ML20044A964

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Rev 0 to Comm Ed Topical Rept on Benchmark of PWR Nuclear Design Methods Using PHOENIX-P & Advanced Nodal Code (Anc) Computer Codes.
ML20044A964
Person / Time
Site: Byron, Braidwood  Constellation icon.png
Issue date: 07/31/1990
From: Lee D, Wise D, Yang S
COMMONWEALTH EDISON CO.
To:
Shared Package
ML20044A963 List:
References
NFSR-0081, NFSR-0081-R00, NFSR-81, NFSR-81-R, NUDOCS 9007170093
Download: ML20044A964 (56)
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COMMONWEALTH EDISON COMPANY TOPICAL REPORT ON BENCHMARK OF PWR NUCLEAR DESIGN METHODS USING THE PHOENIX-P AND ANC COMPUTER CODES i

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JULY,1990 Commonwealth Edison Company

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.' j I NFSR-0081-

'Rev.'O.

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[ COMMONWEALTH EDISON COMPANY TOPICAL REPORT j i

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BENCHMARK OF PWR NUCLEAR DESIGN METHODS USING THE PHOENIX-P AND ANC COMPUTER CODES j

j JLly, 1990 by D. K. Lee D. J. Wise S. T. Yang D. J. Hanner

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Prepared by: db 8'4.

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. Approved by: /[Ld2u tm // 4dt '; '

NuclearFuelServices{Mafager I

I Commonwealth Edison Company 72 W. Adams St., Room 900 Chicago, Illinois 60603 I '

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w I STATEMENT OF DISCLAIMER I This report was prepared by Commonwealth Edison Company

(" Edison") for filing with the United States Nuclear Regulatory Commission ("USNRC") for the sole purpose of obtaining approval I of Edison's PWR nuclear design methods. . Edison makes no warranty' or representation and assumes.no obligation, responsibility, or.

liability with respect to the contents of this report or-its .l j

accuracy or completeness. Any use or reliance on the report or I the information in this report is at the sole risk of the party using or relying on it.

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I NFSR-0081 Rev. 0 J

I ABSTRACT j

This Topical Report summarizes Commonwealth Edison

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'W Commpany's. Nuclear Design Methodology for its pressurized water reactors. Results of the benchmark program are presented to demonstrate Edison's capability to use the PHOENIX-P and.the I Advanced Nodal Code (ANC) computer codes to independently perform the nuclear analyses required for the fuel management, licensing, y

i operation,l testing,' and surveillance of a PWR reload cycle.

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TABLE OF CONTENTS J

CHAPTER PAGE (OF 55) l 3 TITLE PAGE 1 STATEMENT OF DISCLAIMER- 2

'I' ABSTRACT TABLE OF CONTENTS 3

4 l

1.0 INTRODUCTION

SUMMARY

AND CONCLUSIONS 9 1

.. 1.1 Introduction and Background 9 l

1.2 Overview 10 i 1.3 Summary and Conclusions 11

2.0 DESCRIPTION

OF METHODOLOGY 12 l -ca- 2.1 Introduction 12

'g 2.2 Basic Neutronic Computer Codes 2.3 Scope of Analyses performed 12 13 H 3.0 VERFICATION OF PWR NEUTRONIC METHODS 21 3.1- Introduction 21 3.2 -Comparisons to Station Measurements 21 3.2.1 Core Reactivity 22 3.2.2 Control and shutdown Rod Worth 23 3.2'.3 Isothermal Temperature Coefficient 24 3.2.4 Power Distribution 25 3.3 Comparison with Vendor Rt 26 y 3.3.1 Model Comparisons 26 L .

3.3.2 Delayed. Neutron Fraction 27 3.3.3 Rod Ejection Analyses 27 I .

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NFSR-0081 Rev. O.

TABLE OF CONTENTS (cont.)-

M PAGE (OF 55) 4.0 CALCULATIONAL UNCE.RTAINTIES 51

. . 4 .1. Justification For Use of Westinghouse Uncertainties 51

5.0 REFERENCES

52 APPENDICES Appendix As List of Acronyms 54

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NFSR-0081 Rev. O LIST OF TABLES TABLE PAGE'(OF 55) 2-1 Neutronic Computer Codes Currently Used by Edison 14 2-2 Neutronic Computer C0de Applications - 15 l Current Edison Codes  !

2-3 Neutronic. Computer Codes - PHOENIX-P/ANC 16 Design system 1

-. Wl. 2-4 Neutronic Computer Code Applications - 17 q PHOENIX-P/ANC Design System 2-5 Key Neutronic Parameters Required For 18 Safety Analyses 4

2-6 Neutonic Parameters Required for Operation, 19 Testing, and Surveillance  ;

3-1 PHOENIX-P/ANC Benchmark Scope 28 -

3-2 Heasured Versus Predicted Startup Critical 29- -l l Boron Concentrations - BY1C4 and 21C11 f 1

3-3 BY1C4 BOC HZP Measured Versus Predicted Rod 30 Worths Comparison

I 3-4 BY1C4 Measured Versus Predicted' Isothermal Temperature Coefficient 31

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. 3-5-Flux Map Comparisons for Power Distribution 32

Benchmark

'3-6 Average of Absolute Value of Measured Minus 33 Predicted Difference in Reaction Rate Integrals (RRIs) l Ll -W 3-7 Comparison With Westinghouse Results -

BR2C2 Boron Letdown 34 j

g 3-8 Comparison With Westinghouse Results - 35 g' BR2C2 Delayed Neutron Fraction l'"

3-9 Comparison With Westinghouse Results - 36 BR2C2 Rod Ejection Analyses 1

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I NFSR-0081 Rev. 0 T.TST OF FIGURES 6

FIGURE PAGE (OF 55)  !

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2 Calculational Sequence with the PHOENIX-P/ANC 20 l Design System-3-1 Z1C11 Critical Boron Concentration Versus Cycle 37 .

I Exposure HFP, ARO Equilibrium Xenon Conditions 3-2 ' Map BY10401 Normalized Reaction Rate Integrals I

38 and Measured Minus Predicted Differences in RRI '

3-3 Map BY10406 Normalized Reaction Rate Integrals- 39 and Measured Minus Predicted Differences in RRI 3-4 Map 1-B-06 Normalized Reaction Rate Integrals 40 and Measured Minus Predicted Differences in RRI 3-5 Map 1-B-17; Normalized Reaction Rate Integrals 41 and Measured Minus Predicted Differences in RRI _l e3- 3-6 Map'l-B-24 Normalized Reaction Rate Integrals 42 ,

and Measured Minus Predicted Differences in RRI 3-7 Core Average Axial Power Map BY10401 Versus 43 3D ANC Prediction l 3-8 Core Average Axial Powers Map BY10406 Versus 44 3D ANC Prediction 3-9 Core Average A:ial Power: Map 1-B-06 Versus- 45 3D ANC Prediction '

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7 3-10 Core Average Axial Power Map 1-B-17 Versus '46 s

S 3D'ANC Prediction 3-11 Core Average Axial Power Map 1-Bc24 Versus 47

'5 3D ANC Prediction 3-12' Comparison of Edison and Westinghouse Predicted 48 Average Assembly Power Distribution at HFP 5 (-0 MWD /MTU )

.3-13 Comparison of Edison and Westinghouse Predicted 49 Average Assembly Power Distribution at HFP

( 8,000 MWD /MTU )

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NFSR-0081' Rev. O I LIST OF FTG11RKs (cont. )

FIGURE -PAGE (OF 55) 3-14 Comparison of Edison and Westinghouse Predicted 50 8 Average Assembly Power Distribution at HFP

( 16,300 MWD /MTU )-

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1.0 INTRODUCTION

SUMMARY

AND CONCLUSIONS ,

1.1 Introduction and Background This report demonstrates Commonwealth Edison Company's (Edison's)

I ability to adapt-the Westinghouse advanced nodal methods into its nuclear design' methodology for its Pressurized Water Reactors (PWRs). I The results of this benchmark program are presented to demonstrate l I Edison 8s ability to use the PHOENIX-P and the ADVANCED NODAL CODE (ANC) computer codes to independently perform the neutronic analyses required for the fuel management, licensing, operation, I testing and surveillance of a PWR reload cycle.

In 1983 Edison submitted to-the NRC a topical report on the benchmark of its nuclear design methods for its PWRs (Reference 1).

The NRC reviewed and approved Edison's neutronic methodology for j performing nuclear design analyses for a reload cycle (Reference 2). :q Since 1983 Edison has1 completed sixteen (16) reload nuclear design

.8, analyses for its Zion, Byron and Braidwood nuclear plants. These analyses included loading pattern determinations, verification of La neutronic parameters and generation of nuclear data required for ig plant startups, operations and surveillances. Other aspects of the reload analyses, such as the thermal-hydraulic, LOCA, transient and accident evaluation were performed and will continue to be ,

I performed by Westinghouse. A topical report describing Edison's l

thermal-hydraulic and transient analysis methods was submitted to L the NRC for review in December 1989.

As indicated in Edison's previous reload analysis topical (Reference 1),

E the computer programs currently used by Edison are Westinghouse lg- Electric Corporation (Westinghouse) programs which have been approved lg by the NRC to perform nuclear design analyses. These include the ,

l FIGHT-H (FIGHT-H) fuel temperature program, the LEOPARD / CINDER (OD) lL cross section generation program, the PANDA (1D) and TURTLE (2D,3D) ll diffusion theory programs, and the PALADON (2N,3N) nodal theory '

.W program (References 3 to 9). The names in parentheses are the Edison names for the corresponding Westinghouse programs. Recently Westinghouse modified their design methodology to use transport theory cross >

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sections and an advanced nodal method. The computer codes used are the Westinghouse developed PHOENIX-P and ANC codes. The PHOENIX-P/ANC methodology has been approved by the NRC for use by Westinghouse for performing nuclear design analyses for PWRs (References 10 and 11).

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D NFSR-0081 Rev. 0-In '1989 Edison contracted with Westinghouse to obtain the PHOENIX-P and ANC' computer codes to incorporate into Edison's nuclear design 5 methodology. . These computer codes were installed on the Edison computer system, and a comprehensive. code verification program was conducted to confirm that the programs gave the same'results at Edisan-as at Westinghouse. To learn the PHOENIX-P/ANC based design methodology, I Edison sent an engineer to Westinghouse for a one year Design  ;

Participation Training Program (DPTP). This training included performance of the complete scope of a reload neutronics design using-I the PHOENIX-P/ANC design system. The DPTP augmented Edison's extensive reload design experience with specific. experience using PHOENIX-P and ANC. .i The current design process employed by Edison is based on the approach described in the Westinghouse Reload Safety Evaluation Methodology Topical Report WCAP-9272-P-A (Reference 12). This I approach will remain the same with the application of the PHOENIX-P/ANC design methodology. It is the purpose of this topical

=

report to. demonstrate that Edison can independently perform. nuclear design analyses utilizing the PHOENIX-P/ANC design system. This will be shown by comparing the results from Edison's benchmark program

  • to plant measurements and to the vendor's calculated resultr.

l.2 overview This report provides an overview of the' Edison nuclear design I process utilizing the PHOENIX-P/ANC design system. Specifically, Chapter 2 provides a description of the specific computer programs (g obtained from Westinghouse and the linkage programs developed in-g house,by Edison. Comparison to current design programs.will also be-made.

Chapter 3 addresses the benchmark of'certain key neutronic parameters. Most of the benchmark consists of comparisons of predictions to. plant data. For parameters which cannot be measured by the plant, comparisons are made on a one-to-one basis with I Westinghouse calculated results. Based on several discussions with the NRC (Reference 13), it was decided that predictions of core reactivity, control and shutdown rod worth, isothermal temperature

.I coefficient and core power distribution would be compared to actual plant measurements. Comparisons to Westinghouse results would be made  !.

for boron letdown, power distribution, delayed neutron fraction and i

,I rod ejection calculations. Finally, Chapter 4 provides a brief justification for the use of the same uncertainties that Westinghouse uses for their PHOENIX-P/ANC based design methodology.

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  • NFSR-0081 Rev. 0 I 1.3 Summary and Conclusions In summary, based on the comparisons with plant data presented in Section 3.2 of this report, Edison can-predict core reactivity,

.g rod worth, temperature coefficient and core' power distribution g within'the acceptance criteria. Also the comparisons with Westinghouse results, as presented in Section 3.3 of this report, show that Edison can develop core models and calculate neutronic parameters in a manner consistent with the Westinghouse calculations..

Based on these comparisons, it is concluded that Edison can independently use the PHOENIX-P/ANC computer programs to perform I' in an acceptable manner all neutronic analyses required for the fuel management, licensing, operation, testing and surveillance of'a FWR reload ~ cycle.. It is further concluded, based on the discussion in Section 4, that Edison can apply calculational uncertainties for 8 =. key neutronic parameters which are identical to those applied.-

by Westinghoure.

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2.0 DESCRIPTION

OF METHODOLOGY 2.1 Introduction This chapter presents a. discussion of the PHOENIX-P and ANC

! :omputer codes that Edison will use to perform PWR neutronic

. calculations. Functional comparison to the computer programs

~

which Edison is currently using to perform PWR reload analyses will be presented. = Though Edison plans to use the PHOENIX-P/ANC I design system, the basic design methodology remains the same as that which Edison has used in performing sixteen (16) cycles of reload neutronic analyses. This design methodology is the same as I the Westinghouse methodology, which is presented in WCAP 9272-P-A

-(Reference 12). .

i 2.2 Basic'Neutronic Computer Codes The PHOENIX-P and ANC computer codes were developed by Westinghouse.

The NRC has reviewed and approved the use of the PHOENIX-P/ANC 5' design system for PWR neutronic analyses (References 10 and 11).

In 1989, Edison obtained IBM versions of the PHOENIX-P and ANC l

computer codes and the required cross section libraries from j I- Westinghouse and installed them on its IBH system. A comprehensive j

-. code verification effort was performed to confirm that the Edison-l installed codes and libraries produced results consistent with those 1 l obtained at Westinghouse. To facilitate data manipulation, Edison L developed-the linkage program PHIX. PHIX is used to prepare the ..

L mechanical input data required to run PHOENIX-P. 'j As mentioned in Chapter 1. the computer codes Edison is currently j using to perform PWR nuclear design analyses are FIGHT-H,- 0D,1D, l

2D, 3D, 2N, and 3N. These are computer programs which Edison obtained I from Westinghouse in the early 1980s. A brief description of these  ;

codes and their application are presented in Tables 2-1 and 2-2 1 respectively. Edison plans to implement the advanced nodal methods ,

o into'its design methodology by replacing some of the current design  !

codes with the PHOENIX-P and ANC codes. Tables 2-3 and 2-4 provide a description of the PHOENIX-P/ANC codes and their application, and  :

l 3 -~ Figure 2-1 illustrates the design path for the generation of the l l3 PHOENIX-P/ANC neutronic models.

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NFSR-0081 1 Rev. O.

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2.'3 Scope o'f Analyses Performed Edison;is currently performing all the reload neutronic analyses ,

required for.the fuel management, licen11ng, operation, testing and ]

surveillance of its six PWR plants. The scope of these analyses -l include cross section generation, assembly loading pattern determination, lI- calculation of all key neutronic parameters required for the safety l analyses, and calculation of neutronic parameters necessary Sr l plant operation, testing, and surveillance.

The same methodology and core conditions-described in Edison's previous

nuclear design methodology topical report (Reference 1) will be used to "

calculate the key neutronic parameters. 'This methodology is identical to the current methodology employed by Westinghouse to perform neutronic

, analyses. A thorough description of the methodology is presented in Section 3 of WCAP-9272, " Westinghouse Reload Safety Evaluation Methodology,"

(Reference 12). For the single and multiple dropped rod events, the-  ;

neutronic analysis is performed according to the Westinghouse dropped rod methodology as described in WCAP-10297 (Reference 14).

A reload core can affect nuclear related key safety parameters in three areas core reactivity parameters and coefficients, control rod worth parameters, and key neutronic parameters required for specific events.

Table'2-5 lists the individual parameters which are calculated by Edison

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in each of these areas. In its determination of these parameters, Edison will apply the same calculational uncertainties as used by ,

Westinghouse.

For a reload design, Edison calculates a set of neutronic parameters l which are required for reactor operation, testing and surveillance.

This includes rod worth predictions using the Rod Swap methodology ,

(References 15 and 16) and the generation of constants used by the movable incore system for power distribution measurements. Table 2-6

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L provides a-list of the neutronic parameters which Edison calculates for each PWR reload cycle.

l l' l By changing to the PHOENIX-P/ANC computer codes, Edison will continue to perform neutronic analyses for the parameters presented in e Tables 2-5 and 2-6. Edison's capability to perform this scope of analyses using the PHOENIX-P/ANC computer codes is justified by the experimental benchmark results and the comparison to certain Westinghouse results presented in Chapter 3.

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  • TABLE 2-1 Neutronic Computer Codes Currently Used By Edison Westinghouse Edison-Code Name Code Name Description-FIGHT-H FIGHT-H Fuel Temperature calculation based on LASER 3 and REPAD4 results-LEOPARD 5/ CINDER 6 OD Macroscopic and microscopic few-group cross-sections. Fission product cross sections I' TURTLE 7 2D,3D Two and three-dimensional spatial few group diffusion theory ID One-dimensional (axial) few-group 8, PANDAS diffusion theory ,

i g; PALADONS 2N,3N Two and'three-dimensional nodal '

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-NFSR.0081 Rev. 0 -;

I TABLE 2-2 Neutronic Computer Code Applications Current Edison Codes '

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  • Code Name Application FIGHT-H Effective fuel temperatures a OD Hacroscopic and microscopic cross sections, kinetics parameters, isotopics 2D,.3D Power distribution, surveillance constants,

-g fuel depletion, xenon distribution, core g ,

reactivity coefficients, rod worths.

1D Axial power and xenon distributions, l I differential control rod worths 2N, 3N ~ Full core spatial calculations, scoping calculations '

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h -TABLE 2-3 Neutronic Computer Codes - PHOENIX-P/ANC_ Design System y i

Westinghouse Edison-I Code Name Code Name. Description ]

I FIGHT-H FIGHT-H Fuel Temperature calculation based on LASER and REPAD results 1 PHOENIX-P- PHOENIX-P Two-dimensional multigroup transport theory code used for lattice physics- 1 calculations '1 i

Multi-dimensional advanced nodal ADVANCED NODAL ANC CODE (ANC) theory

' PANDA 1D One-dimensional (axial) few-group  ;

diffusion theory I .

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TABLE 2-4 J

Neutronic Computer Code Applications -

PHOENIX-P/ANC Design System i

j Code Names Application FIGHT-H Effective fuel temperatures l PHOENIX-P Macroscopic and microscopic cross sections, .

kinetics parameters,.isotopics g ANC Power distribution, surveillance constants,- 1 Jg- fuel depletion, xenon distribution, core  ;

reactivity, reactivity coefficients, rod .

,g! worths, full core spatial calculations,

,5 Sc Pin9 calcu1* tion $

. , ID Axial power and xenon distributions, differential control rod worths 5

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I'- + NFSR-0081 Rev. 0 TABLE 2-s '

Key Neutronic Parameters-Required for Safety Analyses '

l. Core Re' activity Parameters and Coefficients ,

A. Moderator Temperature Coefficient "

8' B.

C.

D.

Fuel. temperature (Doppler) Coefficient Boron' Worth Delayed Neutron Fraction .

Prompt Neutron Lifetime I 2.

E.

Control Rod Worth Parameters 8 'A. Verification of Rod Insertion Limits (RIA) * - F B. Total Rod Worth C. Trip Reactivity (Minimum Value and Shape) 3 Differential Rod Worths D.  !

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3. Neutronic-Parameters for Specific Events i A. Loss of Coolant Accidents l-

- Total Core Peaking Factors From CAOC@ Analyses l

L. B. Boron Dilution (

l'g : C. Control Rod (RCCA) Events 3 -D. Control Rod Ejection L -E. Steamline Break L

  • Rod Insertion Allowance

' .. - @ Constant Axial Offset Control 1

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Rev. 0- .

i 31 TABLE 2-6

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Neutronic Parameters Required for Operation, Testing,  !

.. and Surveillance  !

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.a In addition to some of the parameters presented in Table 2-5,

<5 th' fo22owinS n'utronic Para =*ter8 '8 a function of CYCL

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are required for reactor operation, testing, and surveillance.  ;

I' 1. Radial Power Distributions

2. Boron Concentrations i
j. 3. Core Average Axial Power Distributions l3~ 4. Integral and Differential Rod Worths
5. Integral and Differential Xenon Worths t
6. Boron Worths s

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-Mechanical Data

-Material' Composition  !

I -Fuel Temper.ature PHOE MIX-P

-Cross Sections l

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-Data Transfer ANC I

-Spatial Calculations

. Figure 2-1 Calculational Sequence with the PHOENIX-P/ANC '

Design System Page.20 of-55 t

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I NFSR-0081 Rev. 0 3.0 yI2f FICA'rION OF PWR IdUrKUNIC MMHODS 3.1 Introduction this chapter presents the Edison neutronic calculational techniques I- a.id comparisons of calculated and measured or Westinghouse predicted results for some key core physics parameters. Specifically, 4

Section 3.2 compares Edison predicted parameters to station measucements while Section 3.3 compares Edison predicted parameters I to results of Nestinghouse analyses.

To verify Ec@ on's ability to calculate neutronic parameters required =i I for plant operacion, four parameters were selected for comparison to '

plant measurements. These parameters are core reactivity, control and shutdown rod worth, isothermal temperature coefficient (ITC) and power distribution. Based on several dicussions with the NRC (Refererce 13), it was decided that two reloads, Byron Unit 1 Cycle 4 and Zion Unit 1 Cycle 11, would be used for the benchmark. Table 3-1 provides tLe scope of the benchmark provided to the NRC in Reference 13.

I In addition, Edison predictions are compared to Westinghouse calculations for one reload cycle (Braidwood Unit 2 Cycle 2). Specifically, l3 comparisons are made for core reactivity (boron letdown), power distribution, delayed neutron fraction and rod ejection parameters.  ;

I l3 The parameters sel2cted for comparison to Westinghouse results are '

5 Provided in Table 3  !

l Based on the results presented in this chapter, Edison concludes that l it can independently perform neJtronic anal)ses using the PHOENIX-P/ANC l computer codes to provide results with acceptable agreement with ,

station measurements and Westinghouse predictions.

3.2 Comparisons to station Measu%ements l As mentioned in Section 3.1, credictions of core reactivity, rod worth, isothermal temperature coefficient and power distribution were compared to station measurements performed during startup physics testing, at power testing or surveillances. The measured data is from Zion Nuclear Station, Uait 1 Cycle 11 (Z1C11), and Byron Nuclear Station, Unit 1 Cycle 4 (BY1C4). Z1C11 was chosen because it used the Westinghouse 15x15 Optimized Fuel Assembly (OFA) 13 and full length Wet Annular Burnable Absorber (WABA). The Zion

'W Nuclear Station has used 0FA fuel and WABAs since Cycle 8 for both Units 1 and 2. BY1C4 was chosen because it is the first Edison reload to utilize the Westinghouse Vantage 5 fuel design, Features of the Vantage 5 fuel design such as Axial Blankets and part length Integral Fuel Burnable Absorbers (IFBAs) require nuclear analyses  !

to be performed in three dimensions. Edison prepared a licensing Page 21 of 55

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I I NFSR-0081 '

Rev. 0 I a' submittal and has obtained NRC approval to use vantage 5 fuel I for its Byron and Braidwood Nuclear Stations (References 17 I and 18). Edison also plans to use the Vantage 5 fuel design for the Zion Nuclear Station in future reloads. ,

3.2.1 Core Reactivity The figure of merit used for the core reactivity benchmark is the measured minus predicted soluble boron concentration.

5 Total core reactivity in a PWR is measured in soluble boron concentration and control rod position at a given reactor .

statepoint. As indicated in Table 3-1, comparisons were made i l_ for the startup critical boron concentration of BY1C4 and Z1C11 and for the boron concentrations as a function of 21C11.

burnup, i critical boron concentrat!,ons are measured by an acid base  ;

titration of a Reactor Coolant System (RCS) sample. During che BY1C4 and the 21C11 initial startup tests, the critical I boron was measured at both Hot Zero Power (HZP), All Rod Out (ARO) conditions and at HZP, reference bank inserted conditions. The reference bank is normally the bank with

'I the highest predicted worth when inserted in an unrodded core.

For BY1C4 and Z1C11, the reference banks are control bank B (CBB) and control bank D (CBD) respectively.

During the Z1C11 power operation, boron concentrations were '

measured often et core condittws other than at Hot Full Pcwer (HFP), ARO conditions. Tnese measured values were E adjusted to nominal core conditions (HFP, ARO) prior to  ;

comparison to the predicted values. The adjustment is made to account for off nominal power, temperature, control bank I position, and xenon distributions. However, measured data were excluded from the data base used for the comparison if an adjustment more negative than -50 ppm was necessary to r

I bring the data to nominal conditions. This data was excluded to minimize the error associated with adjusting from off-nominal to nominal conditions.

Critical boron concentrations were calculated at HZP and HFP conditions for different core burnups using ANC in the boron search mode. The calculations were performed using the 3D AMC models. No adjustment was made to the as I calculated boron values before comparing them to the measured data.

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Page 22 of 55

I NFSR-0081 Rev. O c

The predicted and measured startup critical boron concentrations for BY1C4 and Z1C11 are compared in Table 3-2. The absolute

_I value of the measured minus predicted differences are in the range of 2 ppm to 15 ppm which is well within the Technical specification Limit of 11%k/k (approximately 1100 ppm). The

'I' results are also well within the startup physics test design criteria of 150 ppm. .

_k Figure 3-1 shows the comparison between the neasured boron W. concentrations after adjustment to nominal conditions and the 3D ANC predicted boron concentration (solid line) for L C11. .

The Technical Specification limit of 11%Ak/k from the 3D ANC predicted values is also shown on the figure. All of the results fall within the Technical Specification limit.  :

3.2.2 control and shutdown Rod Worth This section presents the results of the integral rod worth benchmark. Measured values of the BY1C4 control and shutdown I rod worths, obtained from the BY1C4 startup physics tests, ,

are compared to values predicted by ANC calculations.

Individual rod bank worths are measured by. the rod swap ,

technique at the station. This technique involves measuring the worth of the highest worth bank (reference bank) by a boron ,

end point measurement. The_ remaining rod bank worths are 5

measured by swapping with the reference bank. This is done by inserting the bank being measured into the core in small,  ;

discrete steps and pulling the reference bank out of the I core by an amount just sufficient to offset the reactivity changes caused by the insertion of the bank being measured.

This is done until the bank being measured is fully inserted.

The reactivity change is determined from the reactivity computer.

.I The worth of ea:h bank is the sum of all the reactivity changes for that bank.

The BY1C4 individual rod bank worths were calculated using l- the rod swap methodology approved in References 15 and 16.

L Individual rod bank cross sections were calculated using the PHOENIX-P code. These cross sections were input into the ANC code which was used to perform the rod swap analysis.

A comparison of the BY1C4 calculated and measured rod bank I. worths is shown in Table 3-3. This table compares the worths of control banks D, C, B, and A and shutdown banks A, B, C, D, and E at HZP, BOC conditions. The figure of merit for the l I- comparison is the percentage differences between measured and predicted integral rod worths. Specifically, percentage I differences in individual bank worths and total band worth Page 23 of 55

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Rev. O l 1

are presentec. ine acceptance criteria for tne measured i versus predicted rod worth are 110% for both th9 reference i

bank.vorth and the total bank worth. For all individual banks other than the reference bank, the acceptance criteria is ,

115% or 1100 pcm, whichever is greater. As shown in Table 3-3, 1 the predicted rod worths all met the acceptance criteria.

3.2.3 Isothermal Temperature coefficient This section presents the results of the Isothermal Temperature Coefficient (ITC) benchmark. The measured value of the BY1C4

. ITC, obtained from the startup physics test, is compared to the predicted value.

The isothermal temperature coefficient is measured at HZP, I ARO conditions by ramping the average moderator temperature up 58F and then ramping back down to the initial equilibrium critical condition. During each temperature change, the reactivity change is measured on the plant reactivity computer.

The ITC is determined as the change in reactivity divided i by the change in temperature.

I The BY1C4 isothermal temperature coefficient was calculated using the ANC code. The parameters calculated were the Moderator Temperature Coefficient (MTC) and the Doppler-only Temperature Coefficient (DTC) at HZP, ARO conditions. The i MTC is defined as the change in the reactivity per unit change in moderator temperature, while the DTC is the change in reactivity per unit change in fuel temperature. The ITC, defined as the reactivity change per unit change in overall l reactor (both fuel and moderator) temperature, is obtained by summing the calculated values of MTC and DTC.

Table 3-4 presents a comparison of the calculated and the measured isothermal temperature coefficients at BOC, HZP and 3 ARO conditions for Byron Unit 1, Cycle 4. Since four different g temperature ramps were performed during the startup physics test, four ITC values were determined. The ITC reported i in Table 3-4 is an average of these four coefficients. The I

results show good agreement between the measured ITC and the  !

predicted value. In particular, the result is well within  !

the startup physics test design acceptance criteria of 13 pcm/8F.

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l NFSR-0081 Rev. 0 3.2.4 Power Distribution I This section presents the results for the power distribution benchmark. Measured values of normalized Reaction Rate Integrals (RRIs) for instrumented assemblies with operable

~g incore detector thimbles are compared to values predicted l 3 by ANC depletion calculations for selected full core flux maps. Also provided is a summary comparison between core average axial power distribution shapes obtained from incore I measurements and axial power predictions obtained from three dimensional ANC models.

l t

l Five full core flux maps were used in this benchmark. As t

indicated in Table 3-6, these include BOC, low power (30%)

and BOC, HTP flux mais for BY1C4,.and HFP flux maps at BOC, MOC,and 50C conditions for ZlC11.

The BY1C4 and Z1Cll power distributions were measured by the movable incore detector system. The system consists of 6 movable fission chamber neutron detectors '

-B. which are inserted into the bottom of the reactor vessel and driven up through instrument guide tubes in the fuel assemblies to the top of the core. Incore flux maps are obtained by taking voltage signal readings from L the detectors as they are slowly withdrawn from the top  !

L of the core. The detector signals are processed by the the Westinghose INCORE code to obtain information on core peaking factors, power distribution and power tilt.

The analytical input constants used in the INCORE analyses l in this report were obtained from ANC calculations.

The figure of merit used in this benchmark is the difference r i

between the measured and predicted normalized Reaction l Rate Integrals (RRIs) for instrumented fuel assemblies with operable incore detector thimbles. Since the mean value t of the difference between the notwalized measured and i

I' predicted RRI values is by definition zero for each flux map, the standard deviation of the measured minus predicted RRIs for each individual flux map is presented in Table 3-5. The standard deviation values in this data base ranged from

'Ie 1.8% to 3.5%, showing good overall agreement for the '

individual flux maps. Table 3-6 shows the average of L the absolute values of the measured minus predicted difference in RRIs for each individual map. Figures 3-2 to 3-6 show the radial distribution of the measured l normalized RRIs along with the actual difference between the measured and predicted RRIs for each flux map. The results show good agreement between measurement and prediction.

Page 25 of 55

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I i t3 l 5 NFSR 0081  !

Rev. O As part of the power distribution benchmark, the measured I core average axial power shape determined by INCORE was graphically compared to the average axial power shape calculated by the ANC code. The ANC calculations were performed in three dimensions using exposure and core conditions at which the maps were taken. Figures 3-7 to 3-11 show good agreement between the measured and predicted results for the five flux maps. The periodic

,l, depressions observed in.the measured data are due to the 5 assembly grids, which are not explicitly modeled by the ANC code. I 3.3 Comparison With Vendor Results [

To supplement the experimental benchmark results presented in Section 3.2, comparisons were also made between Edison and Westinghouse rebolts for some key neutronic parameters j calculated for.the Braidwood Unit 2, Cycle 2 (BR2C2) reload core. The parameters selected for comparison are:

1. Hodel comparison - boron letdown, power distribution
2. Delayed neutron fraction; '
3. Ejected rod worth and total core peaking factor for the IL rod ejection accident.

3.3.1 Model Comparisons The BR2C2 model independently developed by Edison was compared to the Westinghouse BR2C2 model. Both models were developed using the THOENIX-P/ANC code system. ,

The parameters selected for comparison are the critical I. boron concentration and the average assembly power distribution. t

~

.[W Table 3-7 shows the results of Edison and Westinghouse predicted HFP, ARO critical boron concentrations at different I core exposures. The differences observed are in the range  ;

of 2 to 8 ppm. Figures 3-12 to 3-14 show the Edison and I Westinghouse calculated average assembly power distributions at BOC, MOC, and EOC and their percent differences. The differences are all less than 1% for assemblies with a relative power greater than 1.0. None of the differences in any

. assembly are greater than 1.2%. The differences in boron concentration and power distribution are of the magnitude expected from two independently developed models.

p eage s o, ss y

l I

NFSR-0001 Rev. 0 3.3.2 Delayed Neutron Traction Table 3-8 provides the comparison of Edison and Westinghouse delayed neutron fraction analysis results. Included are the predictions for the maximum and the minimum delayed neutron fractions at BOC, HZP conditions. Results shown that Edison calculated values are within 1% of the Westinghouse values.

This is well within the acceptance criteria of Reference 1.

3.3.3 Rod Ejection Analyses Table 3-9 compares the results of Edison and Westinghouse I rod ejection analyses. The parameters compared are ejected rod worths and total core peaking factors calculated at BOC and EOC at both HFP and HZP core conditions. Results I show good agreement between Edison calculated values and Westinghouse predictions. The percent differences are all within the Reference 1 acceptance criteries I

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NFSR-0001 Rev. 0 '

I TABLE 3-1 PHOENIX-P/ANC Benchmark Scope I. Comparison to Station Measurenments A. Core Reactivity I 1. Byron 1 Cycle 4 Startup critical Boron

2. Zion 1 Cycle 11 Lifetime Predictions I B. Control Rod Worths
1. Byron 1 Cycle 4 Rod Swap Data l

l C. Isothermal Temperature coefficient

1. Byron 1 Cycle 4 MTC, DTC and ITC Data D. Power Distribution i.

I 1. Byron 1 Cycle 4 BOC HZP and BOC HTP Flux Maps

2. ion 1 Cycle 11 BOC, MOC and EOC Flux Maps i

II. Comparison to Selected Vendor Results A. Model Comparisons

1. Braidwood 2 Cycle 2 Boron Letdown
2. Braidwood 2 Cycle 2 Power Distributions B. Neutronic Input to 3PIL Calculations I 1. Braidwood 2 Cycle 2 Delayed Neutron Fraction
2. Braidwood 2 Cycle 2 Rod Ejection Calculations III. PHOENIX-P/ANC Code Uncertainties A. Justification for Use of Westinghouse Uncertainties I

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NFSR-0081 Rev. O i TABLE 3-2 '

Measured Versus Predicted Startup Critical Boron Concentrations - l" BY1C4 and ElC11' Critical Boron (ppm) Measured minus Predicted Unit Heasured Predicted Difference (ppm)

  • i I

BY1C4 I - HZP ARO

- HZP CBC in 1206 1107 1213 1105 2

21C11 t

- HZP ARO 1438 1453 -15

- HZP CBD in 1319 1324 -5  :

r I

  • Technical Specification Limit is 11% k/k (approximately 1100 ppm)  !

Startup Physics Test Design Acceptance Criteria is 150 ppm ,

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NFSR-0081 i Rev. 0 I TABLE 3-3 i

BY1C4 BOC HZP Measured Versus Predicted Rod Worths Comparison j

I Bank Measured Worth (pem)

Predicted Worth (pem)

Percent difference Measured vs. Predicted

((M-P)/P)

Acceptance Criteria CBC 882 930 - 5.2% 110%

CBD 504 495 1.8% 115%

CBB 657 630 4.3% 115%

I CBA 339 386 -12.2% 115%

SBE 445 495 -10.1% 115%

SBB 800 804 - 0.5% 115%

SBA 306 282 8.5% 115%

SBC 393 404 - 2.7% 115%

l 1 lW SBD 379 404 - 6.2% 115%

Total bank 4705 4830 - 2.6% 110%

t I CBx

  • Control Bank x, SBx = Shutdown Bank x Reference bank is CBC '

I Hensured worths reported are averaged values from boration and dilution measurements.

  • Acceptance Criteria is 110% for reference bank and total bank worth, and the greater of il5% or 1100 pem for individual bank worth.

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I NFSR-0081 Rev. 0

]'

I TABLE 3-4

)

BY1C4 Measured Versus Predicted Isothermal Temperature Coefficient (BOC ARO HZP conditions) -

I Measured Predicted . Meas-Pred

ITC (pem/eF)

ITC (pcm/8F)

-3.05 -3.95 I Cooldown #1 Heatup #1 -3.59 -3.95 0.90 0.36 i

Cooldown #2 -4.04 -3.95 -0.09 Heatup #2 -3.29 -3.95 0.66 Average -3.49 -3.95 0.46 I

  • Startup Physics Test Design Acceptance Criteria is 13 pcm/8F  ;;

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____o__________._____. _ _ _ _ _ _ . _ _ _ _ _ _ . . _ _ _ . _ _ _ , ___. _ -. .__ __ ._ ,

'I I NFSR-0081 Rev. O I TABLE 3-5 Flux Map Comparisons for Power Distribution Benchmark Cycle Thermal Control Number of Standard **

Burnup Power (% Bank D operable Deviation Unit Map (MWD /MTU) of rated) (steps)* thimbles (%)

BY1C4 BY10401 0 29.2 166 46 3.5 BY1C4 BY10406 1177 99.6 224 51 1.8 Z1C11 1-B-06 165 89.7 193 57 3.0 Z1C11 1-B-17 7681 99.1 228 48 2.3 21C11 1-B-24 14391 99.8 228 48 3.0

  • Fully withdrawn is 228 steps
    • Mean is equal to zero (see Section 3.2.4)

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I TABLE 3-6 Average of Absolute Value of Measured Minus Predicted Difference in Reaction Rate Integrals (RRIs)

Cycle f Burnup lM-Pl '

I Unit Map (MWD /MTU)

Difference (%)

BY1C4 BY10401 0 2.8 BY1C4 BY10406 1177 1.8 1 21C11 3-B-06 165 2.5 ,

21C11 1-B-17 7681 1.9 21C11 1-B-24 14391 2.6 4

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I NFSR-0081 Rev. 0 i

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TABLE 3-7  :

Comparison With Westinghouse Results - BR2C2 Boron Letdown (HFP ARO)

I Edison Westinghouse Edison-Westinghouse  !

Exposure Prediction Prediction Difference (MWD /MTU) (ppm) (ppm) (ppm) 0 1081 1083 -2 150 793 795 -2 1000 794 795 -1 ,

2000 811 812 -1 3000 810 811 -1 4000 794 796 -2 6000 722 726 -4 ,

8000 611 617 -6 10000 482 490 -8 12000 339 347 -8 14000 183 191 8 -

16000 26 32 -4 16300 3 8 -5 I

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I I NFSR-0081 Rev. O TABLE 3-8 Comparison With Westinghouse Results - BR2C2 Delayed Neutron Fraction (BOC,HZP,NOXE conditions)

Percentage Acceptance I Edison Prediction Westinghouse Prediction Difference (W-t)/W Criteria (Ref. 1)

Delayed Neutron 0.005606 0.005614 0.14% 13%

Fraction (minimun)

I Delayed Neutron Fraction (maximun) 0.006363 0.006361 -0.03% 13%

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I NF5R-0081 Rev. O I TABLE J-9 Comparison With Westinghouse Results - BR2C2 Rod Ejection Analyses I Parameter Edison value Westinghouse value Percenatge Difference (W-E)/W Acceptance Criteria (Ref. 1)

HFP-BOC

a. Ejected Rod Worth 0.178% p 0.178% @ 0% 15%

I b. Total core Pecking factor (FQ) 4.08 4.04 -1.0% 14%

a. Ejected Rod Worth 0.219% $ 0.218% @ -0.5% 15%
b. Total core '

Peaking factor (FQ) 4.74 4.76 0.4% 14%

I.-

HZP-BOC

a. Bjected Rod Worth 0.244%A0 0.245%y 0.4% 5% <
b. Total core Peaking factor (FQ) 6.76 6.87 1.6% 14%

I HZP-EOC

a. Ejected Rod Worth 0.704%p -2,3% 15%

0.720%@

g b. Total core

. ,3 Peaking factor (FQ) 16.37 16.34 -0.2% i4%

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9. . .0459. . . . . . . . . .0077. . . . .0006.

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Figure 3-2 Map BY10401 Normalized Reaction Rate Integrals and i Measured Minus Predicted Differences in RRI I

I rage 38 of 55

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Figure 3-3 Map BY10406 Normalized Reaction Rate Integrals and Measured Minus Predicted Differences in RRI l'aga 39 of 55

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4 4

Figure 3-6 Map 1-B-24 Mormalized Reaction Rate Integrals and fleasured Minus Predicted Differences in RRI Pa're 42 of 55 .

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Page 45 of 55

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i NFSR-0081 l

Rev. 0 ,

I rigure 3-12 Comparison of Edison and Westinghouse Predicted  !

Average Assembly Power Distribution at HFP (0 }ND/MTU)

.I 1 2 3 4 5 6 0.961 0.972 1

-1.1 1.050 1.227

'5 2 1.057

-0.7 1.230

-0.2 l

1.305 1.301 1.067  ;

8 -3 1.309

-0.3 1.301 0.0 1.075

-0.7 1.204 1.191 1.162 1.217 5 4 1.214 1.197 1.174 1.223

-0.5 l

-0.8 -0.5 -1.0 l 1.358 1.041 1.018 1.159 0.984 1 5 1.357 1.048 1.026 1.161 0.986 0.1 -0.7 -0.8 -0.2 -

0.2 -

1.278 1.065 1.208 1.238 1.122 0.911 i

6 1.277 1.067 1.205 1.234 1.115 0.903

!~ 0.1 -0.2 0.2 0.3 0.6 0.9 1.206 1.126 1.146 1.008 0.704 0.420 7 1.203 1.122 1.136 0.997 0.696 0.415  :

0.2 0.4 0.9 1.1 1.1 1.2 i 0.605 0.549 0.513 0.330 8 0.604 0.551 0.511 0.329 0.2 -0.4 0.4 0.3 I I  :

LEGEND

'~

West.

Edison

% Diff.

% Diff = ((Wast. -

Edison)/ West.)

  • 100%

I Page 48 of 55 I

9 =

, W m -,----_v

i ,

l' NFSR-0081-

.Rev. 0 'F

I Figure 3-13 Comparison of Edison and Westinghouse-Predicted

,g Average Assembly Power Distribution at HFP (8,000 MWD /MTU)

1. W s 1 '2 3 4 5 6 0.840' I l 1. 0.849

- l '. 0 -

"gL 0.935 1.310 t

'5 2 0.942- 1.303

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3 1.142- 1.383 0.981 {

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{E 4 1.011-

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-0.9 0.2 -0.7 0.5 h5 5.

1.221.

1.221 0.869 0.878 0'898 0.906 1.259 1.252 1.001 1.008 O.0- -1.0 -0.9 0.6: -0.7  ;.

1.0381 0.932 1.273- 1.164- 1.317 1.089 6 1.043- 0.940 1.265- 1.162 1.316 1.087

-0.5 -0.9 0.6 0.2 0.1 0.2 c

1.214 1.246 1.267 1.118 0.741 0.468 J-7 1.220 1.249 1.258 1.108 0.735 0.465

-0.5 -0.2 0.7 0.9 0.8 0.6

.0.602 0.571 0.545 0.364 8 0.605 0.575 0.546 0.364

.-0.5 -0.7 -0.2 0.0

, i LEGEND West.

Edison

,;g. ' "

% Diff.

1 ig)

'Y  % Diff = ((West. - Edison)/ West.)

  • 100%

jl L . Page 49 of 55

I l

NFSR-0081-f

-Rev. 0' Ltl .

Figure 3-14. Comparison of Edison and Westinghouse Predicted. ,

, Average Assembly Power Distribution at HFP (16,300 MWD /HTU).

1 2' 3 4 5 6 0.875 i l' O.879

-0.5 0.944 1.314 5 2 0.944 0.0 1.305 0.7 3

lg 1.097 1.312- 0.983 L g; 3 1.093 1.310 0.982 0,4- 0.2 0.1 1.012 1.258 1.033 1.353-

' (g .'

4 1.011 1.250 -1.032 1.345 0.1- 0.6- 0.1 0.6 1.207 0.908 0.946 1.326 1.002 j.1 5 1.202 0.910 0.949 1.320 1.004 l - -- '.4 -0.2 -0.3 0. 5 .- -0.2 1~ ~

! I1.013 0.939 1.299 1.115 1.244~ 1.044 lk I -

i 1.015. 0.944 1.296 1.117 1.248 1.048-

-0 -0.5 0.2 -0.2 -0.3- -0.4 1.198 1.208 1.079 0.732, 0.481 S e l'.203 1.007 1.077 0.732 0.481 i

-0.4 0.1 0.2 0.0 0.0  ;

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l , .t' O.597 0.572 0.390 l -b6 7 3 I

l LEGEND West.

[t

~ '

Edison 1  ;  % Diff.

% Diff = ((West. - Edison)/ West.)

  • 100%

I 1:

1 I

Page 50 of 55 I- .

1 I

^

i 1

1 8 NFSR-0081' Rev. O'

't 4.0-CALCULATIONAL UNCERTAINTIES L4.1 Justification For Use of Westinghouse Uncertainties the HICENIX-P/ANC code system has been approved-for use in the i l ecre design process by the NRC (References 10 and 11). The  !

previous secticaL cf this report have shown that Edison can

! L use the PHOENIX-P/ ant code system to obtain results which meet 1 Technical Specification limits and startup physics test design-acceptance criteria,and achieve the same level of accuracy as results calculated by Westinghouse with the PHOENIX-P/ANC 7 code system.; These results, therefore, demonstrate that-Edison =

~

can apply the same key neutronic parameter uncertainties-that are  ;

l:. .

applied in-the Westinghouse reload design process'(Reference'12).

Lli vs  ;

i

.f 3

BT i LI l

LI ll Page 51 of 55

, i I I '

8 -NFSR-0081-Rev. 0 5'.0 REFERENCES

^

L p 1 1. " Commonwealth Edison Company Topical Report on Benchmark of' PWR Nuclear Design Methods", NFSR-0016,' H. Cenko et al. , July 1983.

"I 2. -"NRC SER_on Ceco Topical Report on Benchmark of PWR Nuclear Design Methods", NRC letter from S.A. Varga to D.L. Farrar, December 2, 1983. .

~3. " LASER - A Depletion Program for Lattice Calculations Based on .

HUFT and THERMOS", WCAP-6073, C.G. Poncelet, et al. , April, .1966. -[

4. "The Doppler Effect for a Non-Uniform. Temperature Distribution- I in Reactor Fuel Elements," WCAP-2048, J.E. Olhoeft, July, 1962.

II  :

S. " LEOPARD - A Spectrum Dependent Non-Spatial Depletion Code for

.the LBM-7094", WCAP-3269-26, R.F. Barry, September, 1963.

'i

6. " CINDER - A One-point Depletion and Fission Product Program",

WAPD-TM-334, T.R. England, August, 1962. ,

7. "The TURTLE 24.0 Diffusion Depletion Code", WCAP-7213-P-A
(Proprietary) and WCAP-7758-A, S. Altomare and R.F. Barry, January, 1975.
8. "The PANDA Code", WCAP-7048-P-A (Proprietary) and WCAP-7757-A,

-R.F. Barry, et al., January, 1975.

9. "PALADON - Westinghouse Nodal Computer Code", WCAP-9485 (Proprietary) and WCAP-9486, T.M. Camden, . et al. , December,1978.

l l 10. " Acceptance for Referencing of Licensing Topical Report -

WCAP 10965-P and WCAP 10966-NP", NRC letter from C. Berlinger,to E.P. Rahe,-June 23, 1986.

11. " Acceptance for Referencing of the Westinghouse Topical Report WCAP-11596, " Qualification of the PHOENIX-P/ANC Nuclear
gi Design System for Pressurized Water Reactor Cores" ",

NRC letter from A.C. Thadani to W.J. Johnson, May 17, 1988.

12. " Westinghouse Reload Safety Evaluation Methodology",

WCAP-9272-P-A, S.L. Davidson, ed., revision dated July, 1985.

$ 13. "Braidwood Nuclear Power Station Unit 1 - Transition in Neutronic Codes for PWR Reload Design", Commonwealth Edison letter,

]\ g J.A. Silady to T.E. Hurley, February 21, 1990.

.. , i Page 52 of 55

{ .{ .

o

, 4 NFSR-0081 g

Rev. 0

5.0 REFERENCES

(CON'T) c

14. " Dropped Rod Methodology for Negative ~ Flux Rate Trip Plants",

j! .

WCAP.10297-P-A, T. Morita, et al., June 1983, i ..  :

h --

15. "NRC SER on CECO Topical Report on Control Rod Worth Measurement

, Using the Rod Swap Technique 'for Zion Station Units 1 and 2",- ,

NRC letter from S.A.' Varga to D.L. Farrar, December 1983.

16. "NRC SER on CECO Topical Report on Control Rod Worth Measurement Using the. Rod Swap' Technique for Byron Units 1 and 2 and ,

Braidwood Units 1 and 2", NRC. letter from V.S. Noonan to

,!. D.L.-Farrar, September 1986.

.{ 17. " Byron Stations Units 1 and 2 Application for Amendment

y to Facility Operating Licenses NPF-37 and NPF-66",

Commonwealth Edison letter, R.A. Chrzanowski to T.E. Murley, July 31, 1989.

18. "Braidwood Stations Units 1 and 2 Application for Amendment

.. .to Facility Operating Licenses NPF-72 and NPF-77",

i= Commonwealth Edison letter, S.C. Hunsader to T.E. Murley,

. July 31, 1989.

L8; g: .

I L

ih,

l; lli il eeee e3 o, ,,

s .

-t B NFSR-0081'  !

Rev..O.  ;

I APPENDIX A l List of Acronyms

.i W 'ANC Advanced Nodal Code

, ARO All' Rods Out BOC Beginning of Cycle BR2C2 Braidwood Unit 2 Cycle 2 i BY1C4 Byron Unit 1 Cycle 4 ,

5y '

CBA Control Bank A CBB Control Bank B jI CBC CBD Control Bank C Control Bank D if 4

.g' DPTP Design Participation Training Program '

g. ,

DTC Doppler-only Temperature Coefficient i EOC End of Cycle

,I 8F EQXE

Degree Fahrenheit Equilibrium Xenon FQ Total Core Peaking Factor (also termed total nuclear-

^I: .; hot channel factor) ,

1 e HFP Hot Full Power- -!

HZP Hot Zero Power ,

IBM International. Business Machines 'i IFBA Integral Fuel Burnable Absorber .

B ITC Isothermal Temperature coefficient Loss of Coolant Accident

I.

[

LOCA' Measured 1

l M MOC Middle of Cycle MTC Moderator Temperature Coefficient .

MWD /MTU Megawatt (thermal) Days per Metric Tonne of Uranium  ;

ga N0XE No Xenon J5! HRC Huclear Regulatory Commission l' 0FA Optimized Fuel Assembly J

I Page 54 of 55

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4 * ,

l t

i t- NFSR-0081 Rev. 0

, APPENDIX A (CONT.)-

List of Acronyms pcm percent mille (Reactivity x 105) ppm _ parts per million, boron concentration PWR Pressurized Water Reactor RIA Rod Insertion Allowance RRI Reaction Rate Integral i RCCA Rod Cluster Control Assembly SBA Shutdown, Bank A g~ SBB Shutdown Bank B

,B. SBC Shutdown Bank C SBD Shutdown-Bank D SBE' Shutdown Bank E WABA' Wet Annular Burnable Absorber WCAP- Westinghouse Commerical Atomic Power Z1C11 ' Zion Unit 1 Cycle 11 ,

.g AK/K . Reactivity Difference Reactivity Difference

'g l

B I;

.I I

.g

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