ML20010B610

Pdq 07 Discrete Model.

ML20010B610
Person / Time
Site:	Surry, North Anna
Issue date:	07/31/1981
From:	Bowling M, Rhodes J, Matthew Smith VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
Shared Package
ML18130A343	List: ML18130A342 ML20010B610 ML20010B617 ML20010B624
References
VEP-FRD-19A, NUDOCS 8108170337
Download: ML20010B610 (98)

Text

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Vepco THE PDQ 07 DISCRETE MODEL T

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i-i ja8oH88Faa88)lg FUEL RESOURCES DEPARTMENT I VIRGINIA ELECTRIC AND POWER COMPANY

VEP-FRD-19A l

TSE PDQ07 DISCRETE MODEL by l

M. L. Smith Nuclear Fuel Engineering Group Fuel Resources Department Virginia Electric and Power Company Richmond, Virginia.

July, 1981 ,

Recommended for Approval:

NW k 0 M. L. Bowling, Superviso'r Nuclear Fuel Engineering Group Approved:

[/J. T. Rhodes, Director Nuclear Fuel Engineering and Operation

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o UNITED STATES

..UCLEAR REGULATORY COMMISSION

$ ,g( '4 j W ASHINGT ON, D. C. 20555

• I,W( l 81981 Mr. W. N. Thomas, Vice' President Fuel Resources Virginia Electric Power Company.

Richmond, Virginia 23261

SUBJECT:

ACCEPTANCE FOR REFERENCING 0F TOPICAL REPORT VEP-FRD-19 l "THE PDQ 07 DISCRETE MODEL" The Nuclear Regulatory Commission (NRC) staff has completed its review of the Virginia Electric and Power Company Topical Report number VEP-FRD-19 entitled -

"The PDQ 07 Discrete Model". The PDQ 07 Discrete Model is two-dimensional (x-y geometry), two neutron energy group diffusion-depletion model which explicitly represents each fuel rod in the reactor. It was developed by

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the Virginia Electric and Power Company (VEPC0) and utilizes the Babcock and Wilcox developed NULIF, HAFIT, PDQ07, PAPDQ, and SHUFFLE ccdes. It is used specifically to perform reactor phytics analysis, fuel management analysis, and to support the reac't or startup and cycle operation of the Vepco Surry and North Anna nuclear reactors. The accuracy of the model predictions is demonstrated through comparison with measurement data obtained from the Surry reactors. Our summary of the evaluation is -

attached.

As the result of our reviews we conclude that the Virginia Electric and Power Company Licensing Topical Report VEP-FRD-19 entitled "The PDQ 07 Discrete Model" dated July 1976 is acceptable for referencing in licensing actions by VEPC0 to the extent specified and under the limitations in the report and the attached evaluation.  ;

We do not intend to repeat the review of the safety feature 3 described in the report and as found acceptable herein. Our acceptance applies only to tne use of features described in the topical report and as discussed herein. .

"In accordance with established requirements, it is requested that Virginia

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Electric and Power Company issue a revised version of this report within three months of the receipt of this letter. This evaluation letter is to be included in the revised version between tha title page and the abstract and the approved report will carry the identifier VEP-FRD-19A.

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f Mr. W. N. Thomas IAA7 1 a Sc -

Should Nuclear Regulatory Commission criteria or regulations change such that our conclusions as to the acceptability of the report ar1 invalidated

%;rginia Electric and Power Company will be expected to revita and resubmit the topical report or submit justification for the continue effective applic-ability of the topical report without revision.

If you have any questions about the review or our conclusion, please contact us.

Sincerely,

\ n_ , -

Robert L. Tedesco, Assistant Director for Licensing Division of Licensing

Enclosure:

As stated i

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YAY 7 1981 EVALUATION OF VEPC0 TOPICAL REPORT VEP-FRD-19 Report Number: VEP-FRD-19 Report

Title:

The PDQ07 Discrete Model Report Date: July 1976 Originating Organization: Virginia Electric and Power Company Reviewed by: Core Performance Branch The Virginia Electric Power Company has notified the NRC, by letter dated October 17, 1980, of their intention to perform their reload licensing in-house after late 1981. In support of this intention they have submitted and will submit a number of topical reports for our review. Report number VEP-FRD-19 is one of those reports. The Reactor Physics Section of the Core Performance Branch has reviewed this report.

Our evaluation follows.

1. Summary of Report The report VEP-FPD-19 is primarily a document which presents data to qualify the code ?DQ07 in the discrete mode for use by VEPC0 per-sonnel to perform reload analyses for VEPCO-cwned and operated reactors. Brief descriptions of the PDQ07 discrete model and its satellite codes are given. Since VEPCO purchased these codes from Babcock and Wilcox (B&W) they reference reports published by B&W for detailt of the various codes.

A detailed description of the Surry Unit I and Unit 2 reactor cores are presented. Included are thermal-hydraulic design parameters, and mechanical design parameters for the fuel assemblies, ccqtrol rod assemblies, burnable poison rods and core structure including core barrel, thermal shield and side, and axial reflectors. Details of the first and second cycle loadings of both Unit 1 and Unit 2 are given.

A description of the manner in which the cores are modeled is given as well as a discussion of the manner in which the various satellite codes are used to provide input to and process the output from

_ PDQ07. Comparisons of calculated values of certain parameters with measured values and with values calculated by vendor (Westinghouse) t methods are given. Comparisons are made for critical boron con-centrations, core radial power distributions, control bank worth, differential boron worth, and, at end of cycle, the burnup distributions.

Calculations and power distribution measurements are compared in two dimensions only since computer size limits the discrete model calcu-lations to two dimensions. At beginning of cycle comparisons are made for both hot zero power and hot full power conditions. At end of life only hot full power conditions are considered. Comparison are also presented batween the measured data and the vendors calcula-tions which have been used to perform reload analyses up to this time.

L The results of the comparisons between calculation and measurement may be summarized as fcilow:

i. Assembly average power distributions are typically predicted to within a two percent standard deviation. The maximum standard deviation obtained for the first and second cycles of Unit 1 and Unit 2 was 4.3 percent at zero power and 3.6 percent a powers greater than 10 percent of full power.

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2. Peak rod F values are predicted typically within 2.5 percent with the ma$imum difference being 4.3 percent. The VEPC0 model tends to underpredict this quantity.

3. Assembly average burnups are predicted to within 2.5 percent and batch average burnup to within 1.5 percent.

4. Typically, critical soluble baron concentration are predicted to within 30 parts per million and boron worth to within 3 percent.

5. Control rod bank integral worths are predicted typically within 6 percent with a maximum deviation of 9.7 percent.

Comparison of the results of the VEPC0 model to those of the vendor for the same quantities shows that the average absolute deviations for the two models are similar with the VEPC0 model tending to have smaller average deviations.

2. Sumary cf Evaluation We have reviewed the information presented in licensing topical report, VEP-FRD-19, "The P0Q07 Discrete Model ." The following coments summarize our evaluation.

The various codes that comprise the model are described by reference to documents that have been submitted by their developer, the B&W Company. These reports have been previously reviewed and accepted by the staff for reference by B&W. Their use as references for code descriptions by VEPC0 is acceptable inasmuch as VEPC0 purchased the codes from B&W. The procedures used by VEPCO to implement the codes

are described in the report. These are standard procedures in industry-wide use and are acceptable.

We have reviewed the (sta presented to support the conclusions regarding the uncertainties in the calculated results. We conclude the sufficient examples of comparisons between calculation and measurement to pemit the evaluation of calculational uncertainties.

We concur with the particular values of uncertainties given in the topical report and repeated in Section 1 above. We further concur with the conclusion that the VEPC0 modal is an acceptable replace-ment for the vendor models currently in use. These conclusions apply to the calculation of

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- assembly average radial pawer distributions;

- F ' values; 3

- assembly average burnups;

- critical baron concentrations and boron worths; and

- control rod integral bank worths.

3. Evaluation Procedure The review of topical report VEP-FRD-19 has been conducted within the guidelines arovided for analytical methods in the Standard Review Plan, k ction 4.3. Sufficient infomation is provided to pemit a knowledgeable person to conclude that the VEPC0 model described in this report is state-of-the-art and is acceptable.

Sufficient data are presented to permit the conclusion that the derived uncertainties are reasonable and are acceptable.

4. Regulatory Position Based on our review of licensing topical report VEP-FRD-19 we con-clude that it is acceptable for reference in licensing actMns by VEPCO. Such reference may be made for purposes of describir:g the code and for citing uncertainties in the following quantities:

- assemMy average radial power distributions;

-F values; _

- assembly average burnups;

- critical baron concentrations and baron worths; and

- control rod integral bank worths.

We further conclude that this model is an acceptable substitute for vendor calculations of the above named quantities.

We endorse VEPCO's commitment in the report to continue verification and improvements to the PD007 discrete model as more data are obtained from the Surry and North Anna reactors.

_ _ .__ _ _ _-_ _________m_____.m_.-_____ . _ .. - --- ---

CLASSIFICATION / DISCLAIMER The data and analytical techniques described in this report have been prepared for specific application by the Virginia Electric

! and Power Company. The Virginia Electric and Power Company makes no claim as to the accuracy of the data or technique contained in this report if used by other organizations. In addition, any use of this report or any part thereof must have the prior written approval of the Virginia Electric and Power Company.

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ABSTRACT A two-dimensional (x-y geometry), two neutron energy group diffusion-depletion model which explicitly represents each fuel rod in the reactor has been developed by the Virginia Electric and Power Company (Vepco). The model, which is designated as the PDQ07 discrete model, utilizes the Babcock aAd Wilcox developed NULIF, HAFIT, PDQ07, E4PDQ, and SHUFFLE codes. It is used specifically to perform reactor physics analysis, fuel management analysis, and to support the reactor startup and cycle operation of the Vepco Surry and North Anna nuclear reactora. The accuracy of the model predictions is demonstrated through comparison with measurement data obtained from the Surry reactors.

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ACQOWLEDGEMENTS The author would like to thank Mr. S. A. Ahmed for his o

assistance in data analysis and preparation of this report and Ms. Carolyn Watson for her typing of the final manuscript. The r

author would also like to express h'.s appreciation for the contribu-tion of a number of people who were responsible for reviewing this report.

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l TABLE OF CONTENTS ,

1 TITLE PAGE CLASSIFICATION. . . . . . . . . . . . . . . . . . . . . . . . . . . .i ABSTRACT. . .... ................... . . . . . .11 ACKNOWLEDGEMENTS . . ...... . . . . . . . . . . . . . . . . . . .111 TABLE OF CONTENTS ... . ... . .. . . . . . . . . . . . . . . . .iv LIST GF TABLES . . .. . . . . . . . . . . . . . . . . . . . . . . . .v LIST OF FIGURES . . ...... . . . . . . . . . . . . . . . . . . .vil SECTION 1 - INTRODUCTION . . .. . . . . . . . . . . . . . . . . . . .1 -1 SECTION 2 - CORE DESCRlPTION . . . . . . . . . . . . . . . . . . . . .2-1

2.1 INTRODUCTION

, . . . . . . . . . . . . . . . . . . .2-1 2.2 CORE DESIGN . . . . . . . . . . . . . . . . . . . . .2-1

i 2.3 FUEL LOADINGS . . . . . . . . . . . . . . . . . . . .2-3 SECTI.ON 3 - MODEL DESCRIPTION , . . . . . . . . . . . . .. . . . . .3-1 i

3.1 INTRODUCTION

. . . . . . . . . . . . . . . . . . . .3-1 6

3.2 NULIF AND HAFIT . . . . . . . . . . . . . . . . . . .3-4 3.3 PDQW'. . .. . . . . . . . . . . . . . . . . . . . 3-15 i SECTION 4 - COMPARISON OF MODEL PREDICTIONS WITH MEASUREMENT DATA . . . . , . . . . . . . . . . . 4-1

4.1 INTRODUCTION

. . . . . . . . . . . . . . . . . . . .4-1 4.2 ANALYTICAL CALCULATIONS . . . . . . . . . . . . . . 4-1 4.3 MEASUREMENT DATA . . . . . . . . . . . . . . . . . .4-4 4.4 RESTATS . ... . . . . . . . . . . . . . . . . . . .4-7 i

SECTION 5 -

SUMMARY

. . . . . . . . . . . . . . . . . . . . . . . 5-1 SECTION 6 - REFERENCES . . . . ... . . . . . . . . . . . . . . . . .6-1 i

APPENDIX A - Description of the TOTE, FOLLOW, and INCORE computer codes . . . . . . . . . . . . . . . . A-1 APPENDIX B - Representative INCORE Output . . . . . . . . . . . . . .B-1 iv

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! LIST OF TABLES Table Title-2-1 Surry Core Description . . . . . . . . . . . . . . . . . . 2-4 3-1 Contents of Fine-En rgy Group Cross Section Library . . . . 3-5 3-2 Fine Energy Group Cross Sections Liorary Constituents . . . 3-7 3-3 Depletion Equations Used in PDQ07 . . . . . . . . . . . . . 3-18 4-1 Sumary of Comparisons For Both the Initial and Reload Cycles ... ... . . , . . . . . . . .... .. . . . . . 4-2 4-2 Type of Comparisons . .. . . . . . . . . . . . . . . . . . 4-8 4-3 Comparison of Predicted and Measured Assembly Average Power Distributions For Surry Unit 1, Cycle 1 . . . . . . . 4-9 4-4 Comparison of Predicted and Measured Assembly Average l' Power Distributions For Surry Unit 1, Cycle 2 . . . . . . . 4-10 4-5 Comparison of Predicted and Measured Assembly Average Power Distributions For Surry Unit 2, Cycle 1 . . . . . . . 4-11 4-6 Comparison of Predicted and Measured Assembly Average Power Distributions For Surry Unit 2, Cycle 2 . . . . ... . 4-12 4-7 ComparisonofPredictedandMeasuredFfH For Surry Unit 1, Cycle 1 . . . . . . . . . . .. . . . . . . . . . . 4-13

4-8 Comparison of Predicted and Measured F$H For Surry 1, Cycle 2 . . . . . . . . . . . . . . . . . . . . . . . . . .

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4-9 Comparison of Predicted and Measured F{H For Surry Unit 2, Cycle 1 . . . . . . . . . . . . ..... . . . . . 4 4-10 Comparison of Predicted and Measured F[H For Surry Cait 2, Cycle 2 . . . . . . . . . . . ..... . . . . . . 4-16 4-11 Assemblywise Accumulated Burnup and Estch durnup Sharing (103 NWD/MTU) For the Cycle 1 Operation of Surry Unit 1 ........ . . . . ... ...... . . 4-17 4-12 Assemblywise Accumulated Burnup and Batch Burnup Sharing (103 MRD/MTU) For the Cycle 2 Operation of Surry Unit 1 ............ .. .... . . . . ; . 4-18 4-13 Assemblywi3g Accumula:ad Burnup and Batch Burnup Sharing (10 MRD/MTU) For the Cycle 1 Operation of Surry Unit 2 ................. .. ... . . 4-19

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9 LIST OF TABLES (Continued)

Tabis Title 4-14 Assemblywise Accumulated Burnup and Batch Burnup Sharing (103MRD/MTU) For the Cycle 2 Operation of Surry Unit 2 . . . . . . . . . . . . . . . . . . . . . . . 4-20 4-15 Comparison of Predicted and Measured Critical Baron Concentration For Various Control Rod Configurations For Cycle 1 of Surry Unita 1 and 2 . . . . . . . . . . . . . 4-21 4-16 Comparison of Predicted and Measured Critical Boron Concentration For Various Control Rod Configurations For Cycle 2 of Surry Units 1 and 2 . . . . . . . . . . . . . 4-22 4-17 Comparison of Predicted and Measured Differential Baron Worth For Cycle 1 of Surry Units 1 and 2 . . . . . . . 4-23

( 4-18 Comparison of Predicted and Measured Differential Boron Worth For Cycle 2 of Surry Units 1 and 2 . . . . . . . 4-24 4-19 Comparison of Predicted and Measured Integral Bank Worth For Cycle 1 of Surry Units 1 and 2 . . . . . . . . . . 4-25 4-20 Comparison of Predicted and Measured Integral Bank Worth For Cycle 2 of Surry Units 1 and 2 . . . . . . . . . . 4-26 i

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LIST OF FIGURES

. Figure Title Page No.

2-1 Cross Sectional View of Surry Fuel Assemblies . . . . . . . 2-6 2-2 Control Rod Bank Locations. .. .. . . . . .. . . . . . . 2-7 2-3 Surry Units 1 and 2 -- Cycle 1 Fuel Loading . .. ... . . 2-8 2-4 Surry Unit 1 -- Cycle 2 Fuel Loading. .. . ... ... . . 2-9 2-5 Surry Unit 2 -- Cycle 2 Fuel Loading. . . . . . . . . . . . 1-10 2-6 Surry Units 1 and 2 -- Cycle 1 Burnable Poison Bpd Loading . . . . . .. ... . . .. . . . . . . . . .. . . 2-11 2-7 Surry Unit 1 -- Cycle 2 Burnable Poison Rod Loading . . .. . . .. ... ... . .. . . . . . . . . . 2-12 2-8 Surry Unit 2 -- Cycle 2 Bernable Poison Rod Loading . . . . . . . .... . .. . . . . . .. . .. . . 2-13 3-1 Flowchart For the PDQO7 Discrete Model . . .. .. . . . . 3-3 4.1 Surry Unit 1 - Cycle 1 Critical Boron Concentration vs Burnup . . . . ... .. . ... .. .. . ... . . . . 4-27 4.2 Surry Unit 1 - Cycle 2 Critical Boron Concentration Vs 3urnup . . . . . .. .. . . .. . .. ... . . . . . . 4-28 4.3 Surry Unit 2 - Cycle 1 Critical Boron Concentration vs Burnup . . . . . .. ........ . . .. .. .. . . 4-29

j. 4.4 Serry Unit 2 - Cycle 2 Critical Boron Concentration l

va asrn 9 . . . . . . .... . .. . . . . ... . . . . . 4-30 3-1 Incore Calcualted Assemblywise Average Power Distribution For Initial Core At Beginning of Life Condition . . . . . . . B-1 B-2 Incore Calculated Assemblywise Average Power Distribution i

For Initial Core At Beginning of Life Condition . . . . . . . B-2 B-3 Incore Calculated Assemblywise Average Power Distribution For Initial Core At Beginning of Life Condition . . . . . . . B-3 i

B-4 Incore Calculated Assemblywise Average Power Distribution ~

For Initial Core At Middle of Life Condition . . . . . . . . B-4 B-5 Incore Calculated Assemblyvise Average Power Distribution For Initial Core At End of Life Condition . . . . . . . .. . B-5 B-6 Incore C'lculated Assemblywise Average Power Distribution For Reload Core at Beginning of Life . . . . . . . . . . . . B-6 4

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1 LIST OF FIGURES (Continued)  !

Figure Title t B-7 Incore Calculated Assemblywise Average Power Distribution For Reload Core At Beginning of Life . . . . . . . . .. . . .B-7 B-8 Incore Calculated Assemblywise Average Power Distributton For Reload Core At Middle of Life. . . . . . . . . . . . . . .B-8 B-9 Incore Calculated Assemblywise Average Power Distribution For Reload Core At End of Life . . . . . . . . . . . . . . . .B-9 i

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SECTION 1 - INTRODUCTION The Virginia Electric and Power Company (Vepco) is currently developing the capability to perform nuclear reactor analyses for the Surry and North Anna nuclear power statious The objective of this topical report is (1) to dercribe one of the computational models de-veloped at Vepco for the purposes of reactor physics analyses, fuel management evaluation, and core follow support and (2) to demonstrate 6

the accuracy of this model by comparing analytical results generated

, with the model with actual measurements from Surry Units No.1 and 2.

The computational model to be described is a discrete (one mesh line per rod), two-dimensional, two neutron energy group, diffu-sion-depletion (with thermal-hydraulic feedback) calculational package and is designated as the PDQ07 discrete calculational model. The PDQ07 discrete model uses the NULIF1, PDQ072 , SHUFFLE , HAFITO , and S

PAPDQ computer codes which are part of the Fuel Utilization and Per-forrance Analysis Code 0 (FUPAC) system obtained from the Babcock and Wilcox Company. The FUPAC system is currently used by Babcock and Wilcox to perform production reactor analysis and design. A detailed description of the input / output, functioning, and physical model of the

'above computer codes can be obtained from the referenced Babcock and Wilcox computer code manuals. The FUPAC system is maintained by Vepco and updated through contractual arrangements between Vepco and Babcock and Wilcox 0 The types of calculations that can be performed with the PDQ07 di-crete model include:

1. REACTOR PHYSICS .',NALYSIS:

a. Two-dimensional radial power distributions, including 1-1

.-. . . .- - _ .. . _ . = __ _ . ._ - _--

l relative radial peakin factors (Fxy) and enthalpy rise hot channel factors (F H), for both rodjed and unrodded '

planes as a function o burnup.

b. Critical soluble boron concentrations as a functioa of burnup.

a L c. Integral control rod bank worths.

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d. Nuclide concentrations for the fuel and burnable poisoe rods as a function of burnep.

e. Fuel rod and assembly average burnup distributions.

2. FUEL MANAGEMENT:

a. Batch power and burnup sharing,

b. Fuel isotopics as a function of burnup.

, c. Evaluction of alternative fuel loading patterns.

3. CORE FOLI.OW:

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a. Input constants for determining measured core power distributions through use of the INCORE Code.

b. Critical boron concentrations and control rod bank iorths.

To date the PDQ07 discrete model has been used at Vepco (on tho Vepco IBM System 370 computers) to calculate best estimates of the power and burnup sharing for future cycles in order to determine and optimize nuclear fuel requirements. In addition, substantial effort has gone into the evaluation of alternative fuel loading patterns for specific operating cycles in order t) optimize fuel utilization and unit operating flexibility.

The benefits obtained from this limited aiplication'have been significant.

As a result, it is now intended to use the PDQ07 discrete model to perform the complete reactor physics analysis associated with specific operating cycles and to verify (through core follow analysis) that the reactor is operating in accordance with these design predictions. It is anticipated that the extended use of this model will previde additional benefits to-1-2

l Vepco through increased design flexibility, technical support of licensing positions, and greater operational freedom.

The remainder of this report describes the Surry Units No. 1 and 2 reactor core to be modeled, the purpose and interrelationships of

, the various computer codes which comprise the PDQ07 discrete model, the processes through which these codes function, the specific modeling of tnese codes to represent the reactor core, and the comparf <>on of calculated results with reactor measurements from Cycles 1 and 2 of Surry Units No. I and 2.

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SECTION 2 - CORE DESCRIPTION

2.1 INTRODUCTION

The Surry Nuclear Power Station, which c.trrently consists of two operating units, has been selected as the operating system to be modeled for verification of the PDQ07 discreta model. The Surry Units No. I and 2 are identical Westinghouse designed three coolant loop pressurized water reactors with thermal ratings of 2441 Mwt. Initial criticality was achieved for Surry Unit No. 1 on July 1, 1972 and for Surry Unit No. 2 on March 7, 1973. The initial cycle for Surry Unit No. 1 was completed on October 24, 1974 and ror Surry Unit No. 2 on April 26, 1975. Second cycle operation commenced on January 30, 1975 and June 14, 1975 and was completed on Septem-bar 26, 1975 and April 22, 1976 for Surry Units No. 1 and 2, respectively.

2.2 CORE DESIGN The Surry cores consist of 157 fuel assemblies surrounded by a core baffle, barrel, and thermal shield and enclosed in a steel pressure vessel. The pressure inside the vessel is maintained at a nominal 1230 psia. The coolant (and moderator) is pressurized water which enters the 0

bottom of the core at 532 F and undergos an average rise in temperature of 65.5 F befora exiting the core. The average coolant temperature is 5660 F and the average linear power density of the core is 6.2 kw/ft.

Each of the 157 fuel assemblies consists of 204 fuel rods arranged in a 15 by 2* square array. The fuel used in the Surry cores consist of slightly enriched uranium dioxide fuel pellets centained within a Zircaloy-4 clad. A small gap containing pressurized helium exists between the pellets y and the inner diameter of the clad. For the positions in the 15 x 15 array not occupied by fuel rods, these are 20 guide' tube locations for either 2-1

solid burnable poison rods or control rods and one centrally located instrumentation tube. (See Figure 2-1.) The fuel rods in each fuel assembly are supported by seven Inconel-718 grids located along the length of the assembly. These grids are mechanically attached to the guide tubes, which are, in turn, welded to the upper and lower nozzles, and thus provide for assembly structural support.

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There are 48 full-length Rod Cluster Control Assemblies (re-ferred to as control rods) used to control core reactivity as well as five part-length rods for axial power shaping. (It should be noted that the part-length control rods are physically present but are not currently allowed to be insertend into the core). The absorber material of the control rods is an alloy consisting of 80% silver, 15% indium, and 5% cadmium. The f

various control rods are arranged in and moved in symmetrically located groups, or banks, as depicted in Figure 2-2. Banks D, C, B, and A are de-noted as the control banks and are moved in a fixed sequential pattern to

. control the reactor over the twer range of operation. The remaining rods, Banks SA and SB, are denoted as shutdown banks and are used to provide shut-down margin.

In addition to the control rods, a chemical (boric acid) shim is used to control excess core reactivity and to facilitate operational flexi-bility. Above certain concentrations of ebemical shim, burnable poison rods are also used to control excess reactivity. Trash and/or depleted burnable poison rods can also be used to shape (i.e., improve) the core power distribution. The burnable poison rods contain borosilicate in the form of Pyrex glass clad in a stainless steel tube. Burnable poison rods, which may be used in any fuel assembly not under a control rod bank loca-tion, cons'ist of clusters of either 8, 12, 16, or 20 rods which are inserted into the Ziresloy-4 control rod guide tubes.

2-2

Specific values of the principal mechanical and thermal-hydraulic

, parameters of the Surry core are provided in Table 2-1. A complete descrip-tion of the Surry units is given in Reference 7.

2.3 FUEL LOADING i

The initial and reload quarter core fuel loadings (i.e., initial c

enrichments and density, ;cevious cycle location if appropriate, beginning

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of cycle burnup, and number of fresh or depleted burnable poison rods pre- l l

sent) for both Surry units are provided in Figures 2-3 through 2-8. It should be noted that the fuel loadings for Cycle 1 of both Surry units are identical. The fuel management strategy employed in the initial cycle of operation of each unit was the checkerboard loading of the two lower enriched fuel batches in the center of the core and the highest enriched fuel batch I

around the periphery of the core. In the second cycle of operation, this strategy was adhered to by loading fresh fuel around the periphery of the Core.

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Table 2-1

. SURRY CORE DESCRIPTION

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, THERMAL AJD !!TDRAULIC DESIGN PARAMETERS

'l' Total core heat output, Hwt 2441 Heat generated in fuel, % 97.4

!! System operating pressure, psi 2250 Total coolant flow rate, Ib/hr (gpm) 100.7 x 106 (265,500)

Coolant Temperatures, OF (0100% powar)

Nomicsi inlet 532 Average rise in the core 65.5 Average in the core 566

, Nominal outlet of hot channel 642 Aversge linear power density, Kw/ft 6.2 MECHANI'*AL DESIGN PARAMETERS i Fuel Assemblies Design Canless 15 x 15 Number 157

!{ Rod pitch, inches 0.563 Overall dimensions, inches 8.426 x 8.426

, Number of grids per assembly (material) 7 (Inconal-718)

Number of instrumentation tubes 1

l Fuel Rods I Number 32,028

t Number of rods / assembly 204 Batch 1, 2, 4 Batch 3 Outside diameter, inches 0.422 0.422 Diametrical gap, inches 0.0075 0.0085 Clad thickness, inches 0.0243 0.0243

! Clad material Zircaloy-4 Fuel Pellets Material Sintered UO2 Density (% of theoretical) See Figures 3-4 through 3-Batch 1, 2, 4 Batch 3 Outer diameter 0.3659 0.3649 Control Rod Asserblies Neutron absorber 5% Cd-15% In-80% Ag i Cladding Material Type 304 SS-Cold worked Clad thickness, inches 0.019 l~~ 48 Number (full length) '

Number of rods per assembly 20

~._ _ _ _ _ _ - - __-- 2 _4 __ . . , -

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l Table 2-1 (Continued)

Eurnable Poison Rods Material Pyrex glass content B 02 3 (w/o) 12.5

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Core Structure Core barrel I.D./0.D., inches 133.875/137.875 Thermal shield I.D./0.D., inches 142.625/148.000 Core diameter, inches (approximate) 119.5 Reflector thickness (approximate) and composition

! Top - Water plus steel, in. 10 l Bottom - Water plus steel, in. 10 Side - Water plus steel, in. 15 t

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FIGURE 2-1 CROSS SECTIONAL VIEW OF SURRY FUEL ASSEMBLIES

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2-6

FIGURE 2-2 CONTROL ROD BANK LOCATIONS R P N M L K J *H G F E D C B A 1 ,

?

I 2 A D A I

3 S S t 4 C B P B .C .

5 S S 6 A B D C D- B A e

7 . s s s 3

- 8 D P C P C. P D l

9 'S t

S S S I'

10 A B D C D B A 11 s 3

{ -.-

, 12 C B P B C j --

13 S S 11 A D A i 15 i  ;

1 CONTROL ROD ASSEMBLY BANKS

• j Function Number of Assemblies .

Control Bank D 8 Control Bank C 8 Control Bank B 8 Control Bank A 8 -

Shutdown (S) 16 l Part Length (P) 5 i

3

SOURCE ASSDiBLY LOCATIONS 2-7

e Figure 2-3

• SURRY Ul;ITS 1 AND 2 - CYCLE 1 FUEL LOADING 08 09 10 11 12 13 14 15

,. 1 2 1 2 1 2 1 3 l H 0 0 0 , 0 0 0 0 0 Fresh Fresh Fresh Fresh Fresh Fresh Fresh ~ Fresh y 2 1 2 1 2 1 3 3 0 0 0 0 0 0 0 0 Fresh Fresh Fresh Fresh Fresh Fresh Fresh _ Fresh _ ,_

K 0 0 0 0 0 0 0 Fresh Fresh Fresh Fresh Fresh '

Fresh __ Fresh j g 2 1 2 1 2 3 3 0 0 0 0 0 0 0 .

Fresh Fresh Fresh Fresh' Fresh ,

Fresh Fresh g i 2 1 2 1 3 0 0 0 0 0 0 Fresh Fresh Fresh Fresh Fres'h Fresh 2 1 2 3 3 N Batch Enrichment Density 0 0 0 0 0 No. w/o  %

Fresh Fresh Fresh Fresh Fresh 1 1.85 94 1 3 3 p 3 2 2.55 93 0 0 0 0 3 3.10 92 j Fresh Fresh Fresh Fresh ,

3 R ,

i "O O Fresh Fresh ,

LEGEND xx - Batch No.

77 -

i~-luitialBurnup(ISD/MTU)

. zz l--Previous Location (If applicable) 2-8

Figure 2-4 SURRY UNIT 1 - CYCLE 2

+ FUEL LOADING 1

08 09 10 11, 12 13 14 15

! 1 4B 1 4B 2 4B 1 4C 15360 0 12550 t 0 15300 0 14680 0 i H08 Fresh H14 Fresh H13 Fresh H12 Fresh 4B 2 4A 2 2 2 4C 4C J 0 16430 0 14300. 16070 14380 , 0 0

, Fresh LO8 Fresh K13 K11 L12 Fresh Fresh 1 4A 1 4A 2 l 1 4C K 12540 0 11280 0 15920 14190 0 P08 Fresh M12 Fresh J12 K12 Fresh 4B 2 4A 2 2 4C 4 4C L 0 14300 0 16760 16600 O O Fresh N10 Fresh J08,; J10,j Fresh Fresh 2 2 2 2 I4A 4C M 15300 16070 15920 16600 0 0 N08 L10 M09 K09 Fresh Fresh 4B 2 1 4C 4C N O 14380 14190 0' O Batch Enrichment Density Fresh M11 M10 Fresh Fresh --

1 1.85 94 1 4C 4C 4C 2 2.55 93 P 14680 0 0 0 4A 1.85 95 9

M08 Fresh Fresh Fresh .

C 3 4C 4C ,

R 0 0 Fresh Fresh .

LEGEND .

---Batch No.

xx .

~

yy - Initial Burnup (MWD /MTU)

. zz ---Previous Location (If applicable) 08 2-9

l i

. I

~

Figure 2-5

,i SURRY UNIT 2 - CYCLE 2 FUEL LOADLNG

/ 08 09 10 11 12 13 14 15 1 3 3 2 4A 2 2 4B

[ H l -

16690 15420 11410 , 17990 0 18300 16810 0

, H08 G14 H15 H11 Fresh H09 H13 Fresh y 3 4A 2 4A 2 3 43 4B 15420 0 17630 0 . 15750 14730 0 0

.i P09 Fresh K11 Fresh K13 N11 Fresh Fresh 3 2 3 2 4A 2 4B j

K 11410 17630 15420 18150 0 17470 0 R08 L10 J14 l J10 Fresh J12 Fresh 2 l 4A 2 4A 2 4B 43 17990 0 18150 0 15860 0 0 Li LO8 Fresh K09 Fresh L12 Fresh Fresh w .

. g 4A 2 4A 2 4A 45 0 15750 0 15860 0 0 Fresh N10 Fresh M11 Fresh Fresh 2 3 2 4B 4B Batch En'richment Densit't 18300 14730 17470 0 0-No. w/o

• 4 J08 L13 M09 Fresh Fresh t -

1 1.85 94 2 4B 4B 4B 2 2.55 93 l 9 16810 0 0 0 6 9 N08 Fresh Fresh Fresh 4B 3.10 95

' 4 4B R ,

0 0 l Fresh Fresh -

l LEGEND f

Batch No. ,

[

yy --Initial Burnup (MWD /MTU) i z- - Previous Location (If applicable)

I e

l 2-10 l .. - -_ . _ . - .. . - - . . - .-. .

i i

l Figure 2-6 SURRY UNITS 1 AND 2 - CYCLE 1 I Burnable Poison Rod Loading 08 09 10 11 12 13 14 15 i

1 2 1 2 1 2 1 3 i

12 i 12 12

.i -

2 1 2 1 2 1 3 3 12 12 12 12

'l 0 0

~

1 2 1 2 1 2 3 i 12 12 12 ,

1 2 1 2 1 2 3 3 L 12 12 12 12 0

1 2 1 2 1 3 12 12

...J ,

2 1 2 3 3

~

N 12 12 12 1

l 1 3 3 3 l-, P 12 3 3 R r i LEGIND xx Batch No. .

yy No. of Fresh Biarnable Poison Rods E'

2-11

Figur: 2-7 SURRY UNIT 1 - CYCLE 2 BURNABLE POISON ROD LOADING 08 09 10 11 12 13 14 15 1 4B 1 4B 2 4B 1 4C

" 8  ; , 8 12

-l _ .

4B 2 4A 2 2 2 4C 4C 8 20 l

_~ ,..

22 1 4A 1 4A 2 1 4C K

._ - -- l 4B 2 4A 2 2 4C 4C 8 12 2 2 2 2 4A 4C M

I w

4B 2 1 4C 4C N

. 12 12 1 4C 4C 4C 20 ,

4C 4C 1 R

- Batch No.

xx .

j' ,

yy - No. of Fresh Burnable Poison Rods

. zz --No. of Depleted Burnable Poison Rods 6

5 2-12 e

Figure 2-8 SURRY UNIT 2 - CYCLE 2 BURNABLE POISON ROD LOADING 08 09 10 11 12 13 14 15

, i 1 3 3 2 4A 2 2 4B

- H ,

f 12 3 4A 2 4A 2 3 4B 4B J

.I 3 2 3 2 4A 2 4B K

2 4A 2 4A 2 4B 4B L

12 4A 2 4A 2 4A 4B M

2 3 2 4B 4B l .N 12 ~-

12 2 4B 4B 4B i

I 4B 4B R

l LEGEND xx l - Batch No. ,

, zz -No. of Depleted Burnable Poison Rods l*

2-13

_ __ o -

SECTION 3 - MODEL DESCRIPTION

3.1 INTRODUCTION

The PDQ07 discrete model performs a two-dimensional (x-y) diffu-sion-depletion calculation in which each fuel rod, burnable poison rod, control rod, control rod guide tube and the instrument tube within each fuel assembly as well as the core baffle and radial reflector in one quarter of -

the reactor core are represented explicitly, and in which thermal-hydraulic feedback effects are considered. Since a detailed simultaneous calculation of the neutron flux as a function of both energy and position is impractical for large power reactor cores (due to the computational t!.me that would be required), the neutron flux for each material composition in the reactor is first calculated as a function of neutron energy, and then this flux spectrum is used to prepare diffusion theory cross sections by collapsing the spectrum-weighted cross sections into two neutron energy groups (denoted as the fast and thermal energy groups or simply two groupe). The matarial compositions repre-sented in these calculations consist primarily of either a fuel rod, a control

- rod guide or instrumentation tube, a control rod, or a burnable poison rod, and the moderator-coolant associated with the rod. These material compositions are denotea as unit cells, since a lattice of appropriately chosen unit cells can be uced to represent the geometry and material composition of the reactor core.

The heterogeneous effects of the unit cell on the neutron flux are represented through the use of Dresner's method in the fast energy range and the ABH (Amcuyal, Benoist, and Horow:Lcz) approximation to transport theory in the ther-mal range. Fine energy multi-group cross sections (31 epithermal and 80 ther-mal) are required for the calculation of neutron flux in the unit cell as a ft.netion of energy prior to collapsi~g into two-group cross sections.

The spectrum-weighted two-group cross sections associated with each unit cell as well as the core baffle ana reflector are then used in an .

3-1

_ _ _ - _ .-- - m _- . _ - - . _ . - _ __ ___

iterative diffusion .heory calculation of the neutron flux as a function of radial position. From the neutroa flux and cross sections the core l

power (istribution is determined, and subsequently, the fuel and moderator temperatura distributions are calculated. Thermal feedback effects are included in the diffusion theory calculation by recalculating the neutron cross sections, power distribution, and fuel and moderator tems rature distribution iteratively until both the required nuclear and thermal con-vergences are achieved.

The neutron flux in a reactor is not only a function of energy and position but is elso a function of changes in the nuclide concentrations which vary with burnup. A nuclide depletion calculation is performed based on the two-group fluxes and microscopic absorption and fission cross sections representing each fuel rod in the reactor core. The neutron flux is then recalculated in the diffusion theory calculation for the new neclide concen-trations.

Several interrelated computer codes are used to perform the calcu-lations outlined above. The computer codes comprising the PDQ07 discrete model and their interralatonships are presented in the flow chart of Figure 3-1. The ?DQ07 computer code itself is the principal reactor analysis l

l calculational tool in the PDQ07 discrete model and is used to perform the two-group, two-dimensional diffusion theory calculations. The other codes provide either input data, data manipulation, or use the PDQ07 code output.

As indicated in Figure 3-1, the NULIF computer code is used to calculate the required two-group spectrum-weighted cross sections. The HAFIT computer code formats these cross sections for use in the PDQ07 code (as HARMONY table sets). The PDQ07 code calculates relative radial power distributions based on the neutron fluxes at each corner of each fuel rode These values 3-2

)

Figure 3-1 FL:N' CHART FOR THE PDQ07 DISCRETE HODEL FINE ENERCY DESCRIPTION DESCRIPTION NUCLIDE CROUP NEUTPDN OF UNIT CELL OF REACTOR DEPLETION CUEE CROSS SECTION (ITEL, CONTROL, CEOMETRY DESCRIPTION LIBRARY BURNABLE POISON) C TION 1 I I l 5

a pgp MATERIAL C0KCENTRA110N3 AT BEGINNING OF CYCLE

( PNM7 / \

W \ /

Tk'O ENERGY CROUP l CROSS SECTIONS SHUFFLE AS A FL7CTION OF BURNUP, XENON, BORON, IUEL TEMPERATURE, /

AND HODERATOR TEMPERATURE '

d' h ASSDSLY AVERACE CRITI".J. BORON MATERIAL RE!JLTIVE POWERS POWER & FLUXES CONCENTRATIONS CONCE1TRATIONS FOR FUEL AT INSTRUMENT AND CONTROL ROD AT END OF TUBE IDCATIONS WORT 1iS CYCLE

(

RAFIT

\

l pg s PEAK FUEL ROD l

< POWER FOR EACH I ASSDtBLY

% /

CROSS-SECTION TABLES IN THE FOR:n? REQUIRED BY PTOO7

are averaged by the PAPDQ code to determine the relative power density in each fusi rod. PAPDQ is then used to survey the relative power density in each fuel rod within the fuel assembly to determine the peak fuel rod relative power density. The SHUFFLE computer code is a data manipu-lation code that takes appropriate end of cycle avclide concentrations from the PDQ07 computer code and shuffles this data in the reactor core according to a specified scheme ahich dup)icates calculationally the actual replace-met and movement of fuel assemblies in the reactor core as the result of a refueling.

The details of the calculations performed by each of the computer codes in the FDQ07 discrete model are described in the remairder of this section.

3.2 NULIF AND HAFIT 3.2.1 FINE ENERGY GROUP CROSS SECTION DATA:

The source of basic nuclear cross section data for the NUL.IF computer code calculations is the standard fine-group cross section li-brary used at Babcock and Wilcox (see Raference 1). This cross section library was supplied by Babcock and Wilcox as part of the FUPAC system.

The library contains cross sections for 31 fast and 80 thermal energy groups with a thermal energy cutoff of 1.85 eV. The fast library contains smooth cross sections, resonance parameters, and an (n. 2n) inelastic scattering matrix for each nuclide. The thermal library con-tains temparature-dependent cross sections for each thermal energy group and temperature-dependent thermal scattering kernels (both isotropic and anisotropic kernels for the bound atom model). The contents of the files in the cross section library are listed in Tablu 3-1.

The standard fine-group cross section library contains cross sec-tion data for all structural materials, fissionable isotot ts, fission pro-3-4

l Table 3-1

, CONTENTS OF FINE-ENERGY GROUP CROSS SECTION LIBRARY

, FILE 1* GENERAL LIBRARY DATA TAPE LABEL

+ MATERIAL CONTENTS l EPITHERMAL GROUP STRUCTURE THERMAL GROUP STRUCTURE DELAYED NEUTRON' PRECURSOR DATA FISSION SOURCE DISIRIBUTION DATA GENERAL MATERIAL PARAMETERS TEMPERATURE LIST FISSION PRODUCT YIELDS i

RESONANCE ISOTOPE DATA FISSION SPECTRUM DATA DELAYED NEUTRON DATA FILE 2 FAST CROSS SECTION DATA GROUP DJ.TA ,

GENERAL MATERIAL PARAMETERS GENERAL UNRESOLVED RESONANCE DATA UNRESOLVED RESONANCE PARAMETERS RESOLVED RESONANCE FAldNTERS l SMOOTH CROSS SECTION DATA FILE 3 THERMAL CROSS SECTION DATA GENERAL MATERIAL PAR.AFETEPdi SLOWING-DOWN SOURCE DATA

! SMOOTH CROSS SECTION DATA

! ISOTROPIC SCATTERING KERNEL ANISOTROPIC SCATTERING REILNEL 4

3-5

ducts, and moderator-coolant (water) used in the reactor core. The con-stituents of the library are listed in Table 3-2.

The NULIF code is used to calculate composition-dependent energy spectra and then collapse the fine-energy group cross sections to produce two-group cross sections for each unit cell.

3.2.2 NULIF

The NULIF computer code is used to calculate two-group spectrum- . i weighted neutron cross sections for each unit cell type in the reactor core.

A unit cell consists of either a fuel rcd, a control rod guide tube, a control rod, or a burnable poison (BP) rod, and the mode ator-coolant (water) asso-ciated with the rod.

The neutron energy spectrum for each unit cell is computed using a P1 multi-group approximation to the neutron transport equation. The slowing down treatment for hydrogen is exact, the Fermi age model is used for heavy elements, the Grueling-Guertzel model is used for light elements, and (n, n),

(n, 2n), and (n, 3n) reaction effects are included. Resonance absorption is computti by Dresner's method using Sauer's approximation to compute the Laa-

i coff correction for close-packed pin lattices; spectrum reduction corrections are applied for groups containing multiple resonances. The ABH approximation s

[ to transport theory is used to calculate the thermal (below 1.85 eV) neu:ron flux spa'ial t distribution. In the upscattering or thermal energy range, toend 1

f atom scattering kernels are used for the principal scatterers and Doppler-broadened group cross sections are used for resonance absorbers. Two-group cross sections for each isotope of each unit cell are obtained by spectrum-weighted integrals. (See Reference 1 for a more detailed discussion of the above assumptions.)

The 1.85 eV energy cutoff is used in NULIF for the thermal energy group so that the lov energy resonances of the plutonium isotopes will be j 1-A

l 1

Table 3-2 l FINE ENERGY GROUP CROSS SECTION LIBRARY CONSTITUENTS HYDROGEN-1 PROMETHIUM-149 BORON-10 SAMARIUM-149 BORON-11 URANIUM-234 CARBON-12 URANIUM-235 NITROGEN-14 URANIUM-235 .

OXYGEN-16 URANIUM-238 SODIUM-23 NEPTUNIUM-237 NATURAL MAGNESIUM N1.PTUNIUM-239 ALUMINUM-27 PLUTONIUM-239 NATURAL SILICON PLUTONIUM-240 NATURAL CHLCRINE PLUTONIUM-241 NATURAL POTASSIUM PLUTONIUM-242 NATURAL CALCIUM AMERICIUM-241 NATURAL CHROMIUM AMERICIUM-243 MANGANESE-55 BURNABLE POISON (B10)

NATURAL IRON NON-SAT U233 FISSION PRODUCTS NATURAL NICKEL RAP-SAT U233 173SION PRODUCTS NATURAL ZIRCONIUM SLOW-SAT U233 FISSION ?CDUCTS NATURAL MOLYBDENUM NON-SAT U235 FISSION PRCDUCTS SILVER-107 RAP-SAT U235 FISSION PRODUCTS SILVER-109 SLOW-SAT U235 FISSION PRODUCTS CADMIUH-113 NON-SAT PU239 FISSION PRODUCTS 10DEsE-135 RAP-SAT PU239 FISSION PRODUCTS XENON-135 SLOW-SAT PU239 FISSION PRODUCTS l

l L

3-7

l L included in the thermal group. Reactor cores containing significant quan-

, tides of plutonium are represented mora securately when these low energy resonances are included in the thermal group.

NULIF calculates the neutron flux in the unit call for each of 31 fast and 80 thermal energy fine-groups. Macroscopic and microscopic cross i

1 sections are then d stermined for one fast and one thermal energy group by 2

collapsing there 111 fine-groups based on the neutron flux and cross sec-tions calculated for each fine-group. Cross sections are collapsed into two groups for use in PDQ07 calculations because it has been determined that the 1

use of two groups is adequate for large thermal reactors (such as the Surry i reactors) and the use of more emergy groups in FDQ07 would result in sub-stantially longer computer execution times.

The neutron energy spectrum calculated by NULIF for a unit cell r depends on the material concentrations (i.e,, the nuclide concentration or j number density) in the unit cell. The material concentrations change during the operation of the reactor as a result of:

1) Depletion of the material f 2) Changes in the soluble boron (chemical shim) and xenon concen-trations l

3) Changes in material temperature l

i The neutron spectrum is also dependent on the temperature of the fuel due to Doppler broadening of the resonance absorption peaks. The NULIF code is used to calculate the effect of both changes in material concentrations and in the fuel and moderator temperatures on the neutron spectrum and spectrum-weighted two-group cross section.

I NULIF calculates the depletion of unit cells baced on the spectrum-weighted neutron cross sections and the neutron flux. As the material is de-plated, the material concentrations change. This change in concentrations i

affects both the neutron flux and the neutron spectrum and, therefore, requires the frequent recalculation of the spectrum-weighted cross sections.

3-8

3.2.3 HAFIT

The BAFIT computer code is a data manipulation code which is used to prepare HARMONY cross section table sets for input to PDQ07. The HARMONY cross section interpolation and depletion routines have been incor-porated into the PDQ07 computer code to represent spectrum-weighted cross i sections and to perform material depletion calculations. The input to the HAFIT program consigts of a magnetic computsr tape containing the spectrum-weighted cross sections calculated by the NULIF code, and a description of how these cross sections are to be used to create a set of HARMONY tables for input to PDQ07. An automated data processing code like HAFIT must be used to prepare the HARMONY table sets for PDQ07 because of the substantial volume of data involved.

The HARMONY system used in PDQ07 for representation of neutron cross sections is based on the following equation:

, m I " + #

t g

t g i=1 1t 1,g t i,g where t = type (transport, absorption, re2s,v41, fission) of cross section g is the neutron energy group N is the material concentration for nuclide i G is the self-shielding factor m is the total number of nuclides I and a are the macroscopic and microscopic cross sections e

E is either a fixed value or interpolating table for the g

g macroscopic cross section The HARMONY system is designed to represent cross sections as a function of several independent variables. The macroscopic, microsec,pic, and 3-9

G-factor terms in the above equation can be input either as rixed values or in tabular form as a function of up to three independent variables, such as fuel burnup, xenon concentration, and soluble boron concentration.

The G-! actors are not currently used in the PDQ07 discrete model, since the self-shielding effect is already represented in the cross sections obtained from the NULIF code.

, For fuel rod cells, it is necessary to represent cross sections as a function of five independent variables - burnup, boron concentration, xenon concentration, average fuel temperature, and average moderator temperature. Since only three variables (i.e. , burnup, xenon concentration, and boron concentration) can be represented in the HARMONY system, a mechanism for extending this representation (i.e. , to include fuel and moderator temperature) is necessary. This is accomplished throuBh the use of " pseudo" cross sections which are actually partial derivatives of the macroscopic cross section as a function of either moderator t4sperature or the square root of fuel te=perature. These " pseudo" cross sections are treated in the

/ same way as normal cross sections except that the nuclide number density for these cross sections is either the difference between the moderator temperature for the fuel cell and the nominal average moderator temperature or the difference between the square root of the average fuel te=perature for the fuel cell and the square root of the nominal average fuel temperature.

~

The macroscopic cross sections are determined for the nominal average fusi and moderator temperature and then corrected to the actual values through use of the appropriate " pseudo" cross sections.

For each type of table used in the HARMONY system, a table (referred to as a " mask") must be generated which gives values of the independest variables (for example - burnup, borog and xenon concentration) 3-10

for each dependent variable data entry. The dependent variables in the HARMONY system, which are the macroscopic and microscopic cross sections and G-factots, are determined for independent variable values which do not appear in the mask by interpolation between the values which are represented in the mask. A detailed description of the HARMONY system can be found i

t in Reference 2.

3.2.4 GENERAlION OF FUEL UNIT CELL CROSS SECTIONS:

The input to the NULIF computer code for a fuel unit cell consists of:

1) Fuel call dimensions (pellet diameter, clad inside diameter, clad outside diameter, and fuel rod pitch)

2) Material concentration for the fuel pellet, clad, gap, and moderator

3) Average temperatures for the ' fuel, clad, and moderator

4) Average power density '

5) Depletion description including the burnup values at which NULIF calculates the neutron spectrum l

NULIF ,:alculations are then made for the unit fuel cell to determine the >

dependence of the two-group, spectrum-weighted croas sec.tions for each fuel enrichment on:

1) Burnup

2) Soluble boron concentration

3) Xenon concentration >

4) Moderator temperature

5) Average fuel temperature Sets of HARMONY cross section tables based on these NULIF calmulations are prepared by the HAFIT code. These tables represent:

3-11

__ - --- -- l

1) Microscopic fast and thermal energy group absorption and fission cross sections as a fun: tion of burnup, soluble -

boron concentration, and xenon concentration

2) Macroscopic fast transport and removal, and thermal trans-port cross sections as a function of burnup, soluble boron concentration, and renon concentration

3) The effect of fuel and moderator temperature changes on the macroscopic cross sections 6

3.2.5 GENERATION OF BURNABLE POISON (BP) UNIT CELL CROSS SECTIONS:

The input of the NULIF computer code for the burnable poison (BP) cell consists of:

1) BP cell dimensions

2) Material concentrations for the stainless steel clad, Zir-caloy-4 guide tube, pyrex glass, and water

3) Average temperatures for each of the materials above t

The dependence of'the n&utron spectrum of the BP unit cell on the depletion of the Boron-10 in the pyrex glass is determined by performing calculations for a number of Boron-10 concentrations. A set of cross section tables based on these NULIF calculations is prepared by the HAFIT compu'ter code for use by the PDQ07 (EARMONY) code. These tables represent:

1) Microscopic fast and thermal energy group absorption cross sections as a function of Boron-10 concentration

2) Macroscopic fast transport and removal, and thermal trans-port cross sections as a function of Baron-10 concentration An adjustment to the cross sections input to PDQ07 for the BP unit cell is mado to correct for an under-prediction of the flux depression in the BP unit cell (i.e. , over-prediction of the flux) in the PDQ07 com-puter code caletlations. The high thermal absorption cross section for a BP unit cell ccmpared to a fuel unit cell causes a depression of the thermal s

flax in the BF unit es11 and consequently, diffusica theory under-predicts this flux depression when only one mesh block is used to represent each BP unit cell.

a e

3-12

To correct for this, an adjustment factor is determined by (1) comparing PDQ07 quarter assembly calculations with 36 .nesh blocks (6 x 6) fcr each unit cell with calculaticns using one mesh block pe.r unit cell and (2) correcting the under-prediction of the flux depress 4.on with the. one mesh block per unit call representation by reducing the thermal absorption 1

cross sections in the BP unit cell, so that the thermal neutron absorption rate in the BP cell is the same for both the one mesh block par unit cell and the 36 mesh blocks per unit cell representations.

3.2,6 GENERATION OF CONTROL ROD UNIT CELL CROSS SECTIONS:

The input to the NULIF computer code for a control rod unit cell consists of:

1) Control rod cell dimensions

2) Material concentrations for the control rod. (Ag, In, Cd), stain-less steel clad, Zircaloy-4 guide tube, and water

3) Average temperatures for each of the above materials The NULIF calculations for a control rod determine the macroscopic two-group cross sections for the control rod cell which are input to the PDQ07 computer code.

i An adjustment to the control rod unit cell cross sections is calculated in the same way as the adjustment to the BP unit cell cross sections l

in order to account for a similar under-prediction of the flux depression.

3.2.7 GENERATION OF CONTROL ROD GUIDE TUBE UNIT CELL CROSS SECTIONS:

The input to the NULIF computer code for a control rod guide.

tube unit cel] consists of:

1) Control rod guide tube cell dimensions i

2) Material concentrations for Zircaloy-4 guide tube, and water i

3) Average temperatures for each of the above materials b

0 3-13

The neutron spectrum for tne control rod guide tube unit cell is calculated l for sevsral soluble boren concentrations to determine the spectrum-weighted cross sections as a function of solubla boron concentration.

A set of cross section tables based on these NULIF calculations is prepared by the HAFIT code to represent the control rod guide tube unit cell cross sections as a function of soluble boron concentration.

3.2.8 GENERATION OF BAFFLE CROSS SECTIONS The cross sections for the stainless steel baffle surrounding the core are calculated by NULIF. The neutron spectrum and spectrum-weighted cross sections for the baffle are more difficult to calculate aue to the three very different material compositions in the baffle area (i.e., fuel cells, stainless steel, and water). In addition, the baffle has a relatively high the.rmal absorption cross section, and as previously discussed, diffusion theory will not accurately calculate the neutron flux at material interfaces or in highly absorbing media even with correct spectrum-weighted cross sections. Therefore, the cross sections calculated by NULIF for the baffle have to be adjusted to give more accurate calcu-lations of the power density for peripheral assemblies.

The adjustment to the cross sections for the baffle is based on calculations of relative power distributions with the PDQ07 computer code.

The power distribution (as calculatad by the PDQO7 computer code for the beginning of life condition of Cycle i for both Surry units) was compared with the corresponding measured power distribution. The cross sections for the baffle were then adjusted to improve the power distribution com-parison for this situation. These adjus'ted baffle cross sections have provided accurate power distribution calculations for Surry Units 1 and 2, Cycle 1 and 2 as demonstrated in Section 4.

3-14

3.2.9 GENERATION OF REFLECTOR CROSS SECTIONS The cross sections in the reflector region for the Surry reactors e

f~

are calculated by hULIF. The reflector region of the Surry ceres extends from the outside of the core baffle to the reactor vessel wall, including the thermal ehield and core barrel. The stainicss steel and water in this region of the reactor are homogenized (volume-weighted) in NULIF, and a neutron spectrum and spectrum-weighted cross sections are calculated for this region. These calculations are performud for several soluble boron 4

concentrations, and tables representing the cross se.ctiocs of the reflector region as a function of soluble boron concentration are prepared for use l in PDQO7.

3.3 PDQO7 i

} 3.

3.1 INTRODUCTION

The PEQ07 computer code, as used in the PDQO7 discrete calcula-1 J

tional model, is a two-dimensional, two-group diffusion-depletion program which is used to calculate the neutron flux, power, and naclide concentra-tions as a function of radial (x-y) position and burnup. The PDQ07 computer code utilizes the appropriate microscopic and macroscopic :: as sections calculated by NULIF (and as properly formated by HAFlT) along with the initial description of the reactor (i.e., geometry and material composition de.scription) to calculate the neutron flux distribution at discrete spatial mesh points (and for two energy groups) at the desired core power. The spatially dependent neutron flux is then combined with the appropriate nuclide concen-trations and cross sections to obtain the spatially dependent power distri-bution. Once the initial spatially dependent flux and power distributior.s are obtained, the depletion of the nuclide concentrations is calculated.

i l

l 2

3-15

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

i 3.3.2 GEOMETRY DESCRIPTION:

A detailed description of the geometry input and terminology for the CDQ07 code is given in Section 3 of Reference 2. The geometry description for the discrete PDQ07 computational model is summarized

, in this section.

One quarter of the Surry reactor core is represented in two dimensions in the discrete PDQ07 model. Only one quarter of the core has i

tc be represented, because of the quarter core symmetry of the fuel loadic; for the Surry core.

The boundary conditions used in the solution of the diffusion equation in the PDQ07 code are:

1) Zero neutron current for the boundaries located along the core axes y

2) Zero

. wall.neutron flux for the boundaries located at the reactor vessel

[

The PDQ07 code uses one mesh line and therefore, one mesh block, f for each unit cell (i.e., fuel rod, burnable poison (EP) rod, control rod, I

s and control rod guide tube) in the Surry core. The small water gap between assemblies 10 also represented with one mesh line, so that the increased l flux peaking that takes place in the water gap will be calculated I

l l

correctly. The stainless steel baffle at the outside edge of the fuel

! in the Surry cora 1: represented with two mesh lines, because the baffle is approximately two fuel rod pitches in thickness. The reflector region j' outside the baffle of the Surry core is represented with nine mesh lines which extend the region of solution of PDQ07 out to the reactor vessel wall where the zero neutron flux boundary condition is applied. This geometric dascription results in a total of 17,424 mesh blocks for the f quarter core representation, including one for each fuel rod, control rod, 4

3-16

l BP rod, and control rod guide tube.

This detailed two-dimensional representation is required so that the relative radial pcvar density of each fuel rod can be calculated in order to determine the radial peaking factor and the enthalpy rise hot channel factor.

3.3.3 DEPLETION EQUATIONS:

Each mesh block in the PDQ07 code contains a single homogenous composition. The volume-weighted nuclide concentrations for each rash block in the Surry core are input to PDQ07 for beginning of life core conditions. 2 In addition, a set of equations, which is used by PDQ07 to calculate the change in nuclide concentrations with burnup, is input to PDQ07 for each j

i different composition in the Surry core.

l The appropriate set of material (nuclide) depletion equations is assigned in PDQ07 to each mesh block. These equations are used b.y PDQ07 to deplete the nuclide concentrations in each mesh block based on:

1) The average fast and thermal energy group neutron fluxes calculated by PDQ07 for the mesh block

2) The spectrum-weighted fast auf thermal group absorption cross sections determined by PDQ07 from the cross section table set assigned to the mesh block The depletion chains described with thece depletion equations in the discrete model for each unit cell type are shown in Table 3-3. A detailed description l

of how the depletion equations are input to PDQ07 describing these depletion chains is given in Section 5 of Reference 2.

t 3.3.4 THEPEAL-HYDRAULIC FEEDBACK PARAMETERS:

The input to PDQ07 required for thermal-hydraulic feedback consists of:

1) Coolant inlet enthalpy

2) Beated perimeter per unit area of flow 3-17 j

4 Tebis 3-3 Depletion Equations used in PDQ07

1. Neutron Absorptions Not Leading to Fission 38

a. U , Pu depletion chain 238 n 239 U N P

5hPu239 n m 240 n m 241 a m 242 a 243 e Pu , Pu , Pu ,m Am 241 235 i b. U depletion chain 235 n m 236 a m 237 i U U rNp 4

a

2. Neutron absorptions which produce fission are represented with the following fission products:

135 4

a. I g,135, p,149, and Sm 149 whicit are represented explicitly

b. Two groups of fission products which eventually build up to an equilibrium concentration (since they are created by fission reactions and destroyed by decay reactions). One group is characteristic of fission reactions by uranium isotopes and the other group is characteristic of fission reactions by plutonium

, isotopes.

c. Two groups of non-saturating fission products which are either stable isotopes or have half-lives greater than a few years.

Again, one group is characteristic of fission reactions by uranium isotopee and the other group is characteristic of fis-sion reactions by plutonium isotopes.

3-18

3) Hydraulic diameter of the channels

4) Flow area of the fuel assembly per total area of flow

)

5) System pressure-

6) Difference between average fuel temperature and moderator

{ temperature as a function of relative power denisty The strategy used in the feedback calculation consists of first making an initial estimate of the fuel and moderator temperatura for each coolant channel. Based on this initial estimate and the cross section tables for each fuel cell, the PDQ07 code calculates the two-group, spec-trum-weighted cross sections for each mesh block. These cross sections are used in a diffusion theory calculation of power density in each fuel rod cell. This power density is then used in a calculation of the fuel and moderator temperature for each fuel cell. In turn, the new fuel and moderator temperatures are used to calculate new two-grcup, spectrum-weighted cross sections for another diffusion theory power distribution

\

calculation. This process is continued until the power density for each fuel rod in the Nth iteration differs from the power density in the N-1 iteration by less than the convergence criterion.

Thermal-hydraulic feedback effects are represented in the PDQ07 model in order to more accurately calcula:e the power and burnup distributions. .

3-19

i SECTION 4 - COMPARISON OF MODEL PREDICTIONS WITH MEASUREMENT DATA

4.1 INTRODUCTION

The purpose of this section is to present a comparison between the analytical predictions determined with the PDQO7 discrete model and measure-ment data obtained from the Surry reactors. Measured reacecr data encompassing an initial and reload cycle operation are used in these comparisons to demon-s strate both the accuracy and flexibility of the PDQ07 discrete model. A summary of the comparisons is given in Table 4-1.

In addition to comparisons between the Vepco model and reactor measurements, comparisons are made between vendor model predictions and the same measurement: data. For these comparisons, only the average absolute value of the percentage difference between the vendor predicted and measured data is given. The comparisons are given so that the accuracy of the Vepco pre-dictions can also be compared with the accuracy of the predictions from an accepted and verified vendor model which has been used in the design, licensing, and core-follow of the Surry reactors.

)

4.2 ANALYTICAL CALCULATIONS Calculations presented in this section can be conveniently divided into power distribution calculations and reactivity calculations.

The power distribution calculations include fuel assembly average and peak fuel rod (within sach fuel assembly) relative radial power distri-butions representative of both initial and reload cycle operation, batch burn-up sharing, and assemblywise burnups during cycle operation. (it should be noted that the peak fuel rod relative power distribution in the core is con-sidered to be equivalent to the enthalpy rise hot channel factor (F ) for 4-1

Table 4-1

SUMMARY

OF COMPARISONS FOR BOTH THE INITIAL AND RELOAD CYCLES Core Condition

• Parameter Compared HZP-BOL Critical boron concentrations for various control rod bank config-urations (i.e., inserted or not inserted)

' Core radial power distributions for various control rod bank configurations Control rod bank worths e

Differential Boron Worth i

HFP-BOL-ARO Critical boron concentrations Core average radial power distributions t

HFP-BOL to EOL-ARO Critical boron concentration Core radial power distributions Burnup sharing distribution i

• HZP-BOL: Hot zero power - beginning of life HFP-BOL-ARO: Hot full power - beginning of life - all rods out HFP-BOL to EOL-ARO: Hot full p ver - depletion from beginning of life to eno of life - all rods out - equilibrium xenon 4-2

l average core conditions.) The power distribution calculations for cycle l

operation are calculated at various core operating conditions and burnup intervals. Between these intervals, the core is depleted using the +

power distribution calculated at the beginning of the depletion step.

Reactivity calculations presented in this section include cri-tical soluble boron concentrations, differential boron worth, and integral control rod bank worth. The critical soluble boron concentration is that con-centration of boron which maintains the reactor just critical and is obtained by determining the core eigenvalue (or K,ff) from a calculation l

i using a best-estimate boron concentration and then correcting this boron concentration to a value which corresponds to the just critical condition.

! Differential boron worth is obtained by varying the soluble boron concen-

[ tration about the critical baron concentration and then determining the re-l sultant impact on K df. Since the only change made between calculations is in the soluble boron concentration, the change in reactivity due to changes in boron concentration (i.e., the differential boron worth) can l

be directly calculated. Control rod worths are calculated in similar manner except that the soluble boron concentration is held constant l

l while specific control rod bank positions (e.g., all rods out, D bank fully inserted, C and D bank fully inserted) are changed. The control I

rod bank change produces a corresponding change in core reactivity which is directly correlated to control rod worth.

The reactivity and power distribution calculations were per-formed for the following reactor conditions:

1) Rot (547 F), zero power with all rods out, no xenon, D-Bank in, and C and D-Bank in

2) Hot (566 F), full power (2441 Mwt), no xenon and equilibrium xenon, zero and 150 MWD /EIU burnup with all rods out, D-Sank in, and C and D-Bank in 4-3

3) Hot (5660F), full power (2441 Mwt), equilibrium xenon, depletion from 150 MWD /MTU to end of life with all contrcl rods out.

i 4.3 MEASUREMILNT DATA Measurement data is obtained for the Surry units from routine physics testing conducted during the startup of each cycle of operation and from routine core performance monitoring conducted during the depletien of each cycle.

Differential sud integral control bank worths are measured by maintaining the reactor approximately critical and monitoring reactivity changes during exchanges of boron concentration with control rod bank position. Specifically, following the establishment of a constant boron dilution /boration rate, one of the conkrol rod banks is periodically inserted / withdrawn in order to provide reactivity compensation for the changing primary coolant system boron concentration. The reactivity changes resulting from the control bank movements are recorded on a continuous basis by the reactivity computer at the reactor. The differential control rod bank reactivity worth is defined as the ratio of the change in reactivity to the corresponding change in bank position about an average bank position, and the integral worth is then obtained by sinmnhg the individual reactivity changes between measurement endpoints.

The primary coolant critical boron concentration is monitored during startup physics testing when the control rod banks reach their fully inserted (or withdrawn) endpoints during the control rod bank worth measure-ments. For this measurement, the reactor conditions are stabilized, and the base just critical boron concentration is determined. To this base value, a slight adjustment for control rod position is made in order to obtain a just critical baron endpoint at the exactly desired control rod bank 4-4

configuration. The critical boron concentration is also monitored frequently (i.e., several times a day) daring cycle operation. The FOLLOW 8 computer code is used to normalize measured critical boron concentrations obtained during cycle operation to design conditions. (See Appendix A for a more detailed discussion of the FOLLOW computer code.) The normalization process takes into consideration actual control rod configurations, xenon and samarium concentrations, reactivity coefficients, and power levels.

The differential baron worth measurement is also made concurrent with the control rod bank worth measurements. For this measurement, fre-quent (i.e., every fifteen minutes) primary coolant boron concentrations are obtained and control rod bank positions are noted during the dilution and boration phases of the control rod bank worth measurement, Since che control rod bank positions as a funtion of time can be related to integrated reactivity, a relationship (graph) can be constructed of boron concentration as a function of integrated reactivity. The slope of this relationship is the differential baron worth.

The core power distributions are measured through the use of the movable detector flux mapping system. This system consists of five fission detectors which can traverse assembly instrumentation thimbles in 50 core locations. For each traverse, the detectoi output is continuously recorded on a strip chart and is also scanned for 61 discrete axial points by the on-site process computer. Two and three-dimensional core power distribu-tions are then determined using the INCORE 9 computer code. (See Appendix A for a more detailed discussion of the INCORE code.) INCORE couples the experimental flux map measurements from the on-site process computer with analytical power-to-flux ratios and power distribution data for each t:easured and unmeasured location to determine the power distribution for the entire '

4-5

core,includingassemblyaverageandpeak, rod (FfH)relativepowerdistri-bution for each assembly.

The analytical power-to-flux ratio and power distribution data used in the INCORE computer code are determined bf the PDQ07 discrete model. The PDQ07 discrete model calculates radial power distribution on a quarter core basis and stores the results of these calculations on u s-netic tape or disk (in addition to printing out these results) for use by several data handling codes which prepare the analytical data decks for the INCORE computer code. These data handling codes:

1) Expand the radial power distribution calculations for each rodded configuration from quarter core to full core representation

2) Determine the relative power density of the peak rod in each fuel assembly

3) Prepare input for t'ne INCORE code by giving the predicted radial power for each fuel assembly for each rodded configuration used in normal plant operation _

4) Prepare input for the INCORE code by giving the predicted relative radial power for the peak rod in each assembly. When the peak rod changes with control rod bank couliguration, the peak rod for each configuration is represented

5) Prepare input for the INCORE code by giving the= predicted fast and thermal group neutron fluxes for each instrument tube location The INCORE code then uses the above analytical data and the incore flux maps obtained from the movable detector flux mapping system to determine power distributions and' peaking factors.

The TOTE 10 code is used along with the INCORE code to burnup distributions. (See Appendix A for a more detailed discussion of the TOTE code.) Burnup rate information for each fuel assembly, which is obtained from the INCORE code, and core energy generatica rates, which are obtained from heat balances between the hot and cold legs of the secondary cooling system, are inputed 'into the TOTE code for the burnup i 1

4-6

distribution calculation.

4.4 RESULTS The results of the comparison between the power distribution and reactivity predictions and measurements obtained from the Surry units are pre-sented below. In addition, comparisons between the vendor model predictions and the same reactor measurements are presented.

For all of the power distribution comparisons, which are obtained from the incore movable detector flux mapping system, analytical INCORE data decks have been prepared with both the Vepco PDQ07 discrete model and the vendor model. The accuracy of calculated assembly average power distribution compared to the measured distribution is determined with the INCORE code by calculating the standard deviation (a) between measured and predicted assembly average radial power distributions with the following equation:

157 e={1f6 i=1 (i ~ I ) Ib Where: I[ is the calculated assembly average power for the ith assembly and I{isthemeasuredassemblyaveragepowerfortheithassembly The standard deviation of the assembly average power distribution provides a mathematical basis for evaluating the accuracy of the model.

The specific types of results compared are delineated in Tab.le 4-2.

For all comparisons, both the measured and the Vepco model predicted values are presented in addition to the standard esviation or percentage difference in these values. For comparisons with the vendor model, only the average absolute value of the standard deviation or percentage difference for all measurements is given. Representative INCORE output comparing the Vepco PDQ07 discrete model predictions and measurements is provided in Appendix 8.

4-7 L_ _. _ - - - - - - - - - - - - - - - - - - - - - - - - -

1 cura 4-1 TYPE OF COMPARISONS REACTOR Coh.. REFERENCE FCd CCMPARISONS

'TO'd Power Distribution:

Assembly Average Cycle Operation for Unit 1, Initial Cycle Table 4-3 Cycle Operation .for Unit 1, Reload Cycle Table 4-4 Cycle Operation for Unit 2, Initial Cycle Table 4-5 Cycle Operation for Unit 2, Reload Cycle Table 4-6 Peak Rod F yc e Operation for Unit 1, Initiai Cycle Table 4-7 H

Cycle Operation for Unit 1, Reload Cycle Table 4-8 Cycle Operation for Unit 2, Initial Cycle Table I,-9 Cycle Operation for Unit 2, Reload Cycle Table 4-10 Assemblywise Burnup and Cycle Operation for Unit 1, Initial Cycle Table 4-11 Batch Burnup Sharing Cycle Operation for Unit 1. Reload Cycle Table 4-12 Cycle Operation for Unit 2, Initial Cycle Table 4-13  ;

Cycle Operation for Unit 2, Reload Cycle Table 4-14  !

4-E Reactivity:

Critical Boron Concent ation Startup forUnits 1 & 2, Initial Cycle Table 4-15 I For Various Control Rod Con- Startup forUnits 1 & 2, Reload Cycle Table 4-16 figurdtions Critical Boron Concentration Cycle Operation for Unit 1, Initial Cycle Figure 4-1 vs. Burnup Cycle Operation for Unit 1, Reload Cycle Figure 4-2 Cycle Operation for Unit 2, Initial Cycle Figure 4-3

• Cycle Operation for Unit 2, Reload Cycle Figure 4-4 ' i Differential Boron Worth Startup for Units 1 & 2, Initial Cycle Table 4-17 Startup for Units 1 & 2, Reload Cycle Table 4-18 Control Rod Bank Worths Startup forUnits 1 & 2, Initial Cycle Table 4-19 l Startup forUnits 1 & 2, Reload Cycle Table 4-20 l

______-__ A'l

I Tahle 4-3 COMPARISON OF PREDICTED AND MEASURED ASSEMBLY AVERAGE POWER DISTRIBUTIONS FOR SURRY UNIT 1. CYCLE 1 M/D Map Power Control Rod Cycle Burnup Vepco Model Vendor Model Numbe r Level (%) Configuration (MWD /MTU) o (%) o (%)

6 0 ARO 71 2.4 #4 7 0 D-Bank In 71 2.1 11 0 C and D-Bank In 71 2.3 28 75 ARO 339 2.1 35 75 ARO 620 1.5 40 75 ARO 1150 1.5 43 92 ARO 1900 1.8 45 88 ARO 2380 1.5 -

,, 48 90 ARO 3175 1.6 Average 4 49 90 ARO 3550 2.1 absolute

~

51 91 ARO 4190 1.8 value 55 95 ARO 5275 2.4 2.0' 57 95 ARO 5975 1.9 59 94 ARO 6790 1.9 62 95 ARO 7540 1.2 65 94 'ARO 8435 1.4 66 100 ARO 9150 1.3 68 100 ARO 10100 1.3 69 92 ARO 10820 1.3 70 100 ARO 11875 2.8 71 100 ARO 12670 2.5 73 60 ARO 13415 2.1 g r i

\

Table 4-4 COMPARISON OF PREDICTED AND HEASIRED ASSEMBLY AVERAGE POWER DISTRIBUTIONS FOR SURPY UNIT 1, CYCLE 2 H/D Map Power Control Rod Cyc?e Burnup Vepco Model Vendor Model Number Level (%) Configuration (MWD /MTU) o (%) o (%)

1 3 D-Eank In 0 3.0 s '

2 3 ARO O 4.3 l 9 99 ARO 127 3.1 l 10 100 ARO 301 3.6 l Average 12 98 ARO 1103 l 13 100 3.0 absolute ARO 2043 2.8 value 16 99 3102 i

l ARO 2.1 2.8' 17 99 ARO 4015 2.1 18 100 ARO 4899 1.8 l

19 100 ARO 5612 1.5 j-20 100 ARO 6569 1,4 d

5 0

Table 4-5 COMPARISON OF PREDICTED AND MEASURED ASSEMBLY AVERAGE POWER DISTRIBUTIONS FOR SURRY UNIT 2, CYCLE 1 M/D Map Power I Control Rod Cycle Burnup Vepco Model Vendor Model l Number Level (%) Configuration o (%)

(MWD /MTU) o (%)

1 0 ARO O 2.7 d ' '

2 0 D-Bank In 0 2.1 9 0 C and D-Banks In 0 2.2 23 9G ARO 365 2.2 27 88 ARO 630 3.0 31 90 ARO 1300 1.3 34 88 ARO 2030 1.3 l 38 92 ARO 2950 2.9 l 40 92 ARO 3780 1.0 l 42 94 ARO 4670 1.8 Average e- 43 92 ARO 5240 1.7 abap*ute i

1 45 83 ARO 5940 1.2 48 value' 91 f.R0 6780 1.5 50 2.0 91 ARO 7725 1.9 52 98 ARO 8580 1.4 54 98 ARO 9310 1.4 59 100 ARO 9890 1.1 61 100 ARO 11025 1.3 62 100 ARO 11740 1.5 63 100 ARO 12770 2.1 64 98 ARO 13650 1.6 65 100 ARO 14520 2.0 v I

v.,,._.

s. . .. .. _ .

Table 4-6 COMP / ISON OF PREDICTED AND MEASURED ASSEMBLY AVERAGE POWER DISTRIBUTIONS FOR SURRY UNIT 2, CYCLE 2 l

M/D Map Power Control Rod Cycle Burnup Vepcc Model Vendor Model Number Level (%) Configuration (HWD/HTU) o (%) o (%)

1 2 ARO O 4.3 v 4 2 2 D-Bank In 0 4.2 6 100 ARO 170 1.7 7 100 ARO 730 2.2 Average 11 100 ARO 1875 2.5 ahaolute 12 97 ARO 2790 1.6 value-15 100 ARO 3750 1.2 1,7 16 100 ARO 4520 l 1.4 17 100 ARO 5650 1.4 '

20 100 ARO 6875 1.3 c- 21 100 ARO 7125 1.5 1

d. 22 100 ARO 8f~9 i

' " 1.6 I 23 99 %RO 86;d 1.5 g r J

1 l

1 l

)

~

s Table 4-7 COMPARISON OF PREDICTED AND MEASURED F FOR SURRY UNIT 1. LYCLE 1 Vepco Model Measured Predicted M/D Map Power Control Rod Cycle Burnup Number pH yN Vepco Model Vendor Model Level (%) Configuration (MWD /MTU) AH AH  % Difference %' Difference 6 0 ARO 71 1.363 1.331 -2.3 d 4 7 0 D-Bank In 71 1.418 1.406 -0.8 l 11 0 C and D-Bank In 71 1.752 1.734 -1.0 .

i 28 75 ARO 339 1.362 1.329 -2,4 i 35 75 ARO 620 1.373 1.345 -2.0

) 7 40 75 ARG 1150 1.377 1.364 -0.9 l g 43 92 ARO 1900 1.389 1.354 -2.5

, 45 88 ARO 2380 1.375 1.345

-1.5 48 90 ARO 3175 1.360 1.325

. -2.6 49 90 ARO 3550 1.367 1.317 -3.7 51 91 ARO 4190 1.347 1.304 -3.2 Average 55 95 ARO 5275 1.313 1.285 -2.1 absolute 57 95 ARO 5975 1.315 1.273 -3.2 valse 59 94 ARO 6720 1.271 1.260 -0.9 3.6 62 95 ARO 7540 1.277 1.251 -2.0 65 94 ARO 8435 1.256 1.241 -1.2 66 100 ARO 9150 1.254 1.233 -1.7 68 100 ARO 10100 '.. 228 1.277 -0.1 69 92 ARO 10820 1.232 1.223 -0.7 70 100 ARO 11875 1.250 1.214 -2.9 71 100 ARO 12670 1.218 1.207 -0.9 73 60 ARO 13415 1.208 1.200 -0.7 wr

/

Table 4-8 COMPARISON OF PREDICTED AND MEASURED F FOR SURRY 1, CYCLE 2 I Vepco Model M/D Map Power Control Rod Cycle Burnup Measured Predicted Number Vepco Model Vendor Model Level (%) Bank Location (MWD /MTU) F{H F$H  % Difference % Difference 4

1 3 E-Bank In 0 2'

1.653 1.675 1.3 4 3 ARO O d

1.439 1.480 2.9 9 99 ARO 127 1.394 1.386 -0.6 10 100 ARO 307 1.408 1.383 12 98 ARO

-1.8 Average 1103 1.430 1.368 13 100 ARO 2043

-4.3 absolute 1.416 1.371 -3.2 value 16 99 t.R0 3102 y 17 1.417 1.361 -4.0 1.4 99 ARO 4015 1.390 5: 18 1.350 -2.9 100 ARO 4899 1.390 1

19 1.340 -3.6 100 ARO 5611 20 1.373 1.337 -2.6 100 ARO 6569 1.363 1.332 -2.3 if i

l

Table 4-9 COMPARISON OF PREDICTED AND MEASURED F FOR SURRY UNIT 2, CYCLE 1 Vepco Model M/D Map Meaeured Predicted Power Control Rod Cycle Burnup Number Level (%) pH pH Vepco Model Vendor Model Bank Location' (MWD /MTU) 611 ZE  % Difference % Difference 1 0 ARO 2

O 1.350 1.331 -1.4 s k 0 D-Bank In 0 1.406 1.405 0.0 9 0 C and D-Bank In 0 1.704 1.734 +1.8 23 90 ARO 4 365 1.366 1.330 -2.6 27 88 ARO 6 30 1.354 1.345 -0.7 31 90 ARO 1300 1.367 1.363 34 88 ARO -0.3

,d 2030 1.379 1.355 -1.7 38 92 ARO

d. 2950 1.350 1.329 -1.6 40 92 ARO 3780 l 42 94

1. 34 1 1.312 -2.2 ARO 4670 1.331 43 92 1.295 , -2.7 Average ARO 5290 1.331 45 83 1.285 -3.5 absolute ARO 5940 1.302 48 91 1.274 -2.2 value ARO 6780 1.288 50 1.261 -2.1 3.5 91 ARO 7725 52 1.290 1.249 -3.2 98 ARO 8580 54 1.245 1.239 -0.5 98 ARO 9310 1.238 59 100 1.232 -0.5 ARO 9890 1.227 61 100 1.229 -0.2 ARO 11025 1.227 62 100 1.222 -0.4 ARO 11740 1.220 63 10 0 1.215 -0.4 ARO 12770 1.210 64 1.2G6 -0.3 98 ARO 13650 1.205 1.198 -0.6 65 100 ARO 14520 1.200 1 i90 -0.8 S F

3 _. .

Table 4-10 COMPARISON OF PREDICTED AND MEASURED F FOR SURRY UNIT 2, CYCLE 2 I Vepco Model Measured Predicted M/D Map Power Control Rod Cycle Burnup Vepco Model pN pF Vendor Model Number Level (Z) Bank Location (MWD /MTU) Ait All % Difference % Difference 1 2 ARO #'

O 1.426 1.374 -3.7 2 2 D-Bank In 0 1.610 1.550 -3.7 6 100 ARO 170 1.357 1.351 -0.4 l 7 100 ARO 730 1.372 1.352 -1.5 11 100 ARO 1875 1.361 1.341 -1.5 Average 12 97 ARO . 2790 1.331 1.332 +0.1 absolute 15 100 ARO ' 3750 1.325

' 1.371 -0.3 value

e. 16 100 ARO 4520 1.304 1.311 +0.5 1.2 J. 17 100 ARO 5650 1.306 i

"' 1.300 -0.5 20 100 ARO 6875 1.284 1.291 -0.5 21 100 ARO 7125 1.297 1.289 -0.6 22 100 ARO 8040 1.295 1.282 -1.0 23 99 ARO 8850 1.276 1.275 -0.1  % #

i

TABLE 4-11 ASSEMBLTWISE ACCMULATED BURNUP AND BATCH BURSUP SHARING (10 5'D/MI'U) FOR THE CTCLE 1 OPERATICN OF SU2RY UNIT 1 R P N M L E, J H C F E D C B A n

3.1? 10.42 8.17 5,41 10.48 8.35 1

+2.9 +0.6 ** 7 8.37 12.2: 14.03 12.53 14.03 12.22 8.37 8.74 12.1E 13.84 12.35 13.78 12.15 8.67 2

+4.4 -0.3 -1.4 -1.4 -1.8 -0.6 +3.6 8.97 13.33 14.25 13.86 15.30 13.86 14.29 13.33 8.97 -

9.23 3 13.47 14.44 13.81 14.99 13.82 14.15 13.50 9.46

+2.9 +1.1 +1.0 -0.4 -2.0 n1 1 n +1 n +< <

5 97 11.27 14.37 14. 15 15.92 14.69 15.92 14.19 14.37 11.27 8.97 4 9.08 11.51 14.58 14.4C 15.71 14.69 15.80 14.32 14.48 11.48 9.18

+1.2 +2.1 +1.5 +1.5 -1.3 0.0 -0.R +0.9 Ad R +1 o +9 1 3.37 13.32 14.37 14.28 16.02 14.98 16.45 14.98 16.08 14.18 14.37 13.32 8.37 5

8.52 13.07 14.30 14.37 15.85 14.87 16.31 15.09 15.97 14.23 13.89 13.06 8.67 -

+1.8 -1.9 -0.5 +1.3 -1.4 -0.7 -n o m? nv *n t -3:3 -2.0 +3.6 12.22 14.29 14.19 16.08 15.04 16.61 15.25 16.61 15.04 16.08 14.19 14.29 12.22 12.36 14.17 14.12 15.48 14.5E 16.40 15.36 16.44 14.76 6 15.87 14.07 14.07 12.27

+1.1 -0.8 -0.5 -3.7 -3.1 -1.3 +0.7 -1.0 -1.9 -1.3 -0.9 -1.5 +0.4 8.17 14.03 13.86 15.92 14.98 16.61 15.31 1 % 77 15.31 16.61 14.98 15.92 13.86 14.03 8.17! 7 8.54 14.20 13.98 15.84 14.77 16.26 15.28 16.63 15.19 16.44 15.19 16.09 14.01 13.67 8.2d

+4 5 +1.2 +0.9 -0.5 -1.4 -2 1 n? nm 0.8 -1.0 +1.4 +1.1 +1.1 -2.6 +1.4 10.4. 12.53 15.30 -14.62 16.45 15.25 16.77 15.37 16.77 15.25 16.45 14.69 15.30 l'2.53 10.42 12.78 15.55 14.93 16.41 15.2E 16.61 15.32 16.58 15.37 8

10. 55 16.43 15.00 15.44 12.58 10.35-

+4.5 +2.0 +1.6 +1.6 -0.2 +0.2 -1.0 -0.3 -1.1 -0 R n1 +9 1 *n o +o A -0.7 8.17 14.03 13.86 15.92 14.98 16.61 15.31 16.77 15.31 16.61 14.98 15.92 13.86 14.03 8.17 8.43 14.01 13.89 15~86 15.07 16.35 15.11 16.52 15.23 16.18 14.92 15.92 14.11 13.94 8.19 9

+3.2 -0.1 +0.2 -0.4 +0.6 -1.6 1.1 -1.5 -0.5 -2.6 -0.4 0.0 +1 R -o_A an.?

12.22 14.29 14.19 16.08 15.04 16.61 13.25 16.61 15.04 16.08 14.19 14.29 12.22 10 12.03 13.89 14.10 16.04 14.12 16.17 15.01 16.14 14.59 15.75 14.22 14.17 12.22

-1.6 -2.8 -0.6 -0.3 -2.1 -2.7 -1.6 -2.8 -3.0 -2.1 +0.2 -0.4 0.6 8.37 13.32 14.37 14.18 16.08 14.98 16.45 14.98 16.08 14.18 14.37 13.32 8.37 8.64 13.32 14.44 14.35 15.85 14.70 15.96 14.61 15.83 14.24 14.47 13.39 8.69 n

+3.2 0.0 +0.5 +1.2 -1.4 -1.9 -3.0 -2.5 -1.6 +0.4 +0.7 M.9 *.

8.97 11.27 14.37 14.19 15.92 14.69 15.92 14.19 14.37 11.27 8.97 12 9.41 11.69 14.46 14.18 13.53 14.44 15.54 14.08 14.35 11.54 9.26

+4.9 +3.7 +0.6 -0.1 -2.5 -1.7 -2.4 -0.E -0.1 +2.4 +3.2 8.97 13.33 14.29 13.86 15.30 13.86 14.29 13.33 8.97 13 9.22 13.15 14.07 13.73 14.91 13.65 14.06 13.19 9.23

+2.8 -1.4 -1.5 -0.9 -2.6 -1 9 1A 1 1 +2.9 8.37 12.22 14.03 12.53 14.03 12.22 8.37 14 8.60 12.62 14.36 12.43 13.62 12.15 8.48 a 7epco Model +2.R +3.3 +2.4 -0.8 -2.0 -0.6 +1.3 b MEASURED 8.17 10.42 8.17 e  % DIFFERENCE 8.75 10.75 8.19

+7.1 +3.2 +0.2 CORE AVERAGE BURNUP = 13.547 5'D/MTU Vepco Model Vepco Model Vendor Model D Measured Predicted Percent Difference Percent Difference Batch 1 14.25 14.20 +0.4 average Batch 2 15 46 15.62 -1.0 absolute o Batch 3 10.93 10.81 +1.1 value 1.3 '

4-17

TABLE 4-12 ASSIMBLT"ISE ACCUMU1ATED BURSUP AND BATCH BURNUP SHAEING (103 5'D/MIU) FOR THE CTCLE 2 OPERATION OF SURRY UNIT 1 R P N M L K J H C F E D C B A 4.92 6.26 4.92 1

5.00 6.16 4.82

- +1.6 -1.6 -2.0 5.25 7.16 6.99 20.49 6.99 7.16 5 . 25 6.22 7.05 5.38 2 5.47 7.28 6.35 20.19

+4.2 +1.7 -9.2 -1.5 -11.0 -1.5 +2.5 >'

I 5.69 7.62 20.04 21.49 8.64 21.49 20.04 7.62 5.69 6.02 7.44 20.19 21.57 8.27 21.41 20.03 7.39 6.04 3

-10.0 -2.4 +0.8 +0.4 -4.3 -0.4 -0.1 -3.0 +6.2 5.69 6.48 23.26 22.75 23.16 22.84 ?1.16 22.75 23.25 6.48 5.69 6.04 4 6.02 6.85 23.11 22.47 22.73 22.34 22.92 22.63 23.14 6.79

+5.8 +5.7 -0.6 -1.2 -1.9 -2.2 -1.0 -0.5 -0.5 +4.8 +6.2 5.25 7.62 73.26 23.69 7.51 22.11 8.60 22.11 7.51 23.69 23.26 7.62 5.25 -

7.67 23.53 22.98 7.43 5.77 5 5.48 7.42 22.97 23 0' l.54 21.98 8.25 21.95

+4.4 -2.6 -1.3 -0.2 +0.4 -0.6 -4.1 -0.7 +2.1 -0.7 -1.2 -2.5 49.9 7.16 20.04 32.75 7.51 18.31 7.97 19.64 7.97 18.31 7.51 22.75 20.04 7.16 7.61 6 7.22 19.82 22.61 7.56 18.44 8.04 19.42 8.17 18.47 7.62 22.84 19.94

+0.7 +0.7 +0.9 +2.5 +0.9 +1.5 +0.4 -0.5 +6.3 I

+0.8 -1.1 -0.6 -1.1 4.92 6.99 21.49 23.16 22.11 7.97 24.34 8.79 24.34 7.97 22.11 23.16 21.49 6.99 4.92 7 5.16 6.48 21.29 22.47 21.92 8.11 24.29 8.70 24.34 8.20 21.96 22.93 21.07 6.60 5.11

+1.8 -1.0 0.0 +2.9 -3.7 -1.0 -2.0 -5.6 +3.9

+4.9 -7.3 -0.9 -3.0 -0.9 -0.2

'6.26 26.49 8.64 22.84- 8.60 19.64 8.79 22.26 8.7) 1!.64 8.60 22.84 8.64 20.49 6.26 6.38 8 6.48 20.60 8.49 23.03 8.41 19.86 8.73 22.09 8.59 19.72 8.50 22.99 8.60 20.67

+3.5 +0.5 -1.7 +0.8 -2.2 +1.1 -0.7 -0.8 -2.3 +0.4 -1.2 +0.7 -0.5 +0.9 +1.9 4.92 6.99 21.'49 23.16 22.11 7.97 24.34 8.79 24.34 7.97 22.11 23.16 21.49 6.99 4.96 5.16 6.49 21.47 23.04 21.70 8.10 24.26 8.53 23.69 7.95 21.a2 22.67 21.54 6.51 5.07 9

+4.9 -7.2 -0.1 -0.5 -1.9 +1.6 -0.3 -3.0 -2.7 -0.3 -0.1 -2.1 +0.2 -6.9 +3.1 7.16 20.04 22.75 7.51 18.31 7.97 19.64 7.97 18.31 7.51 22.75 20.04 7.16 30 7.25 19,82 22.60 7.64 18.69 8.09 19.39 7.90 18.40 7.49 22.52 19.94 7.35

+1.3 -1.1 -0.7 +1.7 +2.1 +1.5 -1.3 -0.9 +0.5 -0.3 -1.0 -0.5 +3.1 5.25 7.62 23.26 23.69 7.51 22.11 8.60 22.11 7.51 23.69 23.26 7.62 5.25 11 5.53 7.48 23.08 23 55 7.65 21.85 8.33 21.76 7.61 23.43 22.85 7.44 5.55

-0.8 -0.6 +1.9 -1.2 -3.1 -1.6 +1.3 -1.1 -1.9 -2.4 +5.1

+5.3_ -1.8 5.69 6.48 23.26 22.75 23.16 22.84 23.16 22.75 23.26 6.48 5.69 12 6.17 6.95 22.91 22.35 22.85 22.31 22.72 22.25 22.86 6.94 6.07

+8.4 +7.3 -1.5 -1.8 -1.3 -2.3 -1.9 -2.2 -1.72 +7 . u . +6.7 5.69 7.62 20.04 21.49 8.64 21.49 20.04 7.62. 5.69 33 6.11 7.42 19.97 21.63 8.35 21.44 19.81 7.44 6.11

+7.4 -2.6 -0.4 +0.7 -3.4 -0.2 -1.1 -2.81 +7.4 5.25 7.16 6.99 20.49 6.99 7.16 5.25 I'

a vepco Model 5.49 7.43 6.53 20.16 6.71 7.46 5.41 b MEASURED +4.6 +3.8 -6.6 -1.6 -4.0 +4.2 +3.1 c  % DIFFERENCE 4.92 6.26 4.92 15 5.20 6.57 5.25

+5.7 +5.0 +7.3 CORE AVERAGE BURNUP = 6915 ETJNIU Vepco Nbdel Vepco Nbdel Vendor Model Measured Predicted Percent Difference Percent Difference Batch 1A 19.63 19.82 -1.0 Average Batch 2 22.60 22.80 -0.9 absolute batch 4A 7.60 7.49 +1.5 value 2.8 Batch 4B 8.40 8.19 +2.6 Batch 4C 6.25 6.27 -0.3 4-18

a mes.t. ae u s ASSEMBLTWISE ACC"g MU LATED BURNUP AND BATCH BURNUP j

SRARTNG (10 MWD /MTU) FOR THE CYCLE 1 OPERATION OF SURRY UNIT 2 R P N M L . r. J H C F E' D C B A 8.96 11.40 8.96 l 9.34 11.50 9.11 1

+4.2 +0.9 +1.7 9.82 13.40 15.40 13.75 15.40 13.40 9.82 9.57 13.58 15.34 13.60 15.13 13.31 9.41 2

-2.6 +1.3 -0.4 -1.1 -1.8 -0.7 -4.2 9.87 14.69 15.73 15.21 16.80 1.5.21 15.73 14.69 9.87 10.09 14.63 15.65 15.15 16.56 15.08 15.62 14.68 10.15 3

+2.2 -0.4 -0.5 -0.4 -1.4 -0.9 -0.7 -0.1 +2.8 9.87 12.42 15.83 15.59 17.48 16.10 17.48 15.59 15.53 12.42 9.87

. 10.06 12.41 15.62 15.42 17.35 16.03 17.45 15.61 15.97 12.53 10.13 4

+1.9 -0.1 -1.3 -1.1 -0.7 -0.4 -0.2 +0.1 40.9 40.9 A2.6 9.82 14.69 15.83 15.58 17.65 16.40 18.02 16.40 17.65 15.58 15.83 14.69 9.82 9.50 14.56 15.69 15.39 17.28 16.20 17.98 16.43 17.56 15.62 15.76 14.79 9.66 5

-3.3 -0.9 -0.9 -1.2 -2.1 -1.2 -0.2 +0.2 -0.5 +0.3 -0.5 +0.7 -1. 6 -

13.40 15.73 15.59 17.65 16.47'18.19 16.68 18.19 16.47 17.65 15.59 15.7" 13.40 13.69 15.71 15.44 17.30 15.98 17.96 16.62 18.09 16.15 17.49 15.49 15.67 13.56 *

+7.2 -0.1 -1.0 -2.0 -3.0 -1.3 -0.4 -0.6 -1.9 -0.9 -0.6 -0.5 +1 '

8.96 15.40 15.23 17.48 16.40 18.19 16.74 18.35 16.74 18.19 16.40 17.48 15.21 15.40 8.'96 7

9.47 1*. 57 15.22 17.45 16.26 17.94 16.51 18.21 16.65 18.15 16.27 17.26 15.24 15.22 9.20

+5.7 +1.1 +0.2 -0.2 -0.9 -1.4 -1.6 -0.8 -0.5 -0.2 -0.9 -1 1 A0.2 1? A?.7 11.4C.13.75 16.80 16.10 18.02 I 16.68 18.35 16.80f18.35 16.68 18.02 16.10 16.80 13.75 11.40 11.7" 13.85 16.91 16.17 18.10' 16.57 18.11 16.5q18.22 16.73 17.91 16.07 17.03 13.90 11.35 8

+3.: +0.7 +0.7 '+ 0.4 +0.4 -0 . -1.3 -1.3 -0.7 +0.3 -0.6 -0.2 +1.4 +1.1 +1.3 8.9C 15.40 15.21 17.48 16.40 18.1 16.74 18.35 16.74 18.19 16.40 17.48l15.21 15.40 0.56 9.45 15.54 15.22 17.52 16.34 18.0 16.47 18.12 16.49 17.86 16.22 17.35115.34 15.54 9.25 9

+5.5 +0.9 +0.1 +0.2 -0.4 -0. -1.6 -1.2 -1.5 -1.8 -1.1 -0.7I +0.9 +0,9 +3.2 13.40 15.73 15.59 17.65 16.47 18.19 16.68 18.19 16.47 17.65 15.59 15.73 13.40 13.48 15.62 15.52 17.48 16.09 17.90 16.47 17.85 15.98 17.31 15.34 15.53 13.53 '0

+0.6 -0.7 -0.15 -1.0 -2.3 -1.6 -1.2 -1.9 -3.0 -1.9 -1.6 -1.3 +1.0 9.82 14.69 15.83 15.58 17.65 16.40 18s02 16.40 17.65 15.58 15.83 14.69 9.82 9.60 11 9.58 14.82 15.91 15.48 17.4C 16.22 17.9C 16.22 17.40 15.49 15.99 14.88

-2.4 +0.9 +0.5 -0.6 -1.4 -1.1 -0.7 -1.1 -1.4 -0.6 +1.0 +1.3 -2.2 9.87 12.42 15.83 15.59 17.48 16.1C 17.48 15.59 15.83 12.42 9.87 10.29 12.65 15.86 15.44 17.24 15.9: 17.28 15.43 15.93 12. 8 ', 10.34 12

+4.3 +1.9 +0.2 -1.C -1.4 -1.1 -1.1 -1.0 +0.6 +3.4 +4.8i 9.87 14.69 15.72 15.21 16. 8C 15.21 15.73 14.69 9.87 10.10 14.51 15.52 15.02 16.5I 14.98 15.58 14.78 10.37 13

+4.3 -1.2 -1.2 -1.3 -1. ! -1.5 -1.0 +0.6. +5.1 9.82 13.4C 15.40 13.7f 15.40 13.40 9.82 14 9.39 13.4C 15.18 13.4: 15.08 13.40 9.41 a Vepco Model -4.4 0.0 -1.4 -2. -2.1 0.0 -4.2 b MEUURED 8.96 11.4C 8.96 c 2 DIFFERENCE 15 l

~~

9,og 11.21 9.04

+1.5 -1.] +0.9 CORE AVERAGE BURNUP = 14,870' MWD /MrU Vepco Model Vepco Model Vendor Model Measured Predicted Percent Difference Percent Difference

(

Batch 1 15.47 15.57 -0.6 Avarage Batch 2 17.02 17.15 -0.8 absolute Batch 3 12.04 11.88 +1.4 value 1.4 4-19

TABLE 4-14 ASSEMBLYW LE ACCUMCLATED BURNUP AND BATCH BURNUP SHARING (103 WD/MTU) FOR THE CTCLE 2 OPERATION OF SURRY UNIT 2 R P N M I. Y. J H '. F E D C a A 5.74 6.98 5.74 I 5.96 7.10 5.93 -

1

+3.8 +1.7 +3.3 5.98 8.50 10.34 24.71 10.34 8.50 5.98 '

6.26 8.27 10.37 24.79 10.33 8.62 6.18 2

+4.7 -2.7 +0.3 +0.3 -1.0 +1.4 +3.3 6.81 10.08 24.04 24.39 26.25 24.29 26.04 10.08 6'.S1 7.07 9.96 25.90 2/. 28 26.04 24.53 26.08 10.06 70 3 j

+3.8 -1.2 -0.s -t s .n a sc e, - -0.2 +3.11 6.81 9.90 25.13 11.10 25.05 10.94 25.05 11.10 25.13 9.90 6.81 7.07 10.04 24.78 10.97 24.85 10.84 24.86 11.10 25.30 9.99 7.03 4

+3.8 +1.4 -1.4 -1.2 -0.8 -0.6 -0.8 0.0 +0.7 .O . 9 +3.0 l 5.98 10.08 25.13 11.16 27.20 11.11 27.17 11.11 27.20 11.16 25.13 10.08 5.98 8 6.21 10.05 25 .0 1 11.13 26.81 10.83 27.05 10.98 27.02 11.10 24.94 10.05 6.34 5

+3.9 -0.3 -0.5 -0.3 -1.4 -2.8 -0.4 -1 2 -0.7 os -n m n1 +4 n 8.50 26.04 L1.10 27.20 25.10 26.51 21.82 26.51 25.10 27.20 11.10 26.04 8.50 8.59 25.98 L1.06 26.81 24.99 25.98 21.80 26.24 24.53 27.00 10.91 25.83 8.73 6

+1.1 -0.2 -0.4 -1.4 -0.4 -2.0 -0.1 -1.1 -7 1 -n 7 1 ? ne +?  ?

5.74 10.34 24.39 25.05 11.11 26.51 10.94 25.48 10.94 26.51 '.1.11 25.05 24.27 10.34 5 . 7 '.

6.07 10.43 24.20 24.90 10.95 25.98 10.78 25.32 10.68 26.20 10.95 24.83 24.44 10.32 5.95 7

_+5.8 +0.9 -0.8 -0.6 -1.4 -2.0 -1.5 -0.6 -2.4 -1.2 -1.4 -0.9 +0.4 -0.2 +3.7 y 6.98 24.71 26.25 LO.91 27.17 21.82 25.48 24.08 25.48 21.82 27.17 10.91 26.25 24.71 6.38 7.26 24.55 26.07 I3.85 27.20 22.00 25.55 23.60 25.20 21.93 27.02 10.67 26.19 24.60 7 12I ,

+4.3 -0.7 -0.7 -0.6 +0.1 +0.8 +0.3 -2.0 -1.1 +0.5 -0.C -0.46 40.2 -0.5 +2.0 5.74 10.34 24.39 25.05 11.11 26.51 10.94 25.48 10.94 26.51 11.11 25.05 l24.39 10.34 5.74 6.09 10.51 24.29 14.85 10.88 26.14 10.76 25.04 10.71 26.00 10.96 24.69 4.63 10.57 6.02 9

+6.1 +1.6 -0.4 -0.8 -2.1 -1.4 -1.7 31 1 o -1.4 ~1.4 I+1 n +? ' +4.o 8.50 26.04 (1.10 27.20 25.10 26.51 21.82 26.51 25.10'27.20 11.10 26.04 8.50 8.57 26.03 LO.97 26 Jo 24.62 26.12 21.49 26.00 24.91 26.67 10 *6 25 .9 2 8.77 10

+0.8 -0.1 -1.2 1.1 -1.9 -1.5 -1.5 -1.9 -0.8 -2.0 -2.2 -0.9 +1.4 5.98 10.08 25.13 11.16 27.20 11.11 27.17 11.11 27.20 11. 6 25.13 10.08 5.98 6.25 10.12 t5.26 11.04 26.78 10.87 11 26.87 10.78 26.61 11.01 25.34 10.17 6.35

+4.5 +0.4 +0.5 -1.1 -1.5 a ? 11 1n 9 9 -1.3 +0 A '+0.0 +4 '

6.81 9.90 25.13 11.10 25.05 10.91 25.05 11.7.0 25.13 9.90 6.81 7.19 LO.20 25.19 10.94 24.57 12 10.71 2'.70 .' O . 9 4 25.23 10.13 7.23

+5.6 +3.0 +0.2 -1.4 -1.0 -1 a -1_s -1.4 +0.4 42.3 +A.2 6.81 10.08 26.04 24.39 26.25 24.39 26.04 10.08 6.81 7.14 10.09 2 24.48 25.95 24.58 13 25.85 10.05 7.15

+4.9 +0.1I5.80 0.9 +0.4 -1.1 +0.8 -0.7 -0.3 +5.0 5.98 8.50 10.34 24.71 10.34 8.50 5.98 6.23 8.70 10.42 24.93 10.34 8.67 6.19

+4.2 +2.4 +0.7 +0.9 0.0 +2.0 +3.5 5.74 6.e4 5.74 a Vepco Model 5.93 7.09 5.93 15 b MEASURED +3.3 +1.6 +3.3 e 2 DIFTERENCE CORE AVERAGE BURNUP = 9,054 WD/MrU Vepco Model Vepco Model Vendor Model Measv ad Pe rdic ted Percent Dif fe rence Pe rcen t Difference Batch 1A 23.60 24.08 -2.0 Batch 2 Average 25.79 26.00 -0.8 absolute Batch 3A 24.14 24.24 Batch 4A 10.80

-0.4 value 2.3 10.91 -1.0 Batch 4B 8.01 7.84 4-20 +2.2

f Table 4-15 i

COMPARISON OF PREDICTED AND MEASURED CRITICAL BORON CONCENTRATION FOR VARIOUS CONTROL RODCONFIGURATIONS FOR CYCLE 10F SURRY UNITS 1 AND 2 Measured Vepco Model Predicted Vepco H odel Vendor Hodel Control Rod Critical Boron Critical Boron Percent Percent Unit Bank Position Concentration (PPM) Concentration (PPM) Difference Difference s s 1 ARO 1196 1168 -2.3 1 D-Bank In 1077 1050 -2.5 1

1 C and D-Banks In 957 942 -1,6 Average

c. absolute E

" value 2 ARO 1182 1168 -1.2 2.4 l

2 D-Bank In 1056 1050 -0.6 2 C and D-Banks In 947 942 -5.5 sr

Table 4-16 COMPARISON OF PREDICTED AND MEASURED CRITICAL BORON CONCENTRATION FOR VARIOUS CONTROL ROD CONFIGURATIONS FOR CYCLE 2 0F SURRY UNITS 1 AND 2 Measured Vepco Model Predicted Vepco Kodel Vendor Model Control Rod Critical Boron Critical Boron Percent Percent Units Bank Position Concentration (PPM) Concentration (PPM) Difference Difference sL 1 ARO 1033 997 -3.5 1 D-Bank In 917 899 -2.0 1 C and D-Banks In 800 787 -1.6 Average absolute value p 2 ARO 1408 1401 -0.5 0.7 is 2 D-Bank In 1325 1312 -1.0 2 C and D-Banks In 1208 1192 -1.3 S '

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Table 4-19 COMPARISON OF PREDICTED AND MEASURED INTEGRAL MANK WORTil FOR CYCLE 1 0F SURRY UNITS 1 AND 2 Control Rod Measured Vepco Model Vepco Model Vendor Model Bank Integral Predicted Integral Percent Percent '

Unit Position Bank Worth (PCH) Bank Worth (PCM) Difference Difference 1 D-Bank In 1480 1379 -6.8 A 1 C and D-Bank In 1330 1234 -5.1 Average absolute value 2 D-Banks In 1435 1379 -3.9 4.9 l

c 2 C and D-Banks In 1309 1234 -5.7 ,,

b .

Table 4-20 COMPARISON OF PREDICTED AND MEASURED INTEGRAL BANK WOR'HI FOR CYCLE 2 OF SURRY UNITS 1 AND 2 Control Rod Measured Vepco Model :o Model Vendor Model Back Integral Predicted Integral Unit rcent Percent Position Bank Worth (FCM[ Bank Worth (PCM) u* trence Di f ference 1 D-Bank In 1951 1079 2.7 1 C and D-Banks In 1331 1202 -9.7 Average ,

absolute value 2 D-Bank In 880 931 5.8 7.3 2 C and D-Banks In 1244 1249 0.4 ,

1 l

l l

l i

i FICURE 4.1 i

i SURRY UN.IT 1 - CYCI.E 1 4

, CRITICAL BORON CONCENTRATION i

vs.

BURNUP 3

J j 1100 j

1000 -- --- - -

l l i I '

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VEPCO H0 DEL PREDICTED CURVE hm ._.

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SECTION 5 - SLTfARY AND CONCLUSION The PDQC7 discrete model, which contains the NULIF, RAFIT, SHUFFLE, PAPDQ, and PDQ07 computer codes, is operational at Vepco for the purpose of performing detailed reactor physics analyse and supporting the evaluation of core performance. The accuracy of the PDQ07 discrete model h J been established through extensive compari-sons of calculations with measurements from the Surry reactors. The results of these compari.ons are:

)

1) Assembly average power distributions re predicted typi-cally within a standard deviation of 2%, with a maximum standard deviation of 4.3% for low power flux =aps (where uncertainties ia the data are greater due to the low neutron flux level and the drift in the power level) and a mnvimum standard deviation of 3.6% for flux maps at power levels greater than 10%.

2) PeakrodF{Hvaluesarepredictedtypicallywithin2.5%

with a =aximum deviation of 4.3% between predicted and measured.

3) Assembly average burnups are predicted typically within 2.5% and batch average burnups within 1.5%.

4) Critical soluble boran concentrations are predicted typi-cally within 30 ppm and boron worth within 3%.

~

5) C,ntrol rod bank integral worths are predicted typically tichin 6% with a maximum deviation of 9.7%.

In addition, the accuracy of the Vepco model was verified by comparison of the accuracy of Vepco results (i.e. comparison of predicted values with 5-1

1 measurement data) with the accuracy of similar results obtained from a vendor model. These comparisons in'icated that the standard deviation 1

and/or percentage difference between the Vepco PDQ07 discrete model calculations and reactor measurement data were within acceptable industry standards.

Verification, as well as improvements to the PDQ07 discrete

=odel, will continue to be made as more expecience is obtained throuri the continued application of the model to the Surry and North Anna reactors.

e 5-2 ii r ud

l SECIION 6 - REFERENCES

1. W. A. Wittkopf, et. al., NULIF "Neutrtts Spectrum Generator, Few Group Constant Calculator, and Fuel Depletion Code", BAW-10ll5, 5 June 1976.

2. H. H. Hassan, et. al. , " Babcock and Wilcox Version of PDQ07 - User's Manual", BAW-10117P, December 1975.

3. H. H. Hassan, et. al., " Shuffle - Program to Perform Fuel Shuffle in Nuclear Reactor Core", BAW-422, Rev. 1, July 1975.

4. H. H. Hassan, W. A. Wittkopf, et. al., "HAFIT", BAW-425, July 1973.

5. H. H. Hassan, et. al., " Zeus / General Service Programs", BAW-423, Rev. 1 July 1975.

6. Private correspondence from the labcock and Wilcox Company to the Virginia Electric and Power Company dated February 3,1971, and October 6, 1971.

7. Final Safety Analysis Report - Surry Power Station Units 1 and 2, Virginia Electric and Pover Company.

8., R. D. Klatt, W. D. Leggett III, and L. D. Eisenhart, " FOLLOW - a Code Pron. ding a Standard Reactivity Follow Procesure by Calculating Effective Critical Boron Concentrations as a Function of Burnup",

WCAP-7482, February 1970.

9. W. D. Leggett III, and L. D. Eisenhart, "The INCORE Code", WCAP-7149, Dece'mber 1967.

10. W. D. Leggett I1I, "T0iE - a Code for Totaling Local Burnup, Iso-topics, and Urani m Values", WCAP-7309, March 1969.

6-1

____-__--__-_-_c__-_---_-____---____,

~

i i

APPENDIX A Description of th2 INCORE, TOTE, and FOLLO'4

- Cc=puter Codes

l A-1 DiCORE COMPUTER CODE DESCRIPTION INCORE is a data analysis computer code used to process infor-

=ation obtained from the =ovable incore instrumentation syste=, and is therefuce, the pri=ary computer progra= for core follow analysis. Input to the DiCORE progra= consists of:

1) A description of reactor conditions when the =easure=ents were

=ade (such as power leval, control rod positions, etc.)

2) Incore detector readings including which flux thimbles were used and neutron cross sections of the sensor

3) Aaalytical infor=ation calculated by the PDQ07 discrete =odel

4) Options specifying which thi=bles will be e= ployed in local power predictions and what type of calculations are to be per-for ed INCORE corrects raw pointuise flux =easure=ents for leakage current, changes in power level between =easurements, and relative de-tector sensitivites to deter =ine the pointv'se reaction rate in the flux thi=bles. The =easured reaction rates deter =ined are then co= pared with expected values.

DiCORE co=putes the relative local power produced by each fuel asse=bly and in the peak fuel rod for each asse=bly. Local rela-tive power is computed as:

n P ={ I W (R x ci)}

j=1 R Cj 157 d is nor=alized so that I P /157=1 where P i=1 d A-1

1 l

and P = Measured power for the 1* location (which corresponds to an assembly for 1 < 1 < 157 or a fuel rod for i > 157)

R = Measured reaction rate for the j* th 6ble W = Weighting factor for the j thimble (W is 3 3 th based on the distance from the i location ch g to the j )

P C = Power calculated for the i U location by the PDQ07 discrete model R

= Reaction rate calculated for the j thimble by the PDQ07 discrete model n = Nu=ber of thimbles used fcr ceasuring power in the 1* location Different ratios of power to reaction rate (Pd/R ) obtained from the PDQ07 discrete model are used depending on the control rod configura-tions at each elevation.

INCORE calculates 1) the relative power for each assembly and quadrant based on the assembly average local powers and 2) the twenty largestvaluesofF[H in descending order with an identifying nu=ber so that hot spot (peak rod) locations in the core can be determined.

s A-2

A-2 TOTE COMPUTER CODE DESCRIPTION The TOTE computer code is an isotopic and burnup follow pro-gram which calculates material concentrations for the fuel and accumu-laced burnup based on measured power distributions (obtained from the INCORE coda) and tables of caterial concentrations as a function of burnup. The INCORE cade outputs burnup rate infor=ation for every fuel batch and/or assembly, (the total burnup rate for the fuel region is given as well as the value for each of four axial segments of approxi-

=ately equal length). The buraup rate is given as the megawatt-hours generated in a given fuel quantity per 1,000 megawatt-hours generated in the core.

Input to the TOTE code consists of:

1) the core energy (megawatt-hours) associated with each INCORE burnup rate deck

2) a description of each fuel region (including metric tons of uranium, corresponding INCORE source number, prwious burnup, isotopic depletion type, etc.)

3) tables of the change in up to ten caterial concentrations with burnup

4) the burnup rate decks from INCORE Core average, fuel batch and/or assembly, and material con-centrations are outputed by TOTE. Interpolation for =zterial concen-trations as a function of burnup is quadratic (generally using the two preceding and one succeeding table entry). A description of each fuel region (item (2) above) is output for subsequent TOTE runs.

A-3

- _ _ _ - - _ _ - - - - - - - - - _ - - - - - - - - - - - - - - . - - - - - - - - - - - - - - - - - - - - - - - - - - - ---------a

i i

A-3 FOLLOW PROGPAM DESCRIPTION The FOLLOW computer code processes reactor operation data and calculates nominal boren concentrations. The FOLLOW code is designed to describe the nearly linear relationship between available core reacti-vity and cycle burnup at nominal

• conditions. It is most convenient to use boron as a measure of core reactivity with off-nominal corrections being made for power, xenon and samarium, t ., ature, and control rods in ter=s of their boron worth. These off-nominal corrections are cade using the following equation:

off-nominal corrected or measured reactivity correction nominal boron = baron + inverse boron x due to rod group concentration concentration worth 1 position off-nominal off-nominal

+ ra :tivity correction + reactivity correction due to rod group due to moderator 2 positier temperature off-nominal off-nominal

+ reactivity correction + reactivity correction due to power due to xenon and sacarium behavior The boron concentration in Surry Units 1 and 2 is typically measured one to three times per day. After proper nor=alization, this data is plotted against cycle burnup and forms the " boron depletion curve". Since this curve is well behaved and nearly linear from the beginning to the end of the cycle, it can provide the following infor-mation:

1) Extrapolation to end-of-cycle life for verifying refueling dates.

2) Rate of loss of reactivity with burnup for confirmation of de-sign predictions.

A-4

3) Best-esti= ate of beginning cycle, hot-full-power criticality under equilibrium conditions.

• Nominal conditions are associated with hot full power, equilibrium xenon conditions, and all control rods withdrawn from the core.

A-5

C a

APPENDIX B PIPRESENTATIVE INCORE OUTPUT USING THE VEPCO PDQ07 DISCRETE MODEL 1

O l

FIGURE B-1 INCORE CALC"IATED ASSEMBLYWISE AVERAGE POWER DISTRIBUTION FOR INITIAL CORE AT BEGINNING OF LIFE CONDITION e

a a N m L K J M G F E O C 8 A

. Ps!CIC7?O . 0.66 . 0.87. 0.66 .

iE45U6ED~ .. ~ . Pe (OIC7.gg .

.PC7 018FE'ENCE.

. 0.72 . 0.95 . 0.'72 . . MEaiU8E0 . 1

. 9.2 . 9.5 . 9.2 . .PC7 O!PPE8ENCE.

__ ..stt.tr_stster**

...*** *.**.*.****.. ..**.*.**.*...***....*.=*... .****** -*******

. 0.62 . 0.98 . 1.10 . 1.00 . 1.10 0.94 . 0.42 .

. 0.62 1.02 . 1.12 . 0.94 . 1.12 . 1.02 0.62 . 2

. -1.1 . 4. 4 . 1. 7 . - 1. 7 . 1.7 4.4 . -1.1

. 0.63 . 0.94 . 1.05 . t.05 1 14 . 1.05 . 1.05 . 0.94 . 0.63 .

Ot6 5_,_9 9.5_.J . 0 3_.t,_1._04AL . 0,9 . 1.04 . 1.03 . 0.95 . 0.65 4.t . .

3

0. 7 . - t. 8 . - 0. 7 . - 4. 7 . -0 . 7 . - 1. 4 . O'T7 . 41.

..,0.6 3 . ,0. 7 7 . 0.9 9 .

. 1 02 . 1.15 . 1.08 . 1.15 . 1.02 . 0.99 . 0.77 . 0.63 .

. 0.65 3.9 .. 0.79 . 0.99 . 1.02 . 1.18 . L .0 7 . 1.18 . 1 02 . 0.99 . 0.79 . 0.e5 .

. 2.3 . -0. 3 . 0.1 . 2 . 5 . - 1. 5 . 2.5 . 0.1 . -0 3 . 2.3 . 3.9 4

________ . 0.62 ............................................................................................

0.94 0.99 . 1.00 1 14

. . . 1.10 . L.20 . 1.10 . L.t* . 1.00 . 0.99 0.9* . 0.62 .

. 0.41 . 0.96 . 1.02 . 1.02 . 1 12 . 1.09 1. t 7 . 1.09 . 1.12 . 1.02 t.02 . 0.96 . 0.61 .

. -1.5 . 1.4_.__3.3 . 2.1 . - 2. 3 . -0.7 . - 2.2 . -0. 7 . - 2. 3 . 21

. 3.3 5  ;

. -1.5 1.t0 . 1.21 . 1.t3 . 1.21 1.10 . 1.14 1 02 .1.4

. 0.95 . 1.05 . 1.02 . 1.14 .

. 1.05 0.94

._. 0.97 . 1. 0.5 _,,,,,1. Q 4

. -0 3 . -0.2 .

1.9 . -1.5 1.J3 . -13. 9.juj 6 18a._Ld1_, l.18Mt06 . 1.13 .

1.06a_1. 0 5 . 0.97

. 6

- 2. 8 . - 2. 2 . - 2. 8 . - 3. 6 L.9 . -0. 2 . - 0. 3 .

. 0.66 . 1.t0 . 1.05 . 1.15 . 1.10 . 1.21 . t.t4 . 1.24 t.14 1.21

-L.5 .

. 0.69 . ~ ~ . . . 1.10 . 1.15 . 1.05 . 1.to . 0.66 .

• .5 . 1.11 0.a .. 1.0 3.68 ' .~ t .12

-2.5 . - .~ 1 1.7 .0- 83.719 17~~ . - 1 1l .'137. - t .1t .2 . -1 3^.%1 13 . L.'T7

-3.9 -

T i.OsT Ut2'T U0a T t.1t .

0.8 0.69 . 7

.........................................................................1.7 . -2.5 . 3.6

. . . . 4.5 .

.- C . E 7 1.00 . 1.14 ................................

. . 1.08 . 1.20 . 1.13 . 1.24 . 1.16 . 1.24

. C.90 . 1.02 . 1.t5 1.11 1.22

. 1 13 . 1.20 . 1.08 . 1.1= . 1.00 . 0.~57

. 3.4 . 2.7 . 0.7 1.07 . 1.16

-0.9 . -3.3 1.15 . 1.22 . 1.11 . 1.16 . t .0 7 . 1.15 . 1.02 . 0.90 .

4 i

. -2.3 . -1.6 . -0.4 . -L.6 -2.3 -

-0.9 . 0.7 . 2.7 . 3.4 .

............. 1.10......................................................... 3.3 i

. 0.a6 .

. L.05 . 1.t3 . L.10 . 1.21 . 1.14 1.24 1.14 . . 1.21 . 1. '.0 1 15 1.05 . 1.10 . 0.66 t

._0.69.. t . t t . . _1. 0 a_._1.12_,._1. c a_,_J . t t_,J . n_._t . 23 . 1.t3 . 1.17 .

1.cs

. 4.5 . 0.a . 3.6 . - 2. 5 . - L . 7 . - 3 9 . -t.t

1. --t.t . -1 1 . -3.9 . -

1 12 . t.cs . 1 11 . 0.69 . 9 1

-2.5 . 3.6

..........................................................................t.7

. 0.9a . 1.05 . t.02 . t.t* . 1.10 . t.21 .

1.13 . 1.21 . 1 10 . t.14 0.s . *.5 .

. 0.97 1.05 . 1.04 . t.t3 . 1.06 . t.02 . t.05 . 0.9a . .

-0.3 . -0.2 . 1.9 . -1.5 . - 3. 6

. 1.16 . L.11 . 1.18 . 1.06 . 1.13 . 1.04 . 1.05 . 0.97

- 2. a . - 2.2 . -2.6 . -3.. . - . to

..................................................................t.5

. 0.62 . 0.94 . 0.99 . 1.G3 . 1.14 .

L.t0 . 1.20 . 1.10

~

7-[714

. s.9 . -0 2 . - 0. 3 .

. 0.61 . 0.96 . 1.02 . 1.02 . 1.12 . t.09 1.17 1.09

. L.00 . 0.99 . 0.94 . 0 61 .

. -1.5 . 14 . 3.3 . 2.1 .

-2.3 . -0.7 . - 2 2 . -0. 7 . -2.3 .

1.t2 . 1.02 . 1.02 . 0.96 . 0.6L . 11

. 0.63 0.77 . 0.99 L.02 1.15 2.1 . 3.3 .

1.4 . -1.5 . ,

. 0.65 . 0.79 . 0.99

. 1.02

. 1 18 1.08 1.07

. 1.15 1.18

. t.02 1.02

. 0.99 . 0.77 . 0.63 . {

j 1.9 2. 3 .

. . 0.99 . 0.79 . 0.65 . 12

. . -0. 3 . 0.1 . 2.5 . -1.5 . 2. 5 . 0.1 . -0.3 3 .9 .

l

. 0.63 . 0.94 1.05 . 1.05 . 1.14 t.05 . 1.05 . 0.94 . 0.63 .

2.3 .

. 0.65 . 0.95 . 1.03 . 1.~4 . 1.09 . 1.04 . 1.03 . 0.95 . 0.65 .

. 4.1 . 0.7 . - 1. 8 .

- 0. 7 . - 4 . 7 . = 0 . 7 . - . . 0.7 . 4.1 . 13

.................z._...................._..

. 0.42 0.98 . 1.10 . 1.03 . 1.10 . 0.94 . 0.62 22,....................

.. 0-1.162 . t.02

. 4. 4 .

. 1.12 . 0.98 . 1.12 . 1.02 . 0.62 .

1.7 . -1.7 14 1.7 . 4.4 . -t.1 .

. 57:N028 0 .................................................. ................

s ,.JEd!&7!CN

. . 0.66 . 0.87 . 0.66 . . Avfa:GE ,

. 0.027 .

0.7_2 . 0.95_2_1 72 . 807 O!F868ENCE. 15

. 9.2 . 9.5 . 9.2 . .

................ . 2.3 .

M/D Core Burnup Control Rod Unit Cycle Man Power (%) (MWD /MTU) Configuration 2 1 1 0 0 ARO B-1

FICRDRE B-2 INCORE CALCULATED ASSEMBLWISE AVERACE POWER DISTRIBUTION FOR INITIAL CORE AT BEGINNING OF LIFE CONDITION l

l I

A P N M L E J M G F E D C 5 A 3

__ .___P8ECIC'_(0 . .

Os48_z_0.5.6 . 0.48 . .

• s F 0_I C,7_E C .

. MftSutto . . 0.50 . 0.56 . 0.50 . . afa$uaE0 .

,_4

.PC7 O!F8?A ENCE. . 2.8 . 5.2 . 2.8 . .PC7 OIFFE8ENCE.

. . . u _.2.h.u.a u . . .a ,

. . m a. . . s s s . . . . e u . . . v_v e . te . . .Ar . . -

.av............

.0.75 106 . 0.90 . 0.40 . 0.90 . 1.04 . 0.75 . ,'

. 0.75 . 1.03 . 0.88 . 0.39 . 0.88 . 1.03 . 0.75 . 2 j

_,_. _ . - 0.0 _. - l . 4A -2.4

.. ..-2.1 . .. -2.4. ;-1.6 ._.0.0_.

. 0.81 . 1.16 . 1 18 . 1.04 . 1.02 . 1.04 . 1.18 . t.16 . 0.81 . .

._0. 8J .__J_.16__ JJ 4s t 00 . 0.96 . 1.00 . 1. t ?. _._1.16 . 0.83 . 3

. 2.7 . 0.1 . - 3. 8 . - 3. 8 . - 5. 6 . -3.8 . -3.8 . 0.1 . 2. 7 .

. 0.31 . 0.9 9 '. 1.19 . 1 15 . 1.24 . 1 15 . L.24 . 1.13 . 1.L9 . 0.99 . 0.81 . j

. 0.82 1.C2 . 1.20 . 1 15 . 1.20 . 1.11 . L.20 . 1.15 . 1.20 . 1.02 . 0.82 .

1.6 . 3.3 . 14 . - 0.1 . - 2. 9 . - 4. 0 . -2.9 . -0 1 .

1.4 . 3.3 . 1.6 .

........r....._,.,...nzn., .n f_s n e . s e1 = =. . z.n n

• 1 = u sr .a * = e . =r = = * * ==**=* *= = ****=.=***

. 0.75 . 1.16 . 1.19 . 1. ' . 1.0 7 . 1... 1.26 . 1.11 . 1.07 . 1.07 . 1.19 . 1.16 . 0.75 .

. 0.77 . 1.3% . 1.19 . 1.09 . 1.09 . 1.10 . 1.23 . 1.10 . 1.09 . 1.09 . 1.19 .

2.5 . 0.7 .

1.16 . 0.77 . 5

. 0,8 . 20. t.6 . -0.6 _-1 9 . -0.6 . t.6 . 2.0 . 0.8 . o.7 . 2.5 .

. 1.04 . 1.18 . 1.15 . 1.07 . 0.51 . 1.09 . 1.16 . 1.09 . 0.51 . 1.07 . 1.15 . 1 18 . 1.04 . l

. 1.07 . 1.16_._J .15 . 1.08 . 0 53 . 1 10 . 1 16 . 1.10 . 0.53 . 1.08 . 1 15 . 1.16 . 1.07 . 6 }'

. 2.2 . -1.6 . 0.4 . 11. 3. 6 . 0.5 . -0.6 . 0.5 . 3.6 . 11 . 0.4 . -1.6 . 2.2 .

. 0.68 . 0.93 . 1.d4 . 1.24 . 1.11 1.09 . 1.16 . 1.30 . 1 14 . 1.09 . 1.1L . 1 26 . 1.06 - 0.90 . 0.68 .

j t

. 0.51 . 0.92 . 1.07 . 1.25 . 1.12 1 09 . 1.L4 . L.31 . 1.14 . 1.09 . 1 12 . 1 25 . L.07 . 0.92 . 0.51 . 7 I

. 6. 3 . 2.7 . 2.5 . 0.5 . 1.1 . - 0. 5 . 0.6 . 0.2 . 0.4 . -0.5 . 11 . 0.5 . 2.5 . 2.7 . 6.6 .

...._.,.s......a..,s.z.......... m .1..x,__s._.........n . m ....................._.1..n......

. 0.54 . 0.40 . 1.02. .

L.15 . 1.26 . 1.16 . 1.30 . 1.25 . L.30 . 1.16 . 1 26 . 1.15 . 1.02 0.40 . 3754 .

. 0.57 . 0.41 . 1.03 . 1.16 . 1.24 . 1 15 . 1.30 . 1.26 . 1.30 . 1 15 . 1.26 . 1.16 . 1.03 0.41 . 0.57 . 8

. . 5.6 . 3 3., _t,4 . 0 4 ,, __-L.La - 0.) -0.2 . 0.3 . -0. 2 . -o. 6 . -16 . 0.8 . 1.6 . 33. 5.4 .

0.68 . 0.90 . 1.04 . 1.24 . 1.11 1.09 . 1.16 . 1.30 . 1 14 . 1.09 . 1.11 . 1 24 . 1.04 . 1.90 . 0.68 .

0.*1 . 0.92.._1.07s_1.25_.__1 12 . 1.09 . 1.16 . 1.31 1 16 . 1.09 . 1.12 . 1.25 . 1.07 . 0.92 .l .51 . 9

. 6.8 . 2.7 . 2.5 . 0.5 . 1.1 .-05. 0.6 . 0.2 . 0.+ . -0.5 . 11. 0.5 . 2.3 . 2. 7~ . 6.8 .

. 1. 0 6 _.,1.18 , .12 15 . 1.07 . 0.51 . 1.09 . 1.16 . 1.09 . 0.5L . 1.07 . 1.15 . 1.15 . L.04 .

. 1.07 . 1.16 . 1 15 . 1.08 4 53 . 1.10 . 1.16 . 1.10 . 0.53 . 1.08 . 1 15 . 1. 16 . 1.07 .

2.2 . -1.6 . 0.4 . 10

. 1.1 . 36 . 0.5 . -0.6 . 0.5 . 3.6 . 11. 0.4 . -1.6 . 2.2 .

___ ....................s,_.....t................................................................

. 0.75 . 1.16 . 1.19 . 1.07 . 1.a7 . 1 11 . 1.26 . L.11 . 1.07 . 1.07 . L.19 . 1.10 . 0.75 .

. 0.77 . 1.16 . 1.19 . 1.09 . 1.09 . 1 10 . 1.23 . 1.10 . 1.09 . 1.09 . 1 19 . 1.16 . 0.77 . 11

_ _ _ _ . . . 2 . 5_ . 0.7..__0.8 ._2,0, _1 6_.,-0.6 . = 2 .,9_._ -0_ . 6_f _ t . 6_.___2 . 0 . D r.j_. 0. __.l.9

. 0.81 . 0.99 . 1.19 . 1.15 . 1.26 . 1.15 . 1 26 . 1.15 . 1.19 . 0.99 . 0.81 1 . ......

_. . . _ _ _ _ . . 0. s 2 ._.t . 0 2., 1. 20_t. t .15 , . t . 2 0, . _1.11_._ t . 2 0_a__1.11_._J . to . 1,0_2 . 0. 8 2 . 12

. 1.6 . 3.1 . 1.4 . - 0 1 . - 2. 9 . -4.0 . -2.9 . -0.L . 1.6 . 3.3 . 1.6 .

. 0.01 . 1.16 . L .18 . 1.04 . 1.02 . 1.04 . 1.18 . 1.te . 0.81 .

. 0.83 . 1.16 . 1.1.' . 1.00 . 0.96 . 1.00 . L.14 . 1.16 . 071) .

. 2.7 . O .1 . - 3. 8 . - 3. 8 . - 5. 6 . - 3 . 8 . - 3. 8 . 0.1 . 2.7 .

LT-~~

_ _ _ _ _ _ _ _ _ . . _ _ s,....>..... re...** s.**........r. m...........................

. 0.75 . 1 04 _. 0.90 . 0.40 . 0 90 . 1.06 . 0.75 .

. 0.75 . 1.03 . 0.88 . 0.39 . 0.88 . 1.03 . 0.75 .

. -0 . 0 . - 1. 4 . -2.6 14

. 2 1 . -2.6 . -1.4 . -0.0 .

. 57 anca10 . . 0.48 . 0.54 . 0.48 . . AVERAGE .

. Ofv!17trm . .

0 10_. V.56 . 0.50 . .*C7 Of 78 E8 ENC!. LS

. =0.021 . . 2.8 . 5.2 . 2.8 . . . 1.9 .

M/D Core Burnup Control Rod Unit Cycle Map Power (*!) (E'D /EU) Confieuration 2 1 2 0 0 D-In B-2 n-

-i -

FIGURE B-3 1

l INCORE CALCULATED ASSEMBLYWISE AVERAGE PC'n'ER DISTRIBUTION FOR INITIAL CORE AT BEGINNING OF LIFE CONDITION 5 I

m A A  % M L E J M G F G 0 C S A I

. *AE0'.C7E3 . . 0.61 . 0.7* . 0.61 . . PREGIC7ED .

. m1AtuaEG . . G.61 . 4.30 . 0.61 . . MEAtuRED . 1 i

________ .FC7 G1781RINCE. . 0.9 . G.* . 0.9 . .pC7 DIS #EAENCE. - _. _.

. G.ow . G . 40 . 1.02 . 0. * . 1.C2 . 0.90 . 0.60 .

_. __ . C.a1 . 0.93 . 1.02 . 0. *? . 1,C2 0.93 . 0.61 . 2.

. J.1 . 3.a . -t.2 . -2.- . -0.2 . 3.6 . 3.1 .

..___ . . . . 0.e3 . v.*1 . 1.01 . 1 03 . 1.12 . 1.03 . 1.J1 . 0.01 . 0.A3 . , _ _ , _ _ _ . .

. Q.o* . U.*3 . 1.03 . ..C2 . 1.07 . 1.02 . 1.03 . 0.93 . Q.o6 . 3

. 1.7 . 1.- . 1.S . -1.1 . -.i . -1 1 . 1.5 . 1.6 . 1.7 .

_ _ _ , . _ . . 63 . 0.4w . 1.GC . 1.Ci . 1.17 . 1. 11 . 1.17 . 1.CS . 1.0C . Q.a0 . G.63 . . ..

. o.o1 . 0.6w . 1.ou . 1.J. . 1.1S . 1.o= . 1.1% . 1.06 . 1.00 . 4.ee . '.63 . 4

_ ___ . . G.0 . 4.4 . -G.3 . -0.3 . -1 1 . -1.7 . -1.1 . -0.3 . -0.3 . 0.3 . J.0 . ..___

. G.sG . v.41 . 10G . 1.w. . 1.la . 1.15 . 1.21 . 1.13 . 1.14 . 1.G6 . 1.00 . 0.91 . G.e0 .

. . . . ?.c4 . G.5G . 1.GG . 1.w* . 1 18 . 1.17 . 1.23 . 1.17 . 1.18 . 1.04 . 1.00 . 0.90 . 0.00 . .. _$_

. .1 . -G.9 . -G.= . -C.= . 0.1 . 2.2 . G.1 . 2.2 . c.1 . -0.6 . -0.6 . -G.9 . -0.1 .

... ._. 0.64 . 1.01 . 1.55 . ..it . 1.1m . 1.2m . 1.tv . 1.2 1.16 . 1.16 . 1.05 . 1. ' 1 . C.*C , _

. L.4C . ..C2 . 1.34 . 1.;4 . 1.17 . 1.26 . 1.1v . 1.26 . 1.17 . 1.15 . 1.44 . 1.02 . 0.90 . 6__.

. -a.1 . 0.1 . -4.* , -o.= . 1.1 . -0.2 . -G.= . -f.2 . 1.1 . -0.6 . -w.e . c.1 . -0.1 .

. G.a1 . 1.G; . 1.G3 . 1.11 . 1.1% . 1 26 . 1.21 . 1.29 . 1.21 . 1.26 . 1.15 . 1.17 . 1.03 . 1.02 . Q.41 .. ___ _ .

. G.s2 . 1.31 . 1.03 . 1.tw . 1.15 . 1.27 . 1.21 . 1 2v . 1.11 . 1.27 . 1.15 . 1. a o . 1.G3 . 1.01 . 0.62 . 7

. 1.6 . -1 3 . -0.3 . -0.5 ._ w.S . G.7 . -0.0 .

• 1. ? . -0.0 . C.7 . 0.5 . -0.5 . -0. 3 . -1.3 . 1.4 . _ _ _

. 0.74 . 6.45 . 1.11 . 1.11 . 1.23 . 1.1w . 1.29 . 1.22 . 1.29 . 1.19 . 1.23 . 1.11 . 1.12 . C . v5 . 0.79 .

. G.76 . G . *- . 1.1C . 1.1C . 1.2> . 1.14 . 1.29-. 1.22 . 1.29 . 1.19 . 1.2S . 1.10 . 1.10 . G . 4. . 0.19 .. ,__, $_.

. -t.. . -1.C . -2.2 . -0.e . 1.G . -C.4 . G.0 . ~.1 . 0.0 . -0.0 . 1.0 . -G.o . -2.2 . -1.0 . -0.1 .

. G.s1 . i.wt . 1.03 . 1.17 . 1.1! . 1.2s . 1.2% . 1 26 . 1 21 . 1.2m . 1.15 . 1 17 . 1.03 . 1.02 . 0.61 ,

. G.o2 . 1.C1 . 1.J3 . 1.14 . 1.15 . 1.27 . 1.21 . 1.26 . 1.21 . 1.27 . 1.15 . 1.le . 1.03 . 1 01 . 0,42 . 9___

. 1.4 . -1.3 . -G.3 . -G.S . u.S . 0.7 . -0.0 . -0.$ . -C.u . 0.7 . 0.5 . -0.S . -J.3 . - ? .3 . 16 .

. G.v; . 1.G1 . 1.GS . 1..G . 1.Le . 1.26 . 1.16 . 1.26 . 1.lo . 1.18 . 1.05 . 1.01 . 0.50 .

. 0.-C . 1.0 2 . 1. G. . 4.1G . 1.17 . 1.26 . 1.19 . 1.2o . 1.17 . 1.18 . . 04 . 1.02 . C.90 . 10

. _ _ , _ _ . . *.1 . c.1 . -0.= . -0.4 . 1.1 . -G.2 . -G.- . -0 2 . 1.1 . -C.4 . -0.% . C.1 . - 0 .1 . . _ , _____

. G.60 . V.91 . 1.GQ . 1.6 . 1.18 . 1.15 . 1.22 . 1.1$ . 1.16 . 1.Q* . 1.w0 . J.91 . 0.60 .

G.os . C.40 . 1.00 .,1.ue . 1.1* . 1.17 . 1.73 . 1.11 . 1.16 . 1.G* . 1.00 .

_ _ _ _ . 0.90 . G.60 . _ __ . 11_ .

. -0 1 . -3.4 . -C . = . -G.= . 0.1 . 2.2 . G.1 . 2.2 . 0.1 . -0.* . -G. . -G.9 . -0.1 .

, _ . . . . . . .63.

c 0.dG . 1.LC . 1.CS . 1.17 . 1.11 . 1.17 . 1.0$ . 1.00 . 0.AJ . 0.s3 . _.

. G.43 . G.tG . 1.GG . 1 . 54 . 1.15 . 1.Gv . 1.1S . 1.04 . 1 00 . C.40 . 0.e3 __. 12

. 0.0 . 0.3 . -G.3 . -G.3 . -1.1 . -1.7 . -1.1 . -0. 3 . -0. 3 . J.3 . G.C

. 0.e3 . G.41 . 1.01 . 1.w3 . 1.12 . 1.03 . 1.01 . 0.91 . G.a3 .

. 0.44 . 6.=3 . 1.03 . 1.02 1.07 . 1.02 1.03 . 0.93 . 0.6= . 13

_ _ _ _ _ _ _ _. . 1.7 . 1.6 . 1.5 . -1.1 . ..? . - 1.1 . 1.5 . 1.6 . 1.7 . . ___ __ _ ,,___

. s.6G . J.60 . 1.02 . d.*$ . 1.02 . 0.9G . 0.64 .

. _ _ _ . , _ . . w.e1 . 0.93 . 1.02 0.92 . 1.02 . 0.93 . 0.61 . _ . , _ 1* __

. L.1 . 3.S . -0.2 . -a.. . -0.2 . 3.4 . 3.1 .

. _ . 17u. atb. . . 0. 1 . 3.14 . J . 61 . . avfAAGE

. O!1!a7!CN . . 0.61 . G.aG . 0.61 . .p;7 Dip p r ethCt . , _ _ _ _ _ , _ _13 ,_

. ...G11 . . G., . a. . o.5 . . . G.* .

M/D Core Burnup Control Rod Unit Cvele. Mac Power C) (L'D/MTU) Confieuration 1

2 1 31 90 1,300 /0Ro B-3

FIGRE B-4 INCORE CALCULATED ASSEMBLYWISE AVERAGE PORER DISTRIBUTION FOR INITIAL CORE AT MIDDLE OF LIFE CCNDITIONS 1

1 R

• N- M L K J M G F E O C S A

. ikl01CTED .

m. ................

. 0 59 . 0.74 . 0 59 . *

. PRE 0!CTED .

. MEASURE: . . 0.58 . 0. 72 . 0.58 . . MEASUAE0 . I

___ . POT OiFFERENCE. . . . . . -1.6 . -2.5 . -1.4 . . .. ,_ ,___.P(f 01FFERENCR.

. 0.61 0.68 1.01 . 0.93 . 1.01 . 0.6& . 0.61 .

._ _ . 0.o0 . 0.14 . 0.99 . 0.91 . 0.99 . 0.84 . 0.60 . ,

. -2.6 . -4.5 . -2.6 . -2.2 .

. . . . _ , 2

-2. 6 . -4 5 . -2. 4 .

__ . 0.66 . 0.98 . 1.06 .

1.Os . 1.13 . 1.03 . 1.06 . 0.98 . 0.66 .....

. 0.65 . 0.97 . 1.05 .

1.03 . 1 10 . 1.03 . 1.05 . 0.97 . 0.65 . 3

. -1. 7 .

-1. 3 . - 1.2 . -0.4 . -2.4 . -0.4 . -1 2 . -1. 3 . -1.7

. 0.66 . 0.55 . 1.08 . 1.06 . 1 18 . 1.09 . 1.18 1.06 1.08 . 0.85 .. 0.66 .

. C.e5 . 0.80 . 1.08 . 1.08 . 1.38 . 1.09 . 1.18 . 1 06 1.06 . 0.86 . 0.65

,_.-13. 0.6 . 0.4 . 4

. 1,3 . C.1 . ~0 1. . . 0 1 ._ 1.3

. m.61 ..........................................................

. 0.96 . 1.06 . 1.06 . 1.19 . 1.01 1.21

. _ 0.4 . . 0.6.. -l.3 . .

. 1.11 . 1.19 . 1.06 . 1.08 . 0.98 . 0.61 .

_..___. . 0.61 . 0.97 . i.04 . 1.08 . 1. 20 _ . 1 12 . 1.21 . 1.12 . 1.20 . 1.06 . 1.08 . 0.97 . 0.61 . _ _ '

. -0.9 . - 1. * . Q.2 . 1.4 . 1.2 . 0.8 . 0.4 0. 8 . 12. 5

. 1.4 . 0.

. 0.88 . 1.06 . 1.06 . 1 19 . 1 11 . 1.22 . 1.12 . 1.22 . 1.11 . 1.19 .1.06 2 .. . -1.4 . - 0.9 .

_ , .__. . 0.67 . 1.0a .. 1.Co 1.20 1 13 1.23

. 1.06 . 0.8 5 . .

. - 1. * . -0 .1 . -0 . 0 .

0.6 1.3 .

. . 1.14 . 1.23 . 1.13 . 1.20 . 1.06 . 1.06 . 0.87 . 6

. 1.2 . 17. 1.2 . 1.3 . 0.6 . -0.0 1.01.........................................................................-0.1.-1.6

. 0.59 . . 1.03 . 1.14 1 11 . 1 22 ......................

. 0.55 . 1.00 . 1.04 . 1.19 1 13 1.23 1.13 . 1.22 . 1 13 . 1.22 . 1.11 . 1.18 . 1.03 . 1.01 . 0 59 .

. . 1.16 . 1.26 . 1.16 . 1.23 . 1.13 1.19 1.04 . 1 00 . 0 58 .

.. -0.7 . -1 1 ._ 0.5 . 0.8 . 1.5 . 0.8 . 2.7 . 25 . 2.7 . 0.8

. 7

............. 1.5 . 0.8 . 0.8 . -1.1 . -0.7 .

. 0.7. . 0.93 . 1.13 . 1.09 . 1.71 . 1 12 . 1.22 ....................................

_. G.71 . 0.93 . 1.13 . 1 10 . 121 1.15 .

1.13 . 1.22 . 1.12 . 1.21 . 1.09 . 1.13 . 0.93 . 0.7* .

1.25 . 1.le . 1.25 . 1.15.

. -1.6 . 0.2 . -0.0 . 0.6 . 0.1 1.21 . 1.10 . 1.13 . 0.73 . 0.73 . t 2.0 . 2.4 . 3.1 . 2.4 . 2.0 ..

......................................................,0.1

................... . 0. 6 . -0.0 . 0.2 . - 1.4 .

0.59 . 1. 01. . 1.03 . 1.k8 . 1.11 . 1.22 1 13 . 1.22.

1.13 . 1.22 ._1.11 . 1 18 . 1.03 . 1.01 . 0.59 .

.... 0.54 . .1.00 . 1.06 . 1 19 . 1 13 . 1.23 . 1 16 . 1.26 . 1 16 .

1.23 . 1.13 . 1.19 .

. -0.7 . -1.1 . 0.8 . 0.6 . 1.5 . 0.8 1.G* . 1.00 . G.58 . 9

. 2.7 . 2.5 . 2.7 . 0.8 . 1.5 . 0.8 . 0.8 -1.1 . -0.7 .

. O.a6 . 1.06 . 1.06 . ........................................................................

1.19 . 1.11 . 1.22 1.12 . 1.22 . 1.11 . 1.19 1.06

. s.67 . 1 06 . 1.06 . . . 1.06 . 0 . b4 .

1.20 . 1.13 1.23 . 1.14 . 1 23 . 1.13 . 1.20 1 04 1.06

. -1.6 . -0.1 . -0.0.. 0.6 . 1.3 . . . 0.87 . 10 1.2 . 17 . . 1.2 . 1.3 . 0.6 . -0.0 . -0.1 . -1.4 .

..............................s...............................

. G.61 . 0.96 . 1.04 . 1.06 . i.19 . 1.11 . 1.21 . 1.11 . 1.19 .. . . . . . . . . . . . . . . . . . . . . . . . . . . . .

._C. o 1 . 0. 9 7 . 1.08 . 1.06 . 1.06 . 1.26 . 0.98 . 0.61 .

. -0.9 . -1.4 . G.2 . 1.6 . 1.20 12. . 1.12 . 1.21 . 1.12 . 1.20 . 1.08 . 1.08 . 0.97 . 0.61 .

0.8 . 0.4 . 0.6 . _11

12. 1.6 0.2 . -1.4 . -0.9

. G.66 . 0 65 . 1.08 . 1.06 . 1.18 . 1 09 , 1.16 . 1.06 1.06 . 0.85 . 0.66 ._

. 0.65 . 0.46 . 1.08 . 1.08 . 1.18 . 1.09 .

. -1.3 . 0.6 . 0.4 . 1.0 . 0.1 . -0.1 . 1.14 01.

. 1.06 . 1.08 . 0.86 . 0.65 . 12 1.3 . 0.4 G

. 0. 6 6 . 0.96 1.06 6 . -1.3 0.65 . 0.97 . 1.05 . 1.03 . 1.13 . 1.03 . 1.06 . 0.98 . 0. 6 6 '.

1.03 . 1.10 . 1.03 . 1.05 . 0.97 . 0.65 .

_______. -1.7 . -1.3 . -1.2 . -0.4 . -2.* . -0.4 . -1.2 . 13

-1.3.. -1.7 .

. . . . . . .. .0.o1. . . .. . 0.88

. . . . .. .1.01

. 0.93 . 1.01 . 0.88 . 0,61 .

_. .. _ _ _ . _ _ _ . 0.6 0 . 0 . 64 . 0.99 . 0.91 . 0.99 0.84 . 0. 60 . __

. -2.4 . -4.5 . -2.8 . -2.2 . _ 14

. -2.8 . -4.5 . -2.4 .

. STANOALC .

. 0.59 . 0.74 . 0.59 . . AVERAGE

. OEVIATICN .

. 0.54 . 0.72 . 0.56 .

. .Q.015 . . -1.. . -2.5 . -1.4 .

. PCT 0!FFE4EN01'. 15

................ . . 1.2 .

_l M/D Core Burnup Control Rod Unit Cycle Mao Power (%)

(L'D/MTU) Confieuration 2 48 1 91 6,780 AJto B-4

FIGURE B-5 i

l INCORE CALCUIATED ASSEMBLWISE AVERAGE POWER DISTRII JTION FOR INITIAL CJRE AT END OF LIFE CONDITIONS a P N M L E J M G F t D C 5 A

.....t..........

. PAELICit3 . . 0.60 . L.76 . 0.60 .

. ME A$utt o .

. Pat 01C7tc .

. 0 5* . 0.77 . 0.59 .

PF A 5U410 fC7 LikftatMCt. __ . -1.6 . 3 4 . -: . * .

.FC7 0158tktNCE.

. 1 8

. L.66 . 0.bv . 1.03 . 0.*6 . 1.03 . 0.89 . 0.e6 .

._G.66..._0.LS . 1 03 . 0.** . 1.03 . 0.88 . 0.66 .

. 05 . -0 5 . -0.0 . -u.1 . ~0.0 . -0.5 . 0.S . .. ______.____2..

. 0. 6 t _3 0- . _1 10 . .1.06 . 1.1. . 1.C6 . 1.10 . 1.0* . c.aw .

. 0. 6 h '. 1.0. . 1.12 . 1.05 . . , , _ _ _ _ . _ . . . .

1.12 . 1.05 . 1 12 . 1.06 . 0.64 . 3

. -1.1 . G.3 . 1.6 . 1.1 . -2.1 . 11. 1.6 . 0.3 . -1 1 .

..t.,.................,.......................................................

. 0.69. . Q.E* . 1.12 . 1.07 . 1.17 . ". 0 7 . 1.17 1.07 . 1.12 . w.e* . 0.e6 . _ _ _ _ _ .

. 0.71 . 0.91 1.11 . 1.10 . 1.17 . 1.L6 . 1.17 1.10 1 11 . 0.61 0.71 . 6 3._7 . 2.2 . -0.9 . 3.L . _-9 1 . _- 1.1 . -0.1 . 3.5 . -0.9 . 22 .

3.7 . _ , _ _ _ _ _ _ _ _ _

. . . . . . .. .1. 06

.0.<* . . .. . .1 . 12 . . .. . 1.47

. . . . .. .1.16 . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ................. ............... _

1.03 . 1.ls . 1.04 . 1.12 .

t_b.e! . 1 0 $ _._,1 13_._J .4 6 . 1.16_.1.07 ._1.15. 1.07 1 16 . 1.07 1.06

. 1.12 . 1.06 . 0.66 .

. 1 13 1.C5 .

. t.! . 1.6 . 1.1 . 1.1 . -1.3 . -0.6 . -2 6 . -0.6 . -1.3 . 1.1 . 11 . 1.4 . 0.67 .. . , _ _ ,_ 5

.......................................................................................S.5

. .c.L* .

1.lb . 1.07 . 1 18..1.C8 - 1.17 1. C 7, . 1.17 1.08 1.15 . 1.47 1.10 O.t* . . . .

. C.b" . 1 11 . 1. 7 . 1.17 . 1.0f . 1.15 C.)

. 1.0S . 1.1S . 1.07 . 117 107 . . 11 . C.t* . 6

. . 1. 0 . 0.7 . -4.9 . -0.3 . -1.8 . -2 0 . - 1 8 . -0. 5 -0.* 0.7

...................................................................................... 1.0

. . . . 0.7 .

. 0.ou 1.04 . 1.17 . ' t .u t ...................

. 1.03 . . 1.17 . 1.07 . 1.1o . 1.07 . 1.17 . 1.08 . 1.17 . 1.06 . 1.C3 . 0.60 .

~ ~ ' - ' "

. 0.62 . 1.ww . 1.0% . 1 16 . 1.06 . 1.16 . 1.06 . 1. 15 . 1.06 1 16 2.*

. . 1.06 . 1.16 . 1.C* . 1.06 . 0.62 . 7

. . u.! . Q.6 . -2.6 . -1.S . -1.3 . -1.0 . -1.S . -1.0 . -1.3 . -1 5 . -2.* u.- . 26 .

............................................................................................. 0.5

. C.76 . 0.v6 1 16 ............

. . 1.07 . 1 18 . 1.u? . 1.16 . 1.06 . 1.16 . 1.07 . 1.le . 1.C7 . 1.1* . 3 . ** . 0.76

. C.76 . 0.sb . 1.13 . 1.CS . 1.1S . 1.06 . 1.15 . 1.05 . 1.15 . 1.06 . 1.15 . 1.G5 1.13 0.*5 . 0.76 .

. G.1 . 1.3 . -u.* . -1.7 . -2.1~.~-1 3 . -1.0 . -u.e . -1 0 . -1.3 . -2 1 . -1.7

~8 51.+

............................................................................................ 1.5 . 0.1 .

. 0.60 . 1.43 1.06 . 1.17 1.08 1.1' 1.07 ...............

1.C6

. . . . . 1.1. . 1.07 . 1.17 . 1 0b . 1.17 1.U. 1.03 . 0.60

. Q.t2 2.*

. . 1 06 . 1.1* . 1.EE".~1.16'. 1.wt . 1.15 . 1.06 . 1.16 . 1.06 .

1.16 . 1.G. .

1. D. . 0.62 .

~~ 9

. . 0.S . 0.6 . -2.6 . - 1 5 . - 1.3 . -1.0 . -15 . - 1. 0 . -1. 3 . - 1 5 . - 2 . * . 3.- 0.S 2.

. b.e* . 1 10 . 1.L7 . 1.La . 1.9h . 1.17 . 1.c7 . 1.17 . 1.06 . 1.13 . 1.07 . 1.10 . C.ev .

~~~ ~

. 0.L9 . 1 11 . 1.07 . 1.17 1.07 1 15

, 0 . T_ . 1.0 . 0.7,. -w.*

. . . 1.01 . 1.15 . 1.07 . 1.17 . 1.i1 . 1.11 . C.b9 . 10

. . -05 . -1.6 . -2.w . -1.6 . -0 5 . -0.9 v.7 1.0

...................................................................................... 0.7

. 0.6 1.06 .....

. . 1 12 . 1.u? . 1.lb . 1.08 . 1.ls . 1.08 . 1 16 . 1.07 . 1.12 1.0. . 0. m6

. C.o? . 1.0S . 1.13 . 1.CS . 1.16 . 1.07 1.11

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l FMURE B-9 INCORE CALCULATED ASSEMBLYWISE AVERAGE PCWER DISTRIBUTION FOR RELOAD CORE AT END OF LIFE CONDITIONS R P M M G K J H G F k 0 G b A

. PktalG7tG . . G.67 . u.61 . o.o? . . PRE Q1G1 Lu .

...m&AlckLb _ - . 4.69 . Q.s 1 . 0.e9 . - - - - MtASGRis 1

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- - - - - - .. - . G.e 4 . 0.55 . 1 14 . 0.49 . 1 1 = . 0.% $ , . G.6. --

. G.70 . 0.9S . 1.1$ . G.b4 . 1.1) . 0.95 . C.70 . 2

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. G.fo . 1.11 . 0.9. . 1.GS . u.99 . 1. GS . 0.94 . 1.11 . u.76 .

. G.7h . 1.13 . 0.9e . 1.G5 . 0.90 . 1.GS

• 0.96 . 1.13 . G.7b . 3

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. 3.1 . 1.$ . 1.o . 0.3 . -0.6 . -2 6 . -G.4 . G.3 . 1.6 . 1.5 . 3.1 .

. . Q.e4 ,-1.11-.-l.b u - 1.21-. C.vb.- 1.2 2 . 1.GG . 1.12. .-G.s v 1.2 6-1.G G- -l .11 4.6 *

. G.7C . 1.13 . 1.01 . 1.20 . 0.13 . 1.19 . G.49 . 1.19 . 0.9% . 1.20 . 1.01 1.13 . 4.70 . S

. L1. 1.1 . 0.6 . -6.7 . l a . -2.1 . -1.C . -1.1 . G.* . -G.7 . 0.6 . 1.1 . 2.3 .

. 0.9$ . C.v. . 1.21 . 0.99 . 1.05 . 0.96 . 1.12 . u.9h . 1.CS . 0.99 . 1.21 . 0.94 . 0.95 .

. 0.97 . 4.ve . 1.20 . 0.99 . 1.G3 . 0.94 . 1.G9 . 0.98 . 1.01 . 0.59 . 1.2G . 0.9e .=u.97 . 6

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. G.e7 . 1.1= . 1.6S . 1. G 1 . 1.2 2 . G.v6 . 1.' 1 . 1.G6 . 1.21 . Q.44 . 1.22 . 1.01 . 1.GS . 1.1= . 0.67 .

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. 2.1 . G.6 . G.2 . G.9 . -1.5 . -1.6 . -2.3 . -2.1 . -2.1 . -1.5 . -1.S . 0. 9 . G.2 . 4.4 . 2.1 .

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. Loat . G.41 . G.42 . 1.21 . 0.99 . 1.0v . 1.G6 . G.4* . 1.Ge . 1.0w . 0.9v . 1.21 . G.92 . G.91 . G.b1 . t

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. G . 6 a . 1.1 S . 1. Gs . 1.G2 . 1.24 . 0.97 . 1.18 . 1.Go . 1.16 . G.v7 . 1.20 . 1.G2 . 1.GS . 1.15 . 0.6b . 9 1.1-.-G .* . G. 2 . --o . v -. - 1. 3 , - 1. S-. -2 .3- . -2 .1 -2 . 3-.-.= 1. 5- -l .5 G. w- G. 2%. m -2,1

. G.95 . G.9* . 1.21 . G.99 . 1.GS - G.9h . 1.12 . 0.9h . 1.GS . 0.99 . 1.21 . G.94 . 4.9 5 .

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. G.7G . 1.1) . 1.G1 . 1.20 . 0.9v . 1.19 . o.99 . 1.19 . 0.99 . 1.2 w . 1. G 1 . 1.13 . 6.7 0 . 11

. 1.5 . 1.3 . 0.e . -0.7 . G.6 . -2.1 . -1.0 . -2.1 . G.e . -0.7 . 0.s . 1.3 . 2.5 .

. 0.76 . 1.06 . 1.C0 . 1.21 . 1.C1 . 1.20 . 1.G1 1.21 . 1.GG . 1.Gt . G.1 .

. G.79 . 1.10 . 1.02 . 1.22 . 1.01 . 1.17 . 1.61 1.2 2 . 1.02 . 1.10 . C.79 . 12

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. 0.7e . 1.11 . 0.94 . 1.0% . 0.90 . 1.GS . 0.96 . 1.11 . 0.76 .

G.7 4 l.13. Q .96 -. 1.f S . 0.90.. 1.Gb .-C.96 1.11-.-C.7? 13-.

. 2.7 1.v . 1.4 . 0.0 . -0.m . Q.G . 1.3 . 1.9 . 2.7 .

- _.G.64. 0.9S . 1.1* . o.69 . 1.16...G.9 6 -G.64

. 0.70 . 0.9$ . 1.15 . 0.49 . 1. l b . G.95 . 0.7G . 14

. 2.1 . -0.2 . G.S . 0.2 . G.S . -0.2 . 2.1 .

. . 57ANGAkG . ..G.67 . G.61 . 9.=7 . . A9kkAGk .

. OkVIA71Gu . . 0.69 . Q.a t . 0.e9 . .PG7 DIFFEAthGE. 15

....._.0.GtS 3.t -G.7 . . 3.1 - o M/D Core Burnup Control Rod Unit Cycle Map Power (*) (MWD /M'lU) Confieuration 2 2 23 co 8,850 ARO B-9

_ ___ . _ . _