I

1 2

3 4

5 6

7 8

9 10 11 12 13 14 15 16 AVERAGE DIFFERENCE 1 STAND DEV 2

I l

I 3.3.-

5.3 --.

3.2. ----

5.7 ~ ~

-4.7. -..

3 1.4 ~~

1.6 ---

1.2 -..

1.9 --..

2.3.~.

i 2

4

- - - - - - - - - - - - - ~ ~ - - - - - -

5 2.5.--

4.2.~.

5.8 --.

5.3.. -.

5.2.--

5.4.....

~

1,9.....

4,7 -..

1,8.....

1.5 -...-

2.7~~

1.1.-- - -. -

I 1

6

....... - -....~... ~. - -.. ~ -... -....... ~. -.-....... -

l Ll 3 3.8 _-

.i.2.....

6.7 -...

4.4 -.

2.4...-

1.5.....

6.2.....

7 1.5....-

2.1 --.

2.0 - --

1. 6 ---.-

1.2 ----

1.4 -- --

1.7.--

l 4

g 3.5 --

- 1,6. ---

9.0 ----

7.2 ---

- 1.7.-.-.

-9.7 -..

3.3.--

g I

2.0 ---

2,1 - --

2.5 ---

1.8.---

1.2 - -.

1.4 ----

2.0.....

i 10

--... ~ ---.-.. ~ ~ ~ -.....................................................

I 1.2 - --

4.9 --.

0.0 -.--

2.4...-.

4. 4....-

4.4..--.

2.1 11 1,7.....

2.6 ----

2.9 --...

1.2 ---.

1.7 ----

1.0.-~.

1.1 l

12

.......... - - -.-..-....- -.- ~ ~....,.. t.'....

13 3.4 -----

11.7 ~~ - 1.2 ---.-

.1. 4.---

10.4...-.

2.7 -.-

1.5.-...

1.3 ----

2.0.----

2.4 ----

1.9.- --

1.4.--

l l

i

}4 1

.15 9.4 ----

-4.3 ----

6.4 ~~

i l

1.8 ~ ~

3.2 -----

1.5 ---

l l

l.

16 1

l l:

I t

I

- 55 i

I

E FIGURE 4.4 QUAD CTIWL UNIT 1 CYCLE 2 TIP 'IRACE

SUMMARY

- I PREDICIED - MEASURED AVERAGE 'ITP IN'IEGRA14 1

2 3

4 5

6 7

8 9

10 11 12 13 14 15 16 AVERAGE DIFFERENCE

~ ~. - -- -

1 STAND DEV B

2 3

-0.9 --- 0.3 6.3 ---

7.9 - ---

-5.7.--

3 1,0.. ~

1,6 -..

1.6.-.--

2.6 --. -

1.0..~

4

~ ~ - ~ ~ - - - - - - - - - - - - -

~ ----- -~ ~~ ~

I

....... ~..................................................

. 2.3 ----

3.1 ~~

1.2 - -

2.4 ~~

2.3 ~~

4. 8 --- ----

5 1.3 -

2.3 ---

1.9 ----

1.4 -. -.

1.8 ----

2.4 ---

6

~ ~- -- -- -

~ ~ ---- ~ ~ ~ ~ ~ ~

- -- - ~ -- -

I 7

-. -- ~ ~ - 9. 5 ---

5.1 -- -

8.0 - ---

7.7 ~~

4. 8 -----

5.2.---

-4.4 ---

2.3--

1.1 -- --

2.4 - --

1.2 - -

1.3 ---

1.6 ----

1.6 --

8

- - - - - - - - - - ~ ~ - - ~ ~ ~ - ~ ~ - - - - - - - - - - - - - - - - ~ ~ - - - - - - ~

' 3.4 ----

2.2 - - -

5.9 ---

0.3-----

-2.0 -----

-6.0 -- --

2.9 - -

g 1,1 ~..

1.3 --- -

2,0 - --.

0.5 -- -

1.3 - ---

1.6 -- -

~ 1.6 - ---

.I.

10

- - - - - - - - - - - - - - - - - - - - - ~ ~ - - - - - - - - - - - - - - - - -

2.4 -----

5.5 --- -

1.7-6.0~~

5.3 -----

1.7 -----

3.1 11 2.3 ---, 1.9 -----

1.2 ---

0.8 - -.

1.4 -- --

2.3 ---

1.4 8

12

- - - - - - - - - - - - - - - - ~ ~ ~ ~ - - - - - - - ~ ~ - - - - - - - - - - - - -

13

-2.7 ----

6.6 ----- - 1.2 ---- - 1.9 -----

2.4 -----

0.3 e----

1.9..-.

2.0 -----

1.1 ---

2.1 ---

2.5 - --

1.7 -.--

14

- - - - - - ~ ~ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -

5.6 -----

-6.9 ~~

3.2 --~

15 2.0 -----

3.1 ---

1.4 -- -

16 I

56 -

I

......-...+.=.r-

. +

-~ -. -.. ~.. - - -,

~a..

..-w.a

..~a.+

aan.

.-..->~su

.~-a-I F1GURE 4.5 OUAD CmES UNTP 1 CYCLE 1 TIP TRACE COMPARISON

SUMMARY

___7 ma

__m E

E E

m m.

4 a

g W

E a

m!

E!

Q gp a

e e

a E

m E

E 1

g, m

X m

a m

g E

E m

I e

M e

a e

I e

h a

e' e' s

g,a m

a m

a I

E E

E E

E E'

I' E

E I

E

,E I

E m

m E

EI E

X' e

a a!

e a

e a

e a

a e

a a

W W

E E

m m

E E

W

[

m m

K E

E E

E m

a a

e a

e m

E m

m m

E m

e m

av m

a m

e u

a e

a a

e a

e E

E m

E m

E E.

,,g g,m m

E E

E E

E E

E E

I.

a e

e e

a 4

e a

a e

e a

1

_a a

e' I

E E

E E

E a

a

=

=

=

=

=

E E

E E

E I

m m

E E

E 1

e 4

a a

e a

a e

a e

a I

I a

e e

a E'

E E

X.

E I

E.

E E

E m

E 1m e

a 1

m e

le e

b____

edue.

a g'.

um _

g g

1 s

I l1 x

il s

de p

a u

[, E.

y I

3 66 i

E E

E n

q K

K i

o m

e b

d h

a a

n n

E a

a a

2' q

F E

E E

X:

OO F

E I

E E

I E

4 E

E 4

X a

e a

a e

a UN e

a e

a a

e I

!g H

H K

E m

m m

E E'

D E

X.

E a

a e

e e

a N

W E

E E

E E

E E

W a

e a

a a

a a

e e

a a

a a

m I

m m

m E

E E

m E

E e

a a

e a

=

m a

m a

a a

e a

e e

a e

a a

a E

E E

E E

E E

E E

m E

m E

E E

E E

E a

a e

S a

e a

a A

e e

e a

e a

E E

E E

K.

E E

I E

I E

9 E

E A

E.

a e

a a

e la n

a a

n M

e a

e H

I E

I H

E.

E' I

gg E

E E

E e

__a m_

gg vs I

1 t

3 x

8 m

m a

m i

d X

E e

E E

I m

er m

a e

E m

e E

m X

a e

a a

e a

a mE a

e a

K I

E m

E g

g g

m E

m E

E n

a a

4 4

e a

W W

W m

a er a

e a

a E

E E

E E

E E

- I E

E m

m X

X a

e 4

e a

e e

e 5

a a.

a m

E E

E E

E E

E x

E a

ml e

a a

E m

m m

E E

a a

E a

I a

e a

a e

a E

a:

a e

E.

E E

D E

E.

pp 4.

I m

E I

E 1

e A

p.1-u h

___mm_

,g.

e I.

k

\\

/

~

I.:

I FIGURE 4.5 (CONTINUED)

QUAD CmES UNrl' 1 CYCLE 1 TIP TRACE COMPARISON

SUMMARY

g D

r.

p m

m i

e e

a m

m x

m r

i m

a e

e a

m l

m m

a a

m W

w

,m,

m m

m a

m m

m j

e a

e e

e e

(

a a

I e

I K.

K.

e a

a e

a TL i I.

m u

z z

a i l

[

m m

a

,m L. -

es e

e' I

e

,e e

e e

e I

it

.1 a

e e

f g

8 E

E a

Q p

K z

r h

d m

a a

e i

s e

e a

e

=

a is e

a a

a e

i

))

e e

e e

e a: i e

a e

a a

e i

l.

y p

s i

m m

m a

m x

x-a m

z x

x x

r N

N e

e a

e e

a I

e a

e e

e e

I l

l m

m m

x x

x x

x x

x s

a x

r a

e a

m e

e i

e a

e e

a e

i e

m m

a m

I X

a u

I e

X X

Ki m

e e

e e

e i

e a

a e

e e

i

]

m m

m m

u I

m X

X z

l

. I m

a e

e e

a a

e

$11 i

5 88 l

ee,,_,,

WW e

e p

.e g-Ll l

s 4

a e

J j

R q

5 s

m m

s e

e a

e a

d 2

m

=

x x

2 m

a a

e m

a m

1 y

g m

z e

m a

m W

W m

a e

m a

a m

u m

a a

m m

a a

e n

(.

y}

m a

m m

m a

a e

a a

19 2

m x

I E

21 I

m a

e a

m I

gg m

m 49

---==m I

h gg 8

9 0

e 9

I

~

I F100RE 44 QUAD CITIRR UNIT 1 CYCLE 211P TMM'E COMPARISON

SUMMARY

I y

e E

a s,

u m

m S

I E a

a m1 m

g m

a a

m I

i a

M N 'M M

M a

a a

a m

m m

s y

g y

I m

m m

I XI I

m I

m a

m E

a m

m m

m m

E m

M M

M N

m m

M m

m m

m 1

I I

I m

E m

m m

m m

m.

J M

m' 4.,Ll2{I g

m x

n I I

2 I

m a

a a

a m

ya si n

m n

m m

p _

m

~

~

m_

m m

m E

I m

m m

m I

E I

I I

e a

____d g#

8, e

9 4

I ete t#

8

8 ete' e',,,g 6'O 4

0 9

O e

9 0

I il ei j"

/

7 S,,

g a

a m

m m

S D

k I.m X

ws o

m (f

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i F10URE 4.6 (CONUNUED)

QUAD CmES UNTT 1 CYCLE 2 TIP'IRACE COMPARISON

SUMMARY

l I

i pu__innsu R

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I FIOURE 4.7 j

I FORRuARK UATP 1 T P TRAPP EUMMARY CYOf PA 1A *MROUGH 8 PRPhlCTED MEASURED AVERAGE TIP IN'IECRAIS 1

l 1

1 2

3 4

5 6

7 8

9 10 11 12 13 14 15 16 AVERAGE DIFFERENCE

~~.~. -.~.

1 TTAND DEV 2

-....... ~............. ~ ~..... *.

........... ~.....................

m..

I 1

3 0.26.~.

0.94. ~.0.33 ~~ 2.06 ~. ~~

~~ -~ ~~

2.28... -

1.97 - 2.84..~

1.87 ~..~.

I 4

.................................................. ~.......

i 1

5 0.54 ~~.0.49 ~~ 0.41 ~.. 2.80 -...

1.23.~..- ~~

-. - ~ ~

2.55 -~ 2.4 5.~. 3.82 -~ 2.59..~ 2.97 -~ ~~ -~

...... ~

6

~~ ~

-. - - - ~~ ~~ -

~~ - -- ~~ ~~ ~~

I 1

7 2.53.~.

0.82.. -

0.26..~ 2.63 ~~

0.41..~

2.53.-~.0.65 -~ ~~

2.85 ~~ 2.74 ~ ~ 2.60 ~~

2.72 ~.. 3.7 7 -.- 3.32 --- 2.85.~. ~~

l 8

1' 1,46.... 2.60.....

1.86 ~. 0.77.~. 0.87 - -.5.31 ~. 0.98 ~~ - --

g 2.07.~. 2.00.-- 2.85..~

2.95.~.

1.96 -~ 3.89 -~

3.24 ~.-

.. ~

~..

10

---. - ~ ~ - ~ ~ ~ ~.. ~ ~ ~ ~ ~ ~ ~ ~. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

....................................~........................................

11 1.09 -~.0.32 -.

1.28.-~ 0.95 -~.2.28. -- 3.10 -

.*~.~.

2.86..~ 2.77 -~

2.07. - 2.85 ~-- 2.95 -. ~ 2.35 ~ -

...~.....

s

}q

..................... ~.............................................

I 13 1.93..--.2.15 ~~.0.39 -. 0.37 ~..

1.63..~ -.

~~

2,09.....

1,7 7...- 3.51 -~ 2.54. --.

433.~.

~..

I 14

. - - -. ~ ~ - ~ ~ ~ - - -. - - -.

~. ~ ~ -. -.. ~

........................... ~..............................

15 7.41 - - 4.03 ~~ ~~

..-. ~~ ~~

4.13.~. 2.33. -...---

16

~~ ~~ -

~~

i 61 -

~

I FIGURE 4.8 I

PORSMARK UhTP 1 'nP ' MACE

SUMMARY

OYCLFR 1A PRrnlC'lTD. MEASURED AVERAGE *NP th'TEORALS 1

2 3

4 5

6 7

6 9

10 11 12 13 14 15 16 AVERAGE DIFFERENCE

~ ~ -... ~ ~. ~.

1 STAND DEV

~~..~ ~~ ~~

2 3

3.20.....

1.e2.....

1.le....

3.02.....

3 2,1 1.....

1.26 ~~

1.08 -.-

1.04..~

~~

t I

4 5

.-~ ~~.2.76 ~.

.1.56 -~.4.13.~. 3.31 ~. 2.61.~..~..~.

2.10 ~~ 2.17.~.

2.39.~.

1.33.~. 0.9 1 ~ ~ ~.... ~

6 I

5.66.....

2,49.....

1.80.~.

2.73.~. 2.26 ~~

1.10 ~~ 3.90 ~ ~ ~.-

7

~~

1.28 ~~

1.78 ~~

1,62 ~~

1.33 ~

2.62. ~ 0.62 ~~

1.68 ~~.-..

8

....................................................... ~ ~....................

i l E 9

4.90 -.... 2.53..~ 0.65 ~~

0.72 ---

1.57.-.0.36 ~~ 3.36 ~....~

i g

3.12 -~

1.65.-

0.93.~.

1.10.~. 0.90 ~~ 3.04 ~.+

1.58.~..--

t

) f,)

............................................................................ ~

11

~~ --.1.77..~.3.22 -~.0.12 --. 2.28.-- 3.96 ~. 4.35 -~

2.54 ~~

1.22 ~~

1.56 -.- 2.26 -.

1.93 ~~

0.88..-.

I 12

.e...

....e I

13 2.00.~. -3.72 ~.. 3.43 ---. 2.28 ~.. 0.87.-~. ~

1.21 ~ ~

1.32 --.

1.08 -. 2.40.~.

1.78 ~. ~~

I 14

~ ~ ~ ~ ~ ~ ~ ~. -. ~... - ~ ~ ~ ~.. ~ ~ ~ ~ ~

15 6.18.~. 6.38 ~~ ~~

~... ~. ~ ~

1.84 ~~

1.00......--

,6 I

I I

l I 62 I

I FIGURE 4.9 I

FORSMARK UhTP 1 'nP 'IKACE

SUMMARY

CWLES 6 PREDICTED MEASURED AVERAGE TIP IhTEGRAL I

1 2

3 4

5 6

7 8

9 10 11 12 13 14 15 16 AVERAGE DIFTERENCE

.~...~.~...~

1 FrAND DEV I

2 I

3 1.54 -.- 0.96 -.-.2.59.~. 0.17 -- ~~

0.33. ~ 0.44 -- 0.74 -~.

0.27-.-

.m.

~..

I 4

3.15 ~~.1.71 ~~.0.35 ~.. 0.50 ~~.2.47..~ ~~ --

5 0.93.... 0.35

.. 0.52.~. 0.95.~.

1.03 -~ ~~ ~..

6 I

0.14 -...

1.54 ~.. 0.34 ~~.1.50 ~..

1.G4 -..

2.31..~

2.2 6.~. ~ ~

7 0.91.-.

0.53 -~

1.03 -~

0.68.~.

0.99 -- 0.87.-- 0.36 - ~ ~ ~

ll 8

.................................................. ~...........................

................ ~..................... ~........... ~........... ~.......

ig 9

1,29..... 0.67 ~~

1.31 -- 2.14 -.

0.50.-~

.2.59 ~~ 5.08 ~.....-

g 0.85.~.

0.4 5.~.

0.75.~.

0.38 -- 0.71 -.- 0.43 - - 0.36.- - -

10

~ + -.. ~ - - -. - ~.. ~ - - - -. - - - - ~. ~. - - - - - ~ ~. ~.. - -.. ~.

0.82 - ~.1,94 -.- 0.83 --.-

1.41.~.

0.05 ~. 3.05 --.

11

~ ~.....

0 45 ~-- 0.84 - -

0.46.--

0.72 -~ 0.54 ~ 0.72 --

e 12

- - -. ~. - - ~ ~ ~ ~ - - ~ ~ ~ ~ ~ ~ ~ ~ -. ~ ~ ~ ~ ~ ~

t i

I.

2.19 -.- 0.07.~..

1.68 ~~.3.02..-. 0.55 --. -...

13 0.69 ~.. 0.93 - -

0.55 ~~ 0.4 9.~.

0.4 9......-...

I 14

. ~... ~ -. - ~.. ~ ~ ---. -...~ -~. ~ ~.-.....

...................... 4.

4.25 -.

1,85 -- ~..

15

~ ~ ~ ~

0.25 --. 0.47 -. -~

1e I

I i a 63 g

I

. - -. -. -. -.. - ~. _.. ~ _ _. _ - - _.

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=

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, I tbb' it it it 0 is 5 5 4 h h h 5 is i ki i iii t i e l

eeqwns epou p!ry A

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it it it it it it heih5hhbb55i iI ii 5 i N i e seqwnu 6pou p!vy

• 64

• I 5.0 PWR PIN POWER RECONSmUcnON VAUDNNON I

This section prwides validauon of the pin power reconstruction capabiltues of SIMUIAE3, An extensive benchmarking pmcess was conducted consisting of comparisons of SIMUIA'IE-3 data to entical experiments, higher order calculations, and pin by pin diffusion theory (PDQ 7) results.

I The critical experiment comparisons provide a direct validation of One accuracy of die method for the range of independent parameters examined. In addition, higher order transport theory solutions were generated by modeling multiple assembly or colorset geometdes using j

I CASMO 30. Rese benchmarks demonstrate the accuracy of the SIMUIAE3 method versus I

changes in independent parameters such as exposure, enrichment, burnable poison shtms, and control rods.

In addition, a companson of pin by pin diffusion theory (PDQ 7) to SIMUIAE3 pin by pin distributions was performed for an actual PWR core model.

McGuire Unit 2, which was presented in Section 3.1 of this trpon, was modeled with PDQ 7.

The SIMUIATE 3 model l

presented ^n Section 3.1 was used to generate the pin by pin data.

5.1 yaMntionyersus Mesured C !ncal Eroeriments 1

J The Babcox and Wilcox measured critical experiments provide direct venficatitsn of the pin distribution calculation in that actual pin distributions were me.asured and compared to i

SIMUIATE 3.**

These criucal experiments covered various PWR lattices consisting of enrichment

(

vanations, small and large v'ater holes within the latuces, and an array of poison material, l

including gadolinia. ne results demonstrate SIMUIAE3 to be accurate to within an RMS difference in average pin power of 0.8% in lattices not containing gadolinia, and 1.3% for lattices with gadolinta. Table 5.1 presents a summary of the results of these critical expenments, and lI 1s taken directly from Reference 25, 21s verification, however, does not include fuel depletion I

effects, and therefore Yankee Atomic performed further validation of the pin power reconstruction capabilities.

]

65 -

I

~_

I 5.2 Vnhdatinn Venus Mensun d Reaction Rates The instrument thimble reaction rates, presented for the McGuire Urut 2 in Section 3 of this report, provide an integral test of SIMULAE 3 to calculate accurate relative fluxes in the centraDy located instrument thimble of typical PWR lattices.

Dese comparisons are a demonstation of the flux reconstruction capabilities of SIMU1AE 3, which is an integral part of the pin power reconstruction method. his verification demonstrated the overall accuracy of the SIMUIAE 3 calculated reaction rates, compared to measured data, to be within an RMS error I

of 1.5%. De comparisons were made for 19 different Dux maps comprised of approximately 58 measured samples per map, or a total of over 1100 samples of data.

5.3 Validntion versus Higher Order Numerical Calculations I

A series of multiple assembly (2x2) array pin by pin transport theory solutions were calculated as higher onier benchmarks, nese colorsets consisted of 17x17 fuel rod lattices of the standard Westinghouse fuel design. Fifteen different colorsets were constructed. Within the colorsets, combinations of enriclunent, exposure, control rods inserted, number of burnable I

poison pins, and removal of burnable poison pins were represented.

De values of these parameters were selected to be typical of those used in current PWR cort design.

D ese colorsets were depleted te generate pin power distributions as a function of exposure for the lI various lattices. Tabie 5.2 ' presents a description of th: differtni Vtices evaluated.

1 Transport theory solutions for these geometries were constructed using CASMO-30, CASMO 30 has the capability to physically model four different assemblies within the same l

problem description, with reflecting boundary condftions. However, the SIMUIAE 3 model was based upon group constants and flux discontinuity factura produced from the single assembly geometry CASMO 3G cases. No colorset information was used in generating these constants.

In addition, pin by pin diffusion theory (PDQ 7) was used to construct these colorsets, in order to illustrate the relative accuracy of both methods.

l Comparisons were made between assembly and pin powers for each assembly art:7 zed in the colorsets. De comparisons made were for SIMUIAE 3 versus CASMO 3G. L same comparisons were made between PDQ 7 and CASMO 30. Dese results show the SIMUIAE-3 and PDQ 7 assembly powers to be generally within 1% of the reference CASMO-3G solution, even in rudded configurations. Table 5.3 presents a statistical summary of the pin power 66 I

I

I results of the SIMULAE-3 and PDQ 7 models versus CASMO 30.

De results show the I

average and starulard deviation of percent differences for the peak pins, predicted worst pins and all pins for each depletion step of the colorset, ne results show both SIMULAE 3 and PDQ 7 accurately firedict pin power distributions.

The SIMU1AE 3 results indicate no trend with lattice except for the lattices containing control rods. In these latuces, the peak pin is accurately predicted and the overall average of all pins is quite good. De exception is that the worst pin in these lattices is in erTor by appmximately 3%. nese pins are at the central corner of the lattice where controlled and uncontrolled assemblies are acllacent to one another. However, the peak pin does not occur in I

these regions.

The PDQ 7 results are similar to the SIMUIAE 3 results, but with higher percent differences. The worst pins in some of the heavily pois::ned lattices does show PDQ 7 to exhibit somewhat larger differences than SIMUIAE 3. Again. In these lattices the worst predicted pin is not near the peak value.

I The PDQ 7 colorset calculations use a G factor multiplier in the cross secuon formulation for the non-fuel regions, ne non fuel regions are the guide tubes, with or without inserted I

burnable poison pins or curarol rod fingers, ne G factors arsure that the reaction rates in these regions are correct in the diffusion theory calculation, so that reacuvity worths of bunwele poisons and control rods are accurately represented.

Bis is a standard technique in the generation of fine mesh diffurolon theory cross sectic,ns.

I The results in hble 5.3 show, however, that there is a limited abitty to reproduce highly accurate pin powers with standard diffusion theory. In general, peaking in the area of guide tubes is under predicted and peaking in the areas of heavy absorbers is overpredicted. This fact has been recognized for many years and is addressed in a conservative manner in the core I

incenairyg process.

The summary of pin power results in Table 5.3 shows that the SIMUIATE 3 pin power reconstruction is a highly accurate technique which produces average and peak pin powers to within 1% for a variety of realistic reactor conditions in colorset geometry. For control rod insertions, this is also the case, but with worst difference in pin peaking increasing to appmximately 3% in non limiting pins. The accuracy of the technique is superior to standard I

"'~

I

I diffusion theory, which rebes upon addiuonal factors to assure conservative pin power g

B representation.

5.4 PWR Quarter Core SIMULATE 3 Reconstruction Cnmnned to PDO-7 A second benchmark was constructed which consists of quarter core representation of a contemporary Westinghouse design PWR *Iivo cycles of operation were simulated with both PDQ 7 and SIMUIAE 3 for the McGuire unit presented in Section 3.

Figures 5.1 and 5.2 I

present the loading pattems for Cycles 1 and 2 of McGuire Unit 2.

We loading pattems contain a variety of enrichments and number of bumable poison pins, representative of a typical operating PWR The bumable poison pins were removed after one cycle, which is customary for Westinghouse units, in addition, one third of the fuel was removed after one cycle and replaced with fresh fuel. A second cycle is analyzed since there is interest in obsening the ability of SIMULA*IE 3 to accurately calculate the distribution within an assembly that resided near the core periphery in an earher cycle. Such assembbes usually have extremely large in'ra assembly exposure gradients and are generally beheved to be a hmiting case for pin re onstruction using a nodal code.

I Since the version of PDQ 7 used at Yan1ree Atomic does not include thermaShydraulic or Doppler feedback, these effects were neglected in the SIMUIAT&3 model as well. The purpose of t:Dminating these feedbacks from the SIMUIATL3 model was to m.Mntain as much consistency as possible between the models. Also, since the PDQ 7 model is two dunensional, the SIMULATE-3 analysia was conducted in two-dimensions.

I The first cycle was depleted to a cycle average exposure of 14.5 Gwd/Mt. A shuffle into the second loading pattem was made and this pattem was depleted to 3.0 Gwd/Mt. The I

second cycle was not fully depleted. After the first few exposure steps the power and exposure gradients are less pronounced and further depletion would not add any additional information I

than was estabushed in the first cycle comparisons.

We results of this PDQ 7 benchmark demonstrated the SIMUIA'IE 3 accuracy in predicting peak pin powers per assembly for all assemblies to be within 2.5% of the PDQ 7 calculated data for Cycle 1.

Figure 5.3 presents the distributions for the differences of the peak pin per assembly at beginning and end of Cycle 1.

ne difference in average assembly powers for the first cycle of depletion were less than 1%, with the worst assembly exhibiting a 1.6% difference I

68 -

l I

l

I in power compared to PDQ 7 as shown in Figures 5.4 and 5.5.

Die peak pin in the entire core versus exposure exhibited an average agreement to within 1.1% over the first cycle with no j

trend versus exposure. Table 5.4 summarizes the results of the comparison of the peak pin in the core versus exposure.

The Cycle 2 comparisons to PDQ 7 from beginning of cycle to 3.0 Gwd/Mt were similar to the Cycle 1 results in terms of assembly powers and peak pin. The peak pin per assembly and assembly power comparisons yielded much the same results as Cycle 1.

Figure 5.6 and 5.7 I

present the comparisons for the peak pin per assembly and assembly relative powers at the beginning of Cycle 2.

The codes again yield much the same results, despite large intra-assembly exposure gradients produced by the Cycle 1 depletion and shuffled to the interior of the core.

I Further evidence of the accuracy of the SIMULATE 3 method is illustrated by examining detailed intra assembly pin distributions at the beginning of Cycle 2.

The assembly containing the peak pin in the entire core at the beginning of Cycle 2 resided at the core periphery in the first cycle. The assembly is location number 28 in Figure 5.2, moved from location number 16 I

in Cycie 1, adjacent to the core periphery. At the beginning of Cycle 2, this assembly contains an exposure gradient from one edge of the assembly to the other that ranges from 5.5 Gwd/Mt to 14,0 Gwd/Mt.

The pin by pin comparison, of the pin relative powers. in the quadrant containing the peak pin, of this assembly, is presented in Figure 5.8.

As the figurv LUustrates SIMUIA'IE 3 and PDQ 7 yield similar results and predict the same peak pin kca!!on. Such good results are achieved since SIMULATE 3 tracks spectral history efects that influence intra assembly isotopics. Such efects are especially important near the core periphery, where the spectrum I

differs significantly from the infinite medium spectrum. Without properly accounting for this efect, macroscopic depleting models cannot accurately calculate the pin by pin distribution in sucn an assembly.

5.5 Pin Power Reconstruction Summaty The SIMUIATE 3 pin power distribution calculation has been verified against critical experiments, higher-order calculations and currently accepted methods for generating such i_

data. The results demonstrate SIMUIATE 3 to predict pin by pin powers to within 1% of critical l

69 -

( I I

I

+huents, hgher order calculations and cunendy accepted methods for producing such data.

I The results vahdate that accurate pin power distributions are achieved without the need for multiple assembly or colorset spectnun calculauons.

All spectrum data used in the I

SIMUIEE 3 models were infinite lattice, single assembly CASMO 30 calculations. Tracking of intra assembly spectral history effects in the pin power reconstruction technique is I

sufncient to prmide accurate local pin power predicuons.

This is demonstrated by comparison to pin by pin diffusion theory, in which the basic spectrum history effects are accounted for in the representation of local pinwise isotopics.

I I

I I

' I i I I

L I I

I I

70 -

I I

/

l I

1 TABLE 5.1 SIMULATE 3 PIN POWER VAllDATION FOR 'n{E B&W CRmCAL EXPERIMEN%

Core #

1 5

12 14 18 20 Fuel 2.46%

2.46%

4.02 %

4.02 %

4.02 %

4.02 %

i Annembly 15x15 15x15 15x15 15x15 18x16 18x16 Design 0 Gd 12 Gd 0 Gd 12 Gd 0 Gd 16 Gd I

RMS Dference 0.6%

1.5%

0.9%

1.2%

0.8%

1.2%

(S3 Meas)

Error in Peak 0.1%

1.1%

0.5%

0.4%

-0.8%

-0.1%

I Pin (S3 - Meas)

I I

1

' I I

I I

I

I I

71 -

I I

---,e,,

, ~. - -

I TABIE: 5.2 COLORSEP ANALYSIS CASE i_15rnNG AND IDEhMRS lattice Enrichment Initial Number of Contml Rods Number (w/o U 235)

Exposure Burnable Inserted Absorbers 1

2.4 Fresh 0

no 2.4 Burned 0

no I

2 3.1 Firsh 0

no 3.1 Burned 0

no 3

2.4 Fresh 0

no I

2.4 Fresh 12 no 4

3.1 Fresh 0

no 3.1 Firsh 24 no 5

2.4 Fresh 0

no 2.4 Bumed pulled 12' no 6

3.1 Fresh 0

yes 3.1 Fresh 0

no I

7 3.1 Fresh 0

yes 3.1 Burned 0

no 8

3.1 Pn:sh 0

no I

3.1 Burned 0

yes 9

3.1 Fresh 0

no 2.4 Fresh 0

no 10 3.1 Fresh 0

no 2.4 Fresh 12 no 11 3.1 Fresh 0

no 2.4 Fresh 0

yes 12 3.1 Fresh 0

no 2.4 Bumed 0

no I

13 3.1 Fresh 12 no 2.4 Burned 0

no 14 3.1 Fresh 0

no 2.4 Burned pulled 12' no 15 3.1 Fresh 12 no 2.4 Burned pulled 12' yes

' Removal of 12 burnable poison shims prior to depletion of lattice I

72 -

I I

. ~.. -.

I 1

I TABLE 5.3 COLORSLT VERIFICA710N RESULTS I

&VERACE OF ABSOLITIE VALUE IN PERCENT DIFFERENCE',

IN PINS FROM CASMO 30 COLORSET 1ATI1CE DEPLEIlONS i

i Lattice Peak Pin Worst Pin All Pins No.

Avg to Avg to RMSto

- i SIM 3 followed SIM 3 followed SIM 3 followed by PDQ 7 by PDQ 7 by PDQ 7 1,2

.345.351

.520.307

.176.127 1

.9051 201 1.581 233

.689i.120 i

3,4

.1951 186

.835A.226

.2841 072 g

.9841 205 1.971 871

.8341 367 1

' E l

5

.2701 279

.6401 299

.2071 093 7931.179 1.421 178

.659.103 I

6-8.11

.7381 235 3.15i.466

.9461 097 2.721 315 7.981 235 4.03.152 9.12

.3801 238

.6601 218

.225 091 1.241 141 1.781 323

.7281 116 10

.340i.272

.7801 358

.2691 109 1.221.163 2.31 1.10

.8701 354 l

13

.1001.094

.5901 238

.203.090 1.451 290 2.281 679

.799.144 1

14,15

.2701 244

.5601 245

.1961 108 l

lI 1.231 227 2.091.567

.8221 145 l

OVERALL

.3301 237

.9671 295

.3131 099 1.321.215 2.681 523 1.121 188 I

• % Difference = (SIMULNIE-3 or PDQ 7 mtnus CASMO-30) dMded by CASMO 30 times 100%

t

I I

73 -

I

I

~

I' I

TABLE 5.4 I

McOUIRE UhTP 2 CYC121 COMPARISON BETWEEN PDO 7 AND SIMUIATE 3 OF PEAK PIN Exposure Peak Pin Power

% Difl*

Annembly of the (Gwd/Mt)

S3 PDQ S3 PDQ R

0.15 1.349 1.334 1.14 37 37 0.5 1.341 1.323 1.32 37 37 1.0 1.332 1.312 1.52 20 20 I

2.0 1.344 1.323 1.59 4

4 3.0 1.324 1.305 1,41 4

4 4.0 1.305 1.292 1.02 4

4 5.0 1.290 1.276 1.05 37 37 I

6.0 1.281 1.268 1.02 37 37 l

7.0 1.275 1.263 0.96 37 37 8.0 1.268 1.256 0.95 37 37 I

9.0 1.263 1.252 0.90 37 37 10.0 1.257 1.246 0.91 37 37 11.0 1.252 1.240 0.97 37-Si

! g 12.0 1.246 1.233 1.05 37 37 g

13.0 1.238 1.225 1.06 37 37 14.0 1.230 1.216 1.15 37 37 14.5 1.225 1.210 1.24 57 37 Average Percent Difference of Peak Pin for Entire Cycle is 1.1 %

• % D1!Ierence is (SIMU1 ATE 3 PDQ)/SIMU1EIE 3

• 100%

I I

I

I LI lI 74 -

I I

. ~. _.. -

i I

I TIGURE 5.1 McGUIRE UhTP 2 CYCm 1 I AADING PATrEM AND NAMING CONVEhmON I

l I

1 2

3 4

5 6

7 8

l ADO B20 A00 B16 ADO B16 A00 C10 i

10 11 12 13 14 15 16 ADO B20 A00 B16 A00 C20 COO 19 20 21 22 23 24 A00 B16 A00 B16 ADO C10 28 29 30 31 32 A00 B20 A00 C12 COO 37 38 39 B00 B20 COO l

Assembly Number 45 46 Enrichment bps COO C09

.I 1

Enrichment w/o U235 I

A 2.1 B

2.6 C

3.1 I

'I I

I I

I I I

l I

I FIGURE 5.2 McOUIRE UhTI' 2 CYrlP 21 AADINO PATTERN I

AND NAMING CON \\ThT10N i

I t

23 1 62 83 24 4 5 37 6 77

" 8 A00 BR16 CR10 BR20 BR16 B00 A00 D00 I

16 10 15 11 46 12 22 13 11 14 " 15

" 16 COO CR20 CRO9 BR16 BR20 D04 D00 31 19 38 20 24 21 20 22 13 23

" 24 I

CR12 BR20 CR10 BR16 BR16 DOO 16 28 32 29 29 30 " 31

" 32 COO COO BR20 DO4 D00,

31 37 39 38

" 39 CR12 COO D00 Assembly Number 45 45

" 46 Enrichment bps COO D00 I

~.

Enrichment w/o U235 I

A 2.1 B

2.6 C

3.1 D

3.2 OFA NOTE: R DESIGNATES REMOVAL OF BP AFTER CYCLE 1 OPERNI1ON I

LI I

!I I

76 -

LI

'I

I I

FIGURE 5.3 I

McGUIRE UNTT 2 CYCLE 1 BOC AND EOC COMPARISON BE*IWEEN PDQ 7 AND SIMULATE-3 PEAK PIN BY ARREMBLY I

I

+0.19

+ 1.17

+1.12

+2.06

+1.22

+ 1.99

+ 1.58

+ 1.12

+ 1.38

+0.84

+ 1.36

+ 1.24

+ 1.27

+0.82

+0.97

+ 1.28 I

+ 1.06

+1.73

+ 1.70

+2.05

+ 1.87

+ 1.38

+ 1.10

+ 1.35

+1.13

+ 1.46

+ 1.01

+1.17

+ 1.76

+ 1.54

+ 1.77

+ 1.88

+1.84

+ 1.77

+ 1.23

+ 1.01 I

+ 1.52

+ 1.19

+1.14

+0.86

+0.85

+ 1.26

+ 1.75

+ 1.73

+ 1.15

+0.74

+0.03

+ 1,.1 1

+0.66 +0.73

+ 1.89

+2.51

+ 1.14

+ 1.42

+0.25

+1.14

+0.28

+ 1.93 BOC SIMULATE 3 Peak Pin Power vs PDQ 7

+1.04

+ 1.39

- !(S3 - PDQ)/S3)

~

EOC SIMCIATE 3 Peak Pin Power vs PDQ 7

+0.91

+2.14

• 100%

I BOC (15 Gwd/Mt) RMS Difference = 1,44 EOC (14.0 Owd/Mt) RMS Difference = 1.33 I

I I

I I

I 77 -

I I

- _. ~..

9 I

i FIGURE 5.4 4

I McotHRE UNTP 2 CYCLE 1 BOC.15 Owd/Mt COMPARISON BMTEN PDO-7 AND SIMUIEIE 3 OF ASSEMBLY POWER I

I 1.093 1.014 1.159 1.169 1.240 1.123 1.019 0.706

+0.29 1.86

+0.89

-0.13

+ 1.53 0.14

+0.82 1.21 1.120 1.065 1.223 1.172 1.168 1.109 0.778 I

+0.55 1.17

+ 1.34 0.01

+ 1.19 1.21

+0.08 1.195 1.163 1.200 1.086 0.980 0.649 I

+1.12 0.21

+ 1.24 0.51

+0.46 1.46 1.199 1.058 1.089 0.971 0.554

+1.17 1.08

+0.69 1.29 0.99 1.244 0.908 0.832

+0.57 1.18 0.lo I

l l

SIMU1A'IE-3 Assembly Power 0.994 0.482 l

((S3 PDQ)/S3)*100

+0.59 0.42 1E j

RMS Difference = 0.96 I

lI I

I I

q I I I

I I

FIGURE 5.5 I

McGUIRE UhTP 2 CYCLE 1 EOC 14.0 Gwd/Mt COMPARISON BNN PDQ-7 AND SIMUIEIE-3 OF ASSEMBLY POWTR I

I 1.063 1.107 1.000 1.115 1.071 1.126 1.011 0.786

+0.62

+0.51

+0.60

+0.82

+0.58

+0.35 0.20 0.63 1.061 1.107 1.065 1.126 1.074 1.108 0.810 8

+0.55

+0.57

+0.64

+0.63 4.18 4.22

-0.37 1.065 1.126 1.082 1.123 0.981 0.733

-I

+0.61

+0.70

+0.43

+0.21

-0.46 1.02 1.088 1.142 1.062 1.015 0.608

+0.42

+0.10

-0.24

-0.97

-0.73 1.175 1.056 0.863

+0.06 0.89 0.61 SIMU1WIE 3 Assembly Power 1.010 0.572 0.71 1.59

((S3 PDQ)/S3)*100%

I RMS DIEerence = 0.64 I

I:

I I

I I

79 -

I I

I i I I

FIOURE 5.6 McGUIRE UNTP 2 CYCII 2 BOC.15 Gwd/Mt COMPARISON BElWEEN PDO 7 AND SIMUIA'IT 3 OF ASREMBLY PO%TR I

I 0.882 1.037 1.311 1.058 0.830 0.726 0.730 0.801 0.83 0.56 0.31

-0.35

-0.36 0.44 0.32

+0.59 1.031 1.245 1,267 1.333 0.929 0.798 1.051 0.824 0.36 0.58

-0.45

-0.05

+0.19

-0.43

+0.29

+0.37 1.294 1.252 1.300 1.212 1.240 0.866 0.818 0.766 0.23 0.46

+0.32

+0.90

+0.47 0.16 0.33

+0.12 I

1.048 1.321 1.198 1.373 1.351 0.980 1.054 0.604 0.40 0.06

+0.87

+0.56

+0.42 0.01 0.10 0.73 I

0.826 0.925 1.232 1.345 1.226 1.102 0.923 0.46

+0.17

+0.55

+0.43

+0.30 0.03

-0.40 0.729 0.803 0.869 0.984 1.113 0.893 0.595 I

0.41 0.46

-0.16 0.07 0.03 0.41 0.80 0.761 1.071 0.827 1.064 0.936 0.605

-0.28

+0.33 0.31 0.07 0.28

-0.68 0.822 0.842 0.777 0.611 -- SIMU1 ATE 3 ASSEMBLY POWER

+0.65

+0.55

+0.21 0.64

- (S3 PDQ)/S3

• 100 RMS DitTerence =.44 I

I I

I I

80 -

I

I I

F10URE 5.7 I

McOUIRE UhTP 2 CYCLE 2 BOC.15 Owd/Mt COMPARISON BEIWFTN PDO.7 AND SIMUIATT 3 PEAK PIN BY ASSEMBLY I

0.905 1.148 1.390 1.183 0.941 0.762 0.797 1.018 0.13

+0.08 0.50

+0.35

+0.04

+ 1.65

+ 1.79

+ 1.48 l

1.138 1.339 1.370 1.464 1.099 0.874 1.171 1.080

+0.33 021 0.40

+ 1.32

+ 1.24

+0.37

+0;92

+0.82 I

1.374 1.355 1.400 1.318 1.402 0.999 0.883 1.012 0.45 0.48

+ 1.13

+ 2.11

+0.95 40.00

-0.01

+ 1.24 I

0.171 1.449 1.307 1.401 1.454 1.101 1.207 0.940

+0.35

+ 1.28

+2.12

+0.4 9

+ 1.72

+0.99

+0.59

+0.69 0.935 1.092 1.386 1.437 1.324 1.208 1.211 I

0.01

+ 1.26

+2.09

+ 1.68

+ 1,64

+0.08

+0.77 0.733 0.875 0.999 1.106 1.216 1.096 0.955

+ 1.71

+0.18

+0.60

+ 1.07

+0.11

+ 1.41

+0.67 0.833 1.194 0.891 1.218 1.225 0.975

+0.62

+ 1.03

+0.04

+0.66

+0.81

+0.67 1.040 1.103 1.027 0950 SIMULATE 3 Peak Pin I

+1.54

+0.94

+1.30

+0.70 --- (S3 PDQ)/S3

• 100 RMS Difference = 1.047 I

I I

I E

81 -

l-l

l FIGURE 5.8 I

McGUIRE UNIT 2 CYCLE 2 BOC ( 15 Owd/hm i

COMPARISON BIm#EEN PDQ 7 AND SIMUIXIE 3 PLN DISTRIBLTTION OF UPPER f rFT QUADRAST OF ASSEhmIX 28 WHICH CONTAINS PEAK PIN I

l.

1.382 1.383 1.382 1.381 1.373 1.361 1.350 1.343 1.345 I

1.14 -0.86 0.86 0.86 0.79 0.73 0.59 0.37

+0.07 1.414 1.394 1.394 1.418 1.387 1.363 1.345 1.333 1.337 0.91 1.69 1.69 0.56 1.70 1.02 0.06 0.82

-0.07 guide 1.430 1.432 guide 1.445 1.420 1.373 1.342 1.341 tube 0.28 0.28 tube 0.00 0.21 1.29 0.67 0.00 I

1.431 1.411 1.414 1.449 2.462 guide 1.419 1.357 1.351

-0.07 0.91 0.91

+0.28 +0.48 tube

+0.28 0.51

+0.22 1.434 1.415 1.416 1.450 1.440 1.460 1.444 1.383 1.362 I

+0.42 0.35 0.49

+0.69 0.28

+0.97

+0.98 0.50

+0.44 guide 1.440 1.441 guide 1.447 1.445 guide 1.411 1.369 I

tube

+ 1.19

+1.12 tube

+ 1.05

+ 1.05 tube

+0.93

,0.74 1.430 1.409 1.408 1.436 1.408 1.405 1.423 1.382 1.365

+ 1.20 +0.28

+0.07

+ 1.34 0.00

-0.07

+1.14

-0.14

+0.74 1.422 1,400 1.399 1.454 1.397 1.392 1.412 1.374 1.360

+ 1.28 +0.21

+0.07

+1.14

-0.07 0.22

+ 1.00 0.29

+0.59 l

guide 1.410 1.409 guide 1.408 1.405 guide 1.389 1.359

-S3 l

tube

+0.93

+0.79 tube

+0.72

+0.72 tube

+0.73

+0.74

- % Diff

% Difference = ((SS PDQ)/PDQ)'100%

IE RMS Difference of all Pins = 0.79 l 5 Difference of Peak Pin = 0.48 Peak Pin 14 cation is Italicized and Bolded I

I

,I I

6.0

SUMMARY

AND CONCLUSIONS This report focused upon vahdating the SIMU1 ATE 3 code for three major categories: PWR appheation. BWR application, and detailed pin power reconstruction for PWRs.

Key physics parameters were compared to plant measured data and higher order calculations.

I' operating PWRs. his analysis encompassed seven cycles of operation. Table 0.1 presents a summary of the typical accuracy of the SIMULAE 3 code in preditttrig the measured parameters. As the table shows, the SIMULAE 3 piedicts these parameters to a high degree of acCuraty.

I The second category presented vahdation of the code versus plant measured data for two operating BWRs. nts analysis encompassed eight cycles of operaticn. Table 6.2 presents a summary of the accumey SIMULAE 3 achieved in predicting measured parameters. As the table shows SIMULATE 3 predicts these parameters to a high degree of accuracy.

I The final category presented validation of SIMUIAE 3 for the purposes of calculating detailed pin by pin relative power distributions.

Wis capability was vertfled by comparing SIMUIAE 3 calculated data to higher onler transport theory solutions as well as to the currendy accepted method of generating such data. PDQ 7.

ne analysis demonstrated Otat SIMULATE 3 calculated the peak pin to an accuracy of within 1% of the reference transport theory solution, this was within the same level of accuracy as PDQ 7.

ne pin power capability was also compared between SIMUIATE 3 and PDQ 7 for a quarter I

core depletion model of McGuirr Unit 2 for two cycles of operatkan. Compared to PDQ 7 the results of this study demonstrate that SIMUIATE 3 predicts the peak pin in Ole entire core to within 1.1%.

No trends existed with cycle exposure. Recycled assemblies with large intra-assembly exposutt gradients were also predicted to the same degree of accuracy. Table 6.3 presents a summary of the pin power reconstruction analysis.

By virtue of the analysis presented in this report it is concluded that the SIMUIAE 3 code, coupled with spectrum data produced by the CASMO-30 code, is acceptable as a spatial incore reactor phpics analysis model.

The code is acceptable for the performance of all I I

t I

t calculauona, including detailed pin by pin power neonstniction for PWRs, that are curren0y i

conducted with such codes as PDQ 7 and SD4UlWIE 2.

Furthennore, it is concluded that the code perfonns to a level of accuracy sumelent for the perfonnance of reload physics analysis for heensing appbcations, i

i I

i I

I lI l

1 l

i B

I 4

I 84 -

I I

l_

I TABEE 6.1 suulwARr_or stuutATE-3 ACCURACY FOR PWR APPf) CATIONS

]

j PIArtr PARAMETg PQDm5 CYCIES ACCURACY N

McGuire HFP Critical Baron 50 3

13 ppm io 10 ppm 36 ppm 48 4

12 ppm _wr 9 ppm 38 ppm f

Farley 1

j McGuire EOC HFP Critical 2

2 7 ppm 11 ppm Farky Boron 4

4 4 ppm 8 ppm McGuire HFP Detector Fission 19 flux maps 3

1.5% _wr O.3%

5.8%

Farley Rates 15 flux maps 4

1.0% _wr O.2%

2.9%

McGuire HFP Assembly Rdative 19 flux maps 3

1.5% io 0.3%

4.0%

Farley Power 15 flux maps 4

1/3% w_r O.2%

4.6%

McGuire HFP Detector F1ssion 19 Dux maps 3

1.3a% do 1.2a%

3.4a%

m Farley Rate - Axial Offset 15 flux maps 4

1.0a% _wr O.8a%

3.6a%

McGuire IIFP Xenon "nansient 61 1

1.2a% _wr 1.08%

3.68%

Axial Offset McGuire HZP. BOC Co.4m1 11 3

3.5% _wr 2.7%

9.8%

Farky Rod Worths (pcm) 16 4

3.2% _wr 2.5%

9.8%

McGuire IIZP. BOC Critical 14 3

21 ppm _wr 17 ppm 59 ppm Boron vs. Rods McGuire HFP Isothermal 3

3 I.6 o pc m / T 3.9 apcm/T Temp. CoefBctent McGuire HZP.BOC Isothermal 5

3 1.2 Apon/T 1.5 opcm/7 Farley Temp. Cwn+..:

7 4

0.9 epcm/T 1.2 apem/T

~.----..s

-n,n.

-a,.-----.,+.a..wa.n---

._-_--.-n--.--.-

-.-.au..-.

.. ~. -

n-.w~>

a--.a.,,

9 1

I 1.

I 4

I 5,

e,

'0I l

gl-ilI n

lI l la m

i g

I jl lli[aJ I

~

Is I

I 1i i

,I 86 l.

l-LI-

-M M

M M

M M'

M M

M.

M M

'M' M

'M M

M

. TABLE 6.2

SUMMARY

OF SIMUIATE-3 ACCURACY FDR BWR APPUCATIONS l

PIAlfr PARAMEIER POUfIS CYCLES ACCURACY WORST l

Quad Cttles Hot Eigenvalues

-15 (F4 Xenon) 2 1.00472 g 0.00072 1.00650 Quad Cities 29 (All data) 2 1.00415 g 0.00214 0.99909 Foisk 109

'6 1.00429 y 0.00106

- N/A -

I Quad Cities Cold In-Sequence 7

2 0.99586 8 0.00071 0.99479 f

Forsmark Criticals 37 6

0.99857 30 0.00108

- N/A ~

[

Quad Cities BOC Cold Imal 10 1

0.99829 y 0.00053 0.99913 Criticals i

Quad Cities Integrated Axial 27 flux ruaps 2

5.46% y of 5.42%

13.8 %

Fommark

" IIP Traces 109 flux maps 6

4% - 7% y of 3%

7.4%

l

[

Forsmark Integrated Gamma 13 ilux maps 3.9%

4.3%

9 TIP 'IYaces

)

e

• The Forsmark data (Ru'mumt 24) did not contain deta9ed information with regards to these pmanders i

1 l

l l

I i

(*

1 Oi i

I I

l t.u

'W--

W E

'W E.

M E

E E: M

' W W

W l

l l

TABIE 6.3

SUMMARY

OF SIMUIATE-3 ACCURACY PWR PIN POWER RECONSIRUCI1ON 1

PIANT PARAMEIER POINIS ACCURACY W ORST i

e i

B&W Criticals Peak Pin 6 Cores 1.1%

2.9%

Entire Lattice Colorsets Peak Pin 150 0.3%

0.7%

l Entire Lattice Quarter Core Assernbly Power 20 Statepoints O.7%

1.4% -

PDQ-7 Quarter Core Peak Pin per Assernbly 20x62 assernblies 1.2%

2.1%

PDQ-7 l

Quarter Com Peak Pin Entire Core 20 1.1%

1.0%

i PDQ-7 f

k l

i l

t l

.t I

i I

=

I 7.0 KEFERENCES 1.

VerPlanck, D. M.,

Smith, K. S., and Umbarger, J.

A., " SIMULATE 3P Advanced Three Dimensional Two Group Reactor Analysis Code "

Studsvik/SOA 88/01, I

February,1988.

2.

VerPlanck, D. M.,

• SIMULATE 2: A Nodal Core Analysis Program for Light Water Reactors," EPRI Research Project 710 1, 1982.

3.

Cadwell, W. R., "PDQ 7 Reference Manual," WAPD TM 678,1967.

4.

DIOlovine, A. S., et al, "CASMO 30 Validation," YAEC 1363 April,1988.

5.

Edenius, M., Ahlin, A. and Haggblom, H., "CASMO 3 A Fuel Assembly Burnup Program, "Studovik/NFA 86/7, November 1986, Studsvik Energitenik,1986.

6.

Carew, J. F, to Cacciapouti, R. J. " Yankee Atomic Electric Company Calculation of the BNL PWR Core Standard Problem," letter from Brookhaven National Laboratory,

, I dated July 29,1988.

l 7.

Smith, K. S. et al., " SIMULATE-3: The Studsvik Steady State Nodal Reactor Analysis Code," Studsvik/SOA 88/03, August,1988.

8.

Smith, K.

S., Rempe, K.

R., ' Testing and Applications of the QPANDA Nodal Model," Proceedings International Meeting on Advances in Reactor Physics and Computation, Volume 2, p. 861, Paris, France, April,1987.

9.

Ver Planck, D. M., et al., ' TABLES 3P Library Preparation Code for SIMULATE 3P "

Users Manual Version 2.0, Studsvik /SOA 88/02, February,1988.

10. Ver Planck, D. M., " TABLES 2 Manual", YAEC 1391P, April,1983, l -

11. Smith, K. S., Rempe, K. R., " SIMULATE 3 Pin Power Reconstruction: Methodology and Benchmarking," accepted for publication, International Reactor Physics Conference, Jackson Hole, Wyoming September,1988.

12. Denver, D. J., et al., " Application of Yankee's Reactor Physics Methods to Maine Yankee," YAEC 1115. October 1976.

l

13. Smith, K.

S.,

Ver Planck, D.

M., "SS TESTER A SIMULATE-3 Installation and Verification Benchmark Problem Series," Studsvik/SOA 87/13 May 1987.

14. Smith, K. S., Ver Planck, D. M., "KWU PWR SIMULATE 3 Solution to the KWU Two Cycle PWR Depletion Benchmark Problem," Studsvik/SOA 87/14. Rev B May, 1987.

15. DiGiovine, A. S., et al., "McGuire Unit 2 SIMULATE 3 Benchmark Analysis Cycles 1 through 3," YAEC 1608, October 1987, 1

':E

16. Smith, K.

S., et al., " SIMULATE-3 PWR Benchmark Report Farley Unit II,"

3 Studsvik/SOA 87/10, April 1987,

17. Personal Communication with K. S. Smith, Studsvik of America.

I I I

I

18. Larsen, N. H., et al., " Core Design and Operating Data for Cycles 1 and 2 of Quad Cities 1," EPRI report EPRI NP 240, November 1976.

19. larsen, N. H., et al., " Core Design and Operating Data for Quad Cities 1 Cycle 3." EPRI report EPRI NP 552, March 1983.

20. Tennessee Valley Authority, 'Ver1Dcation of WA Steady State BWR Physics Methods,"

Topical Repon, January 1979.

21, Ver Planck, D. M., " Methods for the Analysis of Boiling Water Reactors. Steady State Core Physics," YAEC 1238, March 1981.

22. Ahlin, A.,

Edenius, M.,

Gragg, C., "MICBURN 3 - Microscopic Burnup in Burnable

)

'I Absorber Rods," Studsvik/NFA 86/26 November,1986, i

23. Wochlke, R. A., et al., ' Vermont Yankee Cycle 8 Sununary Report," YAEC-1305, August, I

1982 i

l l-

24. Hakansson, H., et al., " SIMULATE 3 BWR Benchmark, Forsmark Unit 1 Cycles 1 through i

6," Studsvik/NFA 88/44 Revised, July 25,1988.

25. Smith, K. S., "SIMUIATE 3 Pin Power Reconstruction: Benchmarking Against the B&W Critical Experiments," Trans. Am. Nucl. Soc., Volume 56, p. 531, San Diego, Ca., June, i

I 1988.

i 1

I I

I I-I I

I I

I IL APPENDIX A I

As part of the SIMULATE 3 validation, the Brookhaven National Laboratory (BNL)

PWR Core Standard Problem was analyzed. The problem was designed for the express I

purpose of testing the validity of a particular code or code package applied to typical reload physics calculations, since this document is a validation of the SIMULATE 3 code, the problem was analyzed with SIMULATE 3, and the results were transmitted to BNL. BNL evaluated these results versus their reference solution, and their evaluation 1s summarized in this Appendix.

The BNL PWR Core Standard Problem consists of analyzing a typical PWR. This PWR includes several fuel types: both with and without burnable poison. On order to analyze this problem several physics calculations are required. They include: cycle I

depletion analysis at Hot Full Power (HFP) conditions, calculation of control bank worths at HFP and Hot Zero Power (HZP), soluble boron concentration at HFP and HZP

]

conditions, and reactivity coefficients at HFP and HZP.

Cross section data for the different fuel designs was calculated using'CASMO 3G in a manner consistent with the analyses of both the McGuire and Farley units. The I

Core Standard Problem calculations were performed using SIMULATE 3, Again, in a manner consistent with the analysis of the McGuire and Farley units.

Table A.1 presents a list of the parameters that were requir'ed to be calculated as part of the Core Standard Problem. As the data in the table demonstrates, a variety of physics calculations are required in order to provide this data. Table A.2 presents a summary of the evaluation of the results Yankee Atomic calculated for these parameters versus the BNL reference solution.

BNL concluded upon evaluating Yankee Atomics results that "In summary, the

~ NL/YAEC depletion and reactivity cal.:ulations are found to be in generally good B

a'greement, and agree to within th', expected uncertainty of the BNL reference I

solution."

I A-1

!I

,I

I TABLE A.1 BNL PWR CORE STANDARD PROBLEM CALCULATED PARAMETERS CORE DEPLETION CALCULATIONS I

Provide the following quantities at beginning, middle and end of cycle at HFP conditions.

.[

Core average assembly wise radial power distribution Core average assembly wise radial exposure distribution Core average axial nodal power distribution

=

Axial Offset Overall three dimensional peaking factor

=

Critical boron concentration I

-h CORE REACTIVITY CALCULATIONS Provide the following core reactivity coefficients, multiplication factors and critical boron concentrations at the given conditions.

Xenon free power defect at HFP and HZP at BOL and EOL Equilibrium xenon power defect at HZP and HZP at BOL and EOL Subcritical boron (Keff=0.95) at HZP and BOL Differential control bank worth at HZP and BOL Integral control bank worth at HZP and BOL Moderator coefficient at HFP at BOL and EOL Doppler defect at BOL g

I I

I I

g

=

A-2 I

I TABLE A.2 BNL PWR CORE STANDARD PROBLEM BNL/YAEC COMPARISON RESULTS I

CORE DEPLETION CALCULATIONS Core average radial power distribution Exposure RMS % Difference

-= I BOL MOL EOL

<2 Core average radial exposure distribution I

Exoosure RMS % Difference BOL MOL EOL

<2 Core average axial power distribution Exoosure RMS % Difference BOL MOL EOL

<2 Power Defect BOL (no Xe)

I BOL (eq Xe)

EOL (no Xe)

EOL (eg Xc)

< 15 Subcritical Boron BOL (keff=0,95)

<2 Doppler Defect BOL, 50% power

'I 75% power 100% power

< 15 MTC BOL

< 3 pcm/*F Differential Rod Worth

< 25 Integral Worth

< 20 I

I A3 I

I I

APPENDIX B RESPONSES TO REQUEST FOR ADDITIONAL INFORMATION CONCERNING YAEC-1659, SIMULATE-3 VALIDATION AND VERIFICATION QUESTIONS AND ANSWERS I

1) Q. The comparison of the measured and calculated axial ofisets for the McGuire Unit 2 end-of-cycle 3 (EOC-3) xenon transient (Figure 3.5)

'I.

shows that SIMULATE-3 significantly overpredicts the observed axial flux diiforence. This EOC-3 discrepancy has been attributed to the lack of a true equilibrium state-point where the transient began. To

.l further determine the cause of the apparent discrepancy, provide E

sensitivities of the amplitude of the xenon oscillation and stability index with respect to: (a) flux convergence, (b) source convergence, (c) thermal-hydraulic convergence, (d) doppler feedback, (e) time I

step size (f) spatial nesh (12 vs. 24 axial avdes), and (g) Bank-D r

worth.

Also provide details of the initial core conditions inmediately prior

'l to the McGuire xenon transient, including the core power, core flow n

and inlet coolant temperature.

A. Figure 3.5 of YAEC-1659 shows measured and calculated data for two I'

McGuire Unit 2 xenon transients in Cycle 3, one at middle-of-cycle (MOC) on the top of the page, and one at end-of-cycle (EOC) on the l'

bottom of the page. The data for these plots were also provided in pl-l YAEC-1659 as Tables 3.8 and 3.9, respectively. Upon review af the l W figure and tables, it is evident that the plotted data in the EOC l

figure does not correspond to Table 3.9 and is in error. A revised g

Figure 3.5 is attached which has the correct data from Tables 3.8 and i

g 3.9 of YAEC-1659.

The revised figure does not show a significant overprediction of the observed axial flux difference. The minor differences observed I.

at EOC are partially accountable to the lack of a true equilibrium condition before the transient. Notice in the HOC data that there is effectively no change in measured axial flux difference (-4.1 to-

-4.4%) and no rod movement before the transient (from 0 through 19 I

hours). At EOC, there is a larger measured offset range (-3.9 to

-4.5%)

just prior to the transient, accompanied by a small ud withdrawal at 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. This is a non-equilibrium situation and I.

results in larger differences between measurement and calculation, l

both before and during the transient. As a result, the overall axial offset prediction at EOC is slightly less accurate than that at'MOC.

The McGuire data consisttd of rod position, axial flux I

difference and power level on an hourly basis. Power was effectively full power, with a lowest power level of 99.2%. No data on core flow 1

or coolant inlet temperature variations was available and these i

parameters were assumed to remain constant during the transient. No data prior to the transient was available. Considering the f act tttt the data is only provided hourly and is incomplete, the comparisons I

are considered good and represent the relative accuracy achievable when trying to predict actual plant axial of f set, with continuous changes in rod motion, power and temperature conditions. This is quite distinct from the use of xenon transients in licensing B-1

I i

applications, which is discussed later.

While the correction of Figurc 3.5 obviates the remainder of the question, an evaluation of the sensitivities of the model I

demonstrates its validity. To increase the impact of the sensitivity to the parameters, a more severe xenon transient is chosen with greater changes in axial offset than the McGuire transients. This' transient is an MOC case from equilibrium, full-power conditions. The

I; transient consists of instantaneous rod insertion to 50% and power reduction to 50% power for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, followed by instantaneous rod

' withdrawal and a return to 100% power. A reference case and sensitivity cases were_run with the latest version of SIMULATE-3.

The following comments are relevant concerning the axial of f set changes. resulting from the changes in each of the following parameters:

The acceptance criterion for flux a)

Flux Convergence convergence was tightened by an order of magnitude (0.00005 to I.-

0.000005). No change in axial offset was found.

I b)

Source Convergence The acceptance criterion for source convergence was tightened by an order of magnitude (0.00005 to 0.000005). No change in axial offset was found.

The number of hydraulic c)

Thermal-Hydraulic Convergence iterations was increased by 50% (20 to 30). No change in axial offset was found.

Two cases were run to study - doppler I

d)

Doppler Feedback effects. In the first, the core average fuel temperature was increased by 100'K and, in the second, it was decreased by 100'K from the nominal value of 968'K. Doppler feedback is maintained in-all cases and the linear relationship _between local power and local fuel temperature is unchanged. The effect of fuel temperature change is shown in Figure 1. Notice that there is a change-in axial offset and the time dependence of ' the I

oscillation.

e)

Time Step Size - The time step size was reduced by 50% (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to 1/2 hour). Hinor changes in axial offsets were observed, as seen Figure 2. The relatively small effect of time step size results from the fact that the average of the beginning and end of step powers is 3ssumed in-the-SIMULATE-3 calculationc f)

Spatial Mesh - The number of axial nodes was doubled (12 to

24). Again, only slight changes in axial of f sets were observed, as seen Figure 3. SIMULATE-3 is less sensitive to spatial mesh than the previous generation of nodal codes since there is a true intranodal flux solution.

,I Measured rod worths are well predicted by g)

Bank D Worth SIMULATE-3 in YAEC-1659 for McGuire and Farley. Considering the fact that there is little rod insertion in the McGuire lI B-2 I

I i

transients, bank worth-is not considered a significant factor.

SIMULATE-3 utilizes a fourth-order polynomial representation of the intranodal' flux solution in both the fast and thermal groups. As I'

such, its sensitivities to-convergence parameters and geometry are less than that of simple nodal solutions using constant nodal flux assumptions.

2) Q. Provide the following iodine and xenon constants used in the

-SIMULATE-3 analysis: fission yields, decay constants and microscopic I

absorption cross sections.

A. These parameters are provided in Table 1.

3) Q. What numerical approximations are used in the determination of the xenon buildup and depletion?

Zndicate the effect of these approximations on the calculated xenon concentration and power axial offset.

I A. The SIMULATE-3 solution to the xenon buildup and depletion equations makes the following assumptions:

o The iodine production rate and xenon destruction equations are solved analytically using the average of the beginning and end of step core power levels during the time step interval selected.

I o

The spatial distribution of the neutron flux used in the buildup and destruction equations is assumed to be unchanged during the: time interval selected.

The end of step flux

_I distribution is used, not an averaged distribution. The impact of this assumption can be seen by varying the time step interval, as shown in Figures 2 and 4, which contain the axial I

offset and xenon concentrations, respectively, for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 1/2 hour time step cases.

I o

All iodine decays into xenon (i.e.,

iodine has a neutron absorption cross section of zero).

Xenon absorption is prcportional to the thermal flux (i.e, an I

o equivalent thermal absorption cross section is defined such that the total xenon absorption rate is preserved.)

I

4) Q. Rave radial power oscillations been observed or calculated for the two (Figure 3.5) McGuire transients? If so, provide core maps showing typical radial powers.

A. It is not known for the McGuire units whether radial power oscillations have been observed or calculated. Westinghouse provides I

I

.B-3 I

... ~.

lJ l

45 :

the following general statement in RESAR-3 concerning radial power oscillations. These statements have also found to be true for l

operation of our Maine Yankee plant:

"The core is designed so that diametral and azimuthal oscillations due to spatial xenon effects are self-damping and 1

no operator action or control action is required to suppress c

them. The stability to diametral' oscillations is so great that this excitation is highly improbable. Convergent azimuthal g

oscillations can be excited by prohibited motion of individual

{g control rods. Such oscillations are readily observable and lI alarmed, using the excore long Lon chambers. Indications are i

~

also continuously available from the incore thermocouples and I.

loop temperature measurements. Noveable incore detectors can be activated to provide more detailed information.

In all

[

presently proposed cores these horizontal plane oscillations are self-damping by virtue of reactivity feedback etfects j gg designed into the core."

5) Q. Does. the EOC-3 (F1gure. 3.5) comparison suggest a systematic U*

underestimate of the core axial offset by the measurement system? If what is the effect of this underestimate on the core operating s o, margins?

A. Figure 3.5 of YAEC-1659 was found to be in error and has been-replaced. The new figure shows no systematic error.

I

6) Q. What is the effect of the SIMULATE-3 overprediction of the observed xenon oscillation amplitude (Figure 3.5) on the intended core reload safety analyses?

A. As indicated previously, Figure 3.5 of YAEC-1659 has been replaced.

i l3 The new figure shows no systematic error.

g Free-running xenon oscillations are typically used in the core I

lidenring process. The oscillations are a mechanism to generate a spectrum of power distributions to span the large range of axial l

offsets required for trip limit and setpoint evaluation.

The l 3 oscillations are excited by depletion with heavy rod insertion to a high xenon / iodine imbalance and correspondingly bottom-peaked power shape..With doppler feedbacks purposely omitted, instantaneous-rod I

withdrawal in this situation -leads to a divergent, free-running oscillation. The xenon distributions saved from these oscillations are used in calculations at selected power level and rod insertion combinations with appropriate feedback effects. These calculations allow us to infer a. conservative total peaking for the-particular power level and rod insertion combination as a function of the axial offset indication of the excore detec$ ors.

' I' This process has been demonstrated to produce total peaking l

f actors versus axial offset which are conservative relative to those seen under normal operation in which axial offset is controlled. The licensing requirement of the process is to provide conservative total I

B-4

I peaking as a function of axial offset. Following the axial offset history of a particular xenon oscillation in time, with changing power level, temperature and rodded conditions, is a more difficult test of the core modelling and provides an added level of confidence-I in the model accuracy.

7) Q. The xenon transients presented in Figure 3.5 are relatively mild when compared to design transients with respect to angnitude and rate of bank reactivity insertion.

How would SIMULATE-3 be expected'to compare with measurement for the stronger design transients?

A. As indicated previously, Figure 3.5 of 'iAEC-1659 has been replaced.

The new figure shows no systematic error. Since this is the case, it

-g is expected that there is no significant systematic error for g-stronger transients.

Licensing applications,

however, rely upon a series of

'[

benchmarks, comparisons and conservative assumptions to assure that E

applications to conditions under which measurements are not available will not be affected by minor trends in predictive capability. For instance, for bank withdrawal transients, toe following factors I

provide assurance that the licensing analysis as conservative:

o Bank worths are measured at the beginning of each cycle to

' I'<

assure that both the ratt of reactivity insertion with rod motion and the total reactivity insertion in the licensing analysis is conservative.

o i'he maximum rate of bank withdrawal assumed in the licensing analysis is conservative relative to that achievable by the hardware at the plant.

Reactivity coefficients for moderator temperature, doppler and 1

o power are measured at the beginning of each cycle and during the cycle to assure that the values assumed in the licensing I

analysis are conservative.

The total peaking associated with bank withdrawal is the most o

I adverse of the possible axial off set conditions, accounting for axial offset and excore system measurement uncertainties. The total peaking is conservative based on its generation from a free-running xenon oscillation technique, as described above.

Core parameters are conservatively represented with appropriate o

uncertainties in initial core conditions and ' kinetics-parameters, Benchmarking to tests and higher order analysis is performed o

. E.

to demonstrate that the core power response modelling used in

'E the licensing calculations is conservative.

L LI B-5 L -

I Table 1 Yields, Decay Constants and Microscopic Absorption Cross Sections for Iodine, Xenon, Promethium and Samarium

---

Yields --------------------

Nuclide Iodine Xenon Promethium U-235 0.06288-0.00248 0.01080 U-236 0.06133-0.00172 0.01347

-U-238 0.06934 0.00027 0.01618 Pu-239 0.06400 0.01180 0.01223 Pu-240 0.06514 0.00772 0.01330 Pu-241-0.07001 0.00232 0.01473 Pu-242 0.07036 0.00236 0.01586 L

L

---

Decay Constant (seconds **)

Iodine Xenon Promethium 0.00002924 0.00002100 0.000003626 Microscopic Therma 1' Absorption Cross Section (10' barns) at 0 MWD /MT Enrichmont I

(w/o U-235)

Xenon Samarium l

1.60 154.347 4.61859 2.40 143.024 4.32364 3.10 136.137 4.14064 I

l I

I I

I B-6 li

I

(

FIGURE 3.5 McGUIRE UNIT 2 CYCLE 3 XENON TRANSIENT

' I MOC Transient 20 220 I

18 d -

6-6 6

Aa o a

^^A 6

g 6

4^^^^0

- 200 16 -

66 3

4 Aa4

- 180 12 [

10 -

- 160 6-140j I-g 4~

g o SIMULATE-3 Data

~ s

- 120 [

2~

o D Measured Data

~

.E W.

^ sank D position -

_100 e

{

B C

4

- 80 g E

6

(

8-I 10 -

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12 -

40 I; 16 -

- 20 18 -

0 20-i i

i i

i i

i i

i i

i i

i 0 2 4 6 8 10 12 14 16 18 20 22 24 26 20 30 32 S4 36 38 40 42 44 Time (hrs)

EOC Transient 20-220 1

d' O

O 0

000000 A AAAA 4 6 Aaa6

- 200 3 3 3 I4 [

66

- 180 12 10 -

- 160 I

8-6-

- 140 $

4~

g o SIMULATE 3 Data

- 120 [

2-o O Measured Data g

,0 -

a Bank D Position O

- 100,

E c

_.4 2.

- 80 g

,I k

6-T 8-

- 60 10 -

-g 12

- 40 31 14 - 20

'g 0 g

20 i

i i

i i

i i

i i

i i

i i

i i

i

'O 2 4 6 8 10 12 14 16 18 20 22 24 26 28 30 32 34 36 38 40 42 44 Time (hrs)

B-7 I

l';!

m l

l W

m W.

02

' m 8

3' 1

M

-7~

F 6

M 1

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1 tnte n M

e isi n s s a nt 2

)

r ac s

1 r

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f oE

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+

not t

nnk m

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0 x

e 0

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r

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~

Wel (f

n W.

al D sp u

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8 T f

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60 6 91=8 6

7 UU=U FF F M

TT T ep p sm m aCe e T T ehw 4

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0 0

0 4

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W y.

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M

-M M

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M M

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Figure 2 Typical PWR MOC Xenon Transient Axial Offset vs Time into Transient Time Step Size Effects 0.3 o 1 HourTime Steps i

O 1/2 Hour Time Steps eCT"M F

/

f.

/

f 2

0.0

-0.1 lis a

)

9 -0.2 w4

.s N

-0.3[

g

!\\

-0.4

(

-0.5-t A

f

-0.6 _

~

-0.7 t

0 2

4 6

8 10 12 14 16 18 20 Time (7500 Gwd/mt + hrs) e l

l

M M

M M

W.

W.M M

M M

M M

m

m. m -

m

N

, b Rpm 3 Typicai PWR MOC Xenon Transient Axial Offset vs Time into Transient Spatial Mesh Effects 03 o 12 Node Axia5y

^

0 24 Node Ax!agy f

~

0.2 z g

0.1 0.0 i

-0.1 E

y 9 -0.2 -

J i

E E

}

-03 _

g

-0.4

(

-0.5 A

-0.6

-0.7 '

O 2

4 6

8 10 12 14 16 18 20 j

Twne (7500 Gwdfmt + hrs)

. ~....

M M'

W W

W W

m

. :M M

M mmm m

Figure 4 Typical PWR MOC Xenon Transient

. Xenon Concentration vs Time into Transient i

I Time Step Size Effects O.3 o 1/2 Hour Time Steps 0 1 HourTime Steps e

^" -,,

7 w

i I

l

. r c 0.25 i

S 5

l 8

3 i

t 9

o O

E o

8

~

i 8

e W

e e

0.15 l

O 2

4 6

8

- 10 12 14 16 18 20 i

Time (7500 Gwdrmt + hrs) s 4..

sm..

7,

..-a

+...,e