ML20010B624

VEPCO Flame Model

ML20010B624
Person / Time
Site:	Surry, North Anna
Issue date:	07/31/1981
From:	Beck W, Bowling M, Matthew Smith VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
Shared Package
ML18130A343	List: ML18130A342 ML20010B610 ML20010B617 ML20010B624
References
VEP-FRD-24A, NUDOCS 8108170352
Download: ML20010B624 (111)

Text

{{#Wiki_filter:_ VEP-FRD-24A JULY, 19 81 4 Vepco THE VEPCO FLAME MODEL DO QD w 4er r. jaa8F8=2aa89lg FUEL RESOURCES DEPARTMENT VIRGINIA ELECTRIC AND POWER COMPANY t

VEP-FRD-26 A 9 THE VEPC0 FLAME MODEL BY I W. C. BECK NUCLEAR FUEL ENGINEERING GROUP FUEL RESOURCES DEPARTMENT 7 VIRCINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 4 JULY, 1981 ( 1 Recommended for Approvel: M. L. Smith Nuclear Fuel Engineer Approved: /Y.%*$ E. S M. L. Bowling, Director Nuclear Fuel Engineering

f ** %,,%' % [ UNITED STATES E \\* ~ d/ NUCLEAR REGULATORY COMMISSION $.7, JO,Hjj,3 f '4 WASHINGTON, D. C. 20S55 o s, f g o NAY 13 Nai Pr. W. N. Thomas, Vice President Fuel Resources Virginia Electric Power Company Richmond. Virginia 23261

SUBJECT:

ACCEPTANCE FOR REFERENCING OF TOPICAL REPORT VEF FRD-24 "THE VEPC0 FLAME MODEL" The Nuclear Regulatory Commission (f!RC) staff has completed its review of the Virginia Electric and Power Company Topical Report number VEP.-7RD entitled "The VEPC0 FLAPE Model". This VEPCO developed model is a ' three dimensional, one energy group, modified diffusion theory calculational model. The model is used to perform three dimensiortal reactor physics analyses in support of reactor startup and cycle operations of the Surry and North Anna nuclear reactors. A summary of our evaluation of the licensing topical report is attached. 'As a result of our review,~ we find that Licensing Topical Report VEP-FRD-24 entitled "The VEPC0 FLAME itodel" dated October 1978 as augmented by responses to NRC questions submitted in your November 1978 letter is acceptable for referencing in licensing applications 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 safcty features described. in tic report and its amendment as found acceptable herein. Our accept- ,ance applies only to the use of features described in the topical report and !ts amendment as discussed in the attached safety evaluation. In accordance with established requirements, it is reqcested that Virginia Electric and Power Company issue a revised version of this report within three months of the receipt of this letter. The' revised version is'to appropriately incorporate the information subnitted in your October,1978 letter. This ev.11uation letter and the attached safety evaluation 'is to be included in the revised vusion between the title page and the abstract and the approved report will carry the identifier VEP-FRD-24A.

Mr. W. N. Thomas MAY 13 UC' Should Nuclear Regulatory Commission criteria or regulations change such that our conclusions as to the acceptability of the report are invalidated, Virginia Electric and Power Company will be expected to revise and resubmit the topical report or submit justification for the continued effective applic-ability of tha topical report without revision. If you have any questions about the review or our conclusion, please contact us. Sincerely, E' !.c o u t Robert L. Tedesco, Assistant Director for Licensing Division of Licensing

Enclosure:

As stated N a n

MY 7 1531 REVIEW OF TOPICAL REPORT VEP-FRD-24, "VEPC0 FLAME MODEL" Report Number: VEP-FRD-24 Report

Title:

The VEPC0 FLAME Model Report Date: October,1979 Originating Organization: Virginia Electric and Power Company Reviewed By: Walter L. Brooks / Core Performance Branch Introdtetion Virginia Electric and Power Company has submitted licensing topical report VEP-FRD-24 entitled "The VEPC0 FLAME Model." FLAME is three-dimensional nodal analysis code which is capable of calculating: 1. Assembly and core average axial power distributions 2. Differential control rod bank worths 3. Integral control rod bank worths vs. bank positon 4. Control rod bank insertion limits 5. Axial burnup distribution 6. Axial offset 7. Peaking factors (F, Fxy(z), F ) However, the code is currently used to calculate only differential and integral control rod worths and core average axial peaking factors on a production basis. Our review of this r epu '. follows. Summary of Report This report presents a summary description of the FLAME code and severci service codes which provide input to or handle output from the code. In addition the report includes descriptions of the cores of the_Surry Nuclear Power Station, Units 1 and 2 and fuel loadings for the first four cycles of each reactor. Nor-malization of the FLAME results to higher order (e.g., PD007 discrete model) calculations is described. Axial and radial albedoes are chosen to provide i

_2 agreement at boundaries. The migration area values are adjustea for rodded nodes to provide agreement at interior nodes. Total rod bank worchs calculated by the FLAME code are normalized to those calculated by DD007 in order to determine the differential and integral bank worths as a function cf insertion. Extensive comparisons are presented of radial power distrib.utions, axial power distributions and differential and integral rod worths for both Surry reactors for the first four cycles. The radial power distribution comparisons are between FLAME and PD007 calculations. The other comparisons arr between FLAME calculations and experiments. The results of these cesserisons have been surmarized to obtain the following uncertainties for the parameters calculated by the code on a preduction basis: Core average axial peaking factor 8% Differential Bank worth 2.0 PCM/ step Integral Bsnk Worth Individual 15% Cumulative 10% These uncertainties are the result of comparisons with experiment and represent the calculation-experiment differences. In effect the measurement uncertainty has been assuned to be nil. This is a conservative evaluation of calculational uncertainty. Summary of Evaluation We have reviewed the summary descriptions of the FLAME code and its associated service codes. The FLAME code and the input preparation codes, NULIF, PDQ07, and FLAFIT were purchased by VEPC0 from Babcock and Wilcox, who have submitted

' topical reports -describina nethods used in the design of Babcock and Wilcox reactors. These methods should be equally applicable to the Westinghouse supplied reactors for whirn VEPC0 proposes to use them provided they are suitably modelled. We have reviewed the data in VEP-FRD-24 which is presented to support the conclusion that this code is suitable for VEPC0 reactors. These data show that the VEPC0 FLAME model (which includes normalization to the PD007 discrete model) provides the capability to predict axial peaking factors and axial power distributions and :lifferential and integral rod worths. This conclusion is based on the review of the comparisons with measured data presented in the report. VEPCO has analyzed the data to obtain calculational uncertainties for axial peaking factors and control rod worths. The analysis was perforned in a conservative manner and the results are consistent with industry wide values. We conclude that they are acceptable. Evaluation Procedure The review of topical report VEP-FRD-24 has been conducted within the guidelines provided by the Standard Review Plan, Section 4.3. Enough information is provided, directly and by reference, to permit a knowledgeable person to conclude that the methods described are state-of-the-art. Regulatory Position Based on our review we conclude that the VEPC0 FLAME code is a suitable method for calculating axial power distributions and differential and integral rod bank worths when the code is suitably normalized to PD007 calculations. In addition certain other paraneters which depend on these calculations, e.g., axial .,,,,.,...i

1 I burnup distributions, rod: insertion limits, axial offsets, etc., may also be obtained fromi the. calculated results. We further conclude -that the topical report VEP-FRD-24 may be used as a reference to the de:cription of the code and its cuitability for use in performing these calculations. In addition, the report may..be. useo 'as a reference to ' support the use of the uncertainty values given therein for axial peaking factors and rod worths. i t-I + i l 4 n 4 I ,..,. ~,. _ -,,

F CLASSIFICATICN/ DISCLAIMER The data and analytical techniques described in this report have been prepared specifically for application by the Virginia Electric and Pcwer C:rpany. The Virginia Electric and Power Company makes no claim as tc the accuracy of the data or techniques contained in this report if used by other organisations. Any use of this report or any part thereof must have the prior written approval of the Virginia Electric and Power Cc=pany. i 4 1 1

A2STRACT The Virginia Electric and Power Company (VEPCO) has developed a three-dimensional (x, y, z), one energy group, modified diffusion theory. calculational model designated as the Vepco FLAME Model. The model utilizes the NULIF, PDQG7, FLAFII, and FLA'IE3 computer Codes which are part of the Fuel Utilization and Ferformance Analysis Code (FUPAC) system obtained from the Babcock and Wilcox Company. In addition, the EDITQAR, PICCOLO, and FLMSHUFL codes have been written by Vepco for use in the FLAME Model. The model is used to perform three-dimensional-reactor physics analysis in support of reactor startup and cycle operation-of the Vepco Surry and North Anna nuclear reactors. The accuracy of the FLAME model is demonstr.tted through comparisons with measurements taken at Surry Units No.1 and 2. 11

ACKNOWLEDGEMENTS - The author would like to thank Messrs. C. B. Franklin, J. i R. Rodes and M. L. Smith for their technical assistance during the development of the FLAME model and to Ms. Cathy Langston, 1s. Ivy Wilkerson, and Ms. Miranda Cooper for their typing of the draf t and final l manuscripts. The author would also like to express his appreciation-E to a number of people who reviewe.' end provided comments on this report. d J t I iii

~~- t TABLE OF CONTENTS Page CLASSIFICATION /DISCLALMER. 1 ABSTRACT ii ACKNOJLEDGEMENTS iii i TABLE OF CONTENTS. iv LIST OF FIGURES. v 4 LIST OF TABLES ix SSCTION 1 - INTRODUCTION 1-1 SECTION 2 - CORE DESCRIPTION 2-1 2.1 Introduction. 2-1 2.2 Core Design 2-1

2. 3 Fuel Loadings 2-2 SECTION 3 - MODEL DESCRIPTION.

3-1 3.1 Introduction. 3-1 3.2 Input Preparation 3-2 3.3 Thermal-Hydraulic Feedback Paraceters 3-15 3. <. Xenon Concentration Calculation 3-16 SECTION 4-CALCULATIONAL TECHNIQUES. 4-1 4.1 Power Distribution i Normalization 4-l' l 4.2 Differential and Integral Control Rod Worths as a Function of Bank l Position. 4-3 i SECTION 5 - RESULTS. 5-1 5.1 Introduction. 5-1 5.2 Radial Power Distribution 5-1 5.3 Axial Power Distribution. 5-2 5.4 Differential and Integrar Red !!orths. 3-3 SECTION 6 -

SUMMARY

AND CONCLUSIONS. 6-1 SECTION 7 - RFTERENCES 7-1 iv

LIST OF FIGURES Figure Title Page No. 2-1 Cross Sectional View of Surry Fuel Assembly. . 2-7 2-2 Control Rod Bank Locations 2-8 2-3 Surry Units 1 and 2 - Cycle 1 Fuel Loading 2 -9. 2-4 Surry Unit 1 - Cycle 2 Fuel Leading. 2-10 2-6 Surry Unit 2 - Cycle 2 Juel Loading. . 2-11 2-6 Surry Unit 1 - Cycle 3 Fuel Loading. . 2-12 2-7 Surry Unit 2 - Cycle 3 Fuel Lading 2-l? r 2-8 Surry Unit 1 - Cycle 4 Fuel Loading. . 2-14 2-9 Surry Unit 2 - Cycle 4 Fuel Loading. 2-15 2-10 Surry Units 1 ano 2 - Cycle 1 Burnable Poison Rod Loading. . 2-16 2-11 Surry Unit 1 - Cycle 2 Burnable Poison Rod Loading. 2-17 2-12 Surry. Unit 2 - Cycle 2 Burnable Poison Rod Leading. ..-2-18 l 2-13 Surry Unit 1 - Cycle 3 Burnable Poison Rod Loading 2-19 2-14 Surry Unit 2 - Cycle 3 Eurnabla Poison Rod Loading . 2-20 l 2-15 Surry Unit 1 - Cycle 4 Burnable Poison Rod Loading . 2-21 2-16 Surry Unit 2 - Cycle 4 Burnable Poison Rod Loading . 2-22 3-1 Flow Chart for Ver;o FLAME model . 3-3' l l l 5-1 Radial Power Distribution Comparison for Surry 1, Cy:le 1, HZP, ARD, BOC 3-6 i 5-2 Radial Power Distribution Comparison for Surry 1, Cycle 1, HFP, ARO, BOC 5-7 5-3 Radial Power Distribution Comparison for Surry 1, Cycle 1, EFP, ARO, MOC 5-8 5-4 Radial Power Dist?tibution Comparison tor Surry 1, Cycle 1, HFP, ARO, EOC 5-9 v

LIST OF FIGURES (Continued) Figure Title Page No. 5-5 Radial Power Distribution Comparison for Surry 1, Cycle 4, HZP, ARO, BOC. 5-10 5-6 Radial Power Distribution Comparison for Surry 1, Cycle 4, EFP, AR0, BOC. 5-11 5-7 Racial Power Distribution Comparison for Surry 1, Cycle 4, RFP, ARO, MOC.. 5-12 5-8 Radial Power Distrits 'on Comparison for Surry 1, Cycle 4, HFP, ARO, EOC. 5-13 5-9 Core Average Axial Power Distribution Comparison for Surry 1, Cycle 1, BOC, HZP, ARO 5-14 2 10 Core Average Axial Power Distribution Comparison for Surry 1, Cycle 1, BOC, HZP, o Bank in 5-15 5-11 Assembly B-8 Axial Power Distribution Comparison for Surry 1, Cycle 1, BOC, HZP, D Bank -In 5-16 3-12 Core Average Axial Power Distribution Comparison for Surry 2, Cycle 3, BOC, HZP, ARO.. 5-17 5-13 Core Average Axial Power Distribution Comparison for Surry 2, Cycle 3, BOC, HZP, D Bank in 5-18 5-14 Core Average Axial Power Distribution Comparison for Surry 2, Cycle 3, 30C. HFP, ARO 5-19 5-15 Core Average Axial Power Distribution Comparison for Surry 2, Cycle 3, MOC, HFP, ARO 5-20 5-16 Core Average Axial Power Distribution Comparison for Surry 2, Cycle 3, EOC, HFP, ARO .'5-21 5-17 Core Average Axial Power Distribution Comparison for Surry 1, Cycle 4, BOC, HZP, ARO 5-22 5-18 Core Average Axial Power Distribution Comparison for Surry 1, Cycle 4, BOC, HFP, ARO 5-23 5-19 Core Avarage Axial Power Distribution. Comparison for Surry 1, Cycle 4, MCC, HFP,1ARO 5-24 5-20 Core Average Axial Power Distribution Comparison for Surry 1, Cycle 4, EOC, 3FP, ARO 5-25 5-21 Core Average Axial Power Distribution Comparison for Surry 2, Cycle 4, BOC,- HZP, ARO 3-26 vi

' LIST OF FIGURES (Continued). Figure Title Page No. 5-22 Core Average Axial Power Distribution Comparison for Surry 2, Cycle 4, BOC, RFP, ARO. 5_27 5-23 Core Average Axial Power Distribution Comparison for Surry 2, Cycle 4, MOC, HFP, ARO. 5-28 5-24 Centrol Rod Worth Comparison for Surry 1, Cycle 1, BOC, HZP, D Bank 5-29 5-25 Control Rod Worth Comparison for Surry 1, Cycle 1, BOC, HZP, C Bank 5-30 5-26 Control Rod Worth Comparison for Surry 1, Cycle 1, BOC, HZP, B Bank 5-31 5-27 Control Rod Worth Comparison for Surry 1, ' Cycle 1, BOC, HZP, A Bana 5-32 5-26 Control Rod Worth Comparison for Surry 1,- Cycle 1, BOC, HZP, Banks B-D moving 100 step overlap. 5-33 5-29 Control Rod Worth Comparison for Surry 1, Cycle 2, BOC, HZP, D Bank 5-34 5-30 Control Rod Worth Comparison for Surry 1, Cycle 2, BOC, HZP, C Bank 5-35 5-31 Control Rod Worth. Comparison for Surry 2, Cycle 3, BOC, HZP, D Bank 5-36 5-32 Control Rod Worth Comparison for Surry 2, Cycle 3, i BOC, HZP, C Bank 5-37 5-33 Control Rod Worth Comparison for Surry 1, Cycle 4, BOC, HZP, D Bank 5-38 5-34 Control Rod Worth Ccmparison for Surry 1, Cycle 4, BOC, HZP, C Bank 5-39 5-35 Control Rod Worth Cotrparison for Surry 1, Cycle 4, BOC, HZP, B Bank 5-40 5-36 Control Rod Worth Co=parison for Surry 1, Cycle 4, BOC, HZP, A Bank 3-41 5-27 Control Rod Worth Comparison for Surry 1, Cycle 4, BOC, HZP, Banks A-D moving in 100 step overlap 5-42 5-38 Control Rod Worth Comparison for Surry 2, Cycle 4, BOC, HZP, D Bank 5-43 vii

LIST OF FIGURES (Continued) Figures Title Page No. 5-39 Control Rod Worth Comparison for Surry 2, Cycle 4, BOC, HZP, C Bank. 5-44 5-40 Control Rod Worth Comparison for Surry 2, Cycle 4, BOC, HZP, B Bank 5-45 5-41 Control Rod Worth Comparison for Surry 2, Cycle 4, BOC, HZP, A Bank. 5-46 5^42 Control Rod Worth Comparison for Surry 2, Cycle 4, BOC, HZP, Banks, A-D moving in 100 step over-lap. 5-47 \\ viii

LIST OF TABLES Table Title Page 2-1 Surry Nuclear Power Station Operating History. 2-4 2-2 Surry Core Description 2-3 5-1 Su= mary of Comparisons. 5-4 h ix t

I SECTION 1 - IXIRODUCTION The purposes of this report are to describe one of the computational models developed at Virginia Electric and Power Company (Vepco) and to demon-strate the accuracy of this model by comparing analytical results generated by the model to applicable measurements from Surry Units No. 1 and 2. The' capabilities obtained with this codel will be directly applicable to Surry Units 1 and 2 and generally applicable to all of the units at the North Anna Nuclear Power Station. The model described herein is a three-dimensional (x, y, z), one energy group, modified diffusion theory (with thermal feedback) calculational i package and is designated as the Vepco FLAME Model. The Vepco FLAME Model I uses the NULIF(1), PDQ07(2), FLAME 3(3), and FLATIT(3) computer codes which are part of the Fuel Utilization and Performance Analysis Code (4) (FUPAC) system obtained from the Babcock and Wilcox Company. In addition, the t EDITQAR(5), PICCOLO (5), and FLMSHUFL(6) computer codes have been written by Vepco for use in the Vepco FLAME Model. A detailed description of the input requirements, functioning, physical models and output capabilities of the ( above codes can be obtained from the referenced code manuals. { The types of reactor physics calculations which can be performed I ( within the general capabilities of'the Vepco FLAME Model include: 1. Assembly and core averags axial power distribution 2. Differential control rod bank worths 3. Integral control rod bank worths as a function of rod bank position 4. Control rod bank insertion limits 5. Axial burnup distribution 6. Axial offset 7. Peaking factors (F{,Fxy(Z), Fz) 1-1 ~

This report is concerned primarily with the documentation of the capability of the Vepco FLAME Model to accurately compute axial power distributions, axial o f fset, and differential and integral control rod bank worths. The remainder of this report describes the Surry Units No. 1 and 2 reactor cores to be modeled, the purposes and interrelationships of the various computer codes which comprise the Vepco FLAME Model, the specific modeling of a reactor core with these codes, and comparison of calculated results with appropriate results obtaimed wit'i the Vepcei PDQ07 Discrete Model(7) and with core measurements obtained from Surry Units No. 1 and 2. i i' l l 1-2

SECTION 2 - CORE DESCRIPTION 2.1 Introduction The Surry Nuclear Power Station, which consists of two operating units, has been selected as the operating system to be modeled for verification of the Vapco FLAME Model. The Surry Units No. 1 and 2 are identical Westinghouse designed three coolant loop pressurized water reacto-t with thermal ratings of 2441 Mwt. The operating historv of the Surry Power Station is su==arized in Zable 2-1. 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 pre 3sure vessel. The pressure inside the vessel is maintained at a nominal 2250 psia. The coolant (and moderator) is pressurized water which enters the bottom of the core at a nominal 532*F and undergoes a nominal average rise in tempera-ture of 65.5'F before exiting the core. The average coolant temperature is 566*F and the average linear power density of the core is 6.2 kw/f t. Each of the 157 fuel assemblies consists of 204 fuel rods arranged in a 15 by 15 square array. The fuel used in the Surry cores consists of slightly enriched uranium dioxide fuel pellets co.ntained within a Zircaloy-4 clad. A small gap containing pressurized helium exists between the pellets and the inner diameter of the clad. For the positions in the 15 by 15 array not occupied by fuel rods, there are 20 guide tube locations for either 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 gridc located along the length of the assembly. These grids are mechanically attached to the guide tubes, which are, in turn, fastened to the upper and lower noz les, and thus provide for assembly structural support. 2-1

I There are 48 full-length Rod Cluster Control Asse=blies (referred 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 inserted into the core). The absorber =aterial of the control rods is' an alloy consisting of 80% silver, 15% indium, and 5% cadmium. The various ~ control rods are arranged in and cove in sy==etrically located groups, or banks, as depicted in Eigure 2-2. Banks D, C, B, and A are denoted as the control banks and are coved in a fixed sequential' pattern to control the reactor over the pcwer range of operation. The remaining rods, Banks SA and SB, are denoted as shutdown banks and are used to provide shutdown =argin. In addition to the ecntrol rods, a che=ical (boric acid) shi= is uiad tc control excess core reactivity and to facilitate operational flexi-bility. Above certain ccncentrations of chemical sh1=, burnable poisen rods are also used to control excess reactivity. Fresh and/or depleted burnable poisen rods can also be used to shape (i.e., i= prove) the core power distribution. The burnable poison rods contain borosilicate in the for= of Pyrex gir,s clad in a stainless steel tube. Surnable poison rods, which may be used in any f tel assembly not under a control rod bank location, consist of clusters of either 8, 12, 16, or 20 rods which are inserted into the Zircalcy-4 control red guide tubes. Specific values of the principal neeSanical and ther=al-hydraulic parameters for the Surry core are provided in Table 2-2. A ccuplete descrip-tion of the Surry units is given in Reference 8. 2.3 Fuel Loadings The initial and reload quarter core fuel loadings (i.e., initial enrich =ents and density, previous cycle location if appropriate, beginning of cycle burnup predicted by the Vepco PDQ07 Discrete Model and nu=ber of 2-2

..=. - 1 fresh or depleted bu:nable poison rods present) for both Surry units are provided in.Figurea 2-3 through 2-15. It should be noted that the fuel loadings for Cycl.e 1 of both Surry units are identical. The fuel manage-ment 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-I of the core and the highest enriched fuel batch around the periphery of the core. After the first cycle,"the fuel management became more complicated' i-as the result of the need to minimize the impact of fuel densification (which, !i' was most severe in the lower initial density', lower prepressurization Batches 1, 2, and 3). Generally, a modified out-in strategy was followed wherein higher enrichment fresh fuel was loaded on the core periphe y with lower enrichment fresh fuel (and once-burned fuel and twice-burned fuel) checker-board loaded in the inner region of the core. An exception to this was in the third cycle of Unit No. l where no fresh fuel was loaded on the periphery. The only fresh fuel was 16 lower enrichment assemblies. loaded in the inner regten of the core. Beginning in Cycle 4 of both units, a change.was made from the typical 12 month operating cycle to an 18 month operating cycle. However, the basic fuel management strategy (1. e., modified out-in) was.not changed for the Cycle 4 loading patterns. I h r i t 7-3

TABLE 2-1 SURRY SCCLEAR F0WER STATION OPERATING HISTORY Surry Cycle Beginning of End of Cycle Burnup , Unit. No. Ovele Cvelo OEG)hfrU) 1 1 July 1, 1972 October 24, 1974 13547 1 2 January 30, 1975 September 26, 1975 6915 1 3 Decemb'er 6, 1975 October 17, 1976 8944 1 4 Janua ry 17, 1977 April 21, 1978 13107 2 1 March 7, 1973 April 26, 1975 14870 2 2 June 14, 1975 April 22, 1976 9v54 2 3 June 1, 19 76 September 10, 1977 9422 2 4 October 8, 1977 1st Qtr 1979* 14000*

• Proj ected 2-4

TABLE 2-2 SURRY CORE DESCRIPTION THERMAL AND HYDRAULIC DESICN PARAMETERS Total core heat output, Mwe 2441 Heat generated in fuel, % 97.4 System operating pressure, psi 2250 Total coolant flow race, lb/hr (gpm) 100.7 x 106 (265,500) Coolant Temperatures, F (@l002 power) Nominal inlet 532 Average rise in the core 65.5 Average in the core 566 Nominal outlet of hot channel 642 Average linear power density, Kw/ft 6.2 MECHANICAL DESIGN PARAMETERS Fuel Asse=blies Design ~ Number Canless 15 x 15 157 Rod pitch, inches 0.563 Overall dimensions, inches 8.426 x 8.426 Number of grids per assenbly (meerial) Number of instrumentation tubes 7 (Inconel-718) 1 Fuel Rods Nu=ber 32,028 Nu=ber of rods / assembly 204 Outside diameter, inches Batch 1,2,4,5.6 Batch 1 0.422 0.422 Diametrical gap, inches 0.0075 0.0085 Clad thickness, inches Clad material 0.0243 0.0243 2ircaloy-4 Fuel Pellets Material Sintered UO3 Density (% of theoretical) and Enrichment (w/o U235) See Figures'2-4 through 2-12 Outer diameter Batch 1,2,4,5,6 Batch : 0.3659 0 3649 Control Rod Assemblies Neutron absorber 5% Cd-15 In-80: Ag Cladding Haterial Clad thickness, inches Type 304 SS-Cold worked 0.019 Number (full length) 48 Number of rods per assembly 20 2-5

TABLE 2-2 (Continued) Burnable Poison Kods Material Pyrex glass Content B2 3 (w/o) 12.5 0 Core Structure Core barrel I.D./0.D., inches 133.875/137.875 g; Thermal shield I.D./0.D., inches 142.625/148.000 Core diameter, inches (approximate) 119.5 Reflector thickness (approxi= ate) and composition Top - Water plus steel, in. 10 Bottom - Water plus steel, in. 10 Side - Water plus steel, in. 15 e + 2-6

FIGURE 2-1 CROSS SECTIONAL VIE'J OF SURRY FUEL ASSEMBLY fe ,,, y m.,...... o.','.'.'! ? - .j g i.l._.... P',' ~a 000000000000000 00000000C0 ! OOOOGfC.C 0000000O C $-O 0 0 0 G'0 0 OC 0C000000 ~ O O O v O C 0 0! O O O 000 C 'u. c "'Q,o O 0 O O o ')c:0 O O O ' O 0: O O 000 O O1 O O O O O O C' O O O 0l4 O O O O 'O OO O O O O O O O O O O O! O O O O O O! O O O O OC 0 O O O O O! O O*I O O 000000000000000 000000000000G00o/ 000000000000000 OOOOOOOOOOOccW'f O O O O O O OMC"d " - a <* ~ 000 O O OO O O O COcr "u O --""> 0 0 0 0 0 Od O O O 06 O O 0000 OO O O! O O 0 0 0: l O O! O O ~ ~ " " ' 0 C: O O O O 0 0 O + 0 ~ O O Oi - ~ ~ - ~ Oa OO OO O O I O 01 O O O Ol O O O O 0 0 O Ol l OO O O O O, 0 O O O O O! "a'"'- Oi O O O 0000000e0000g_O0! 0000000e0000000 _. 4....,. I Z'g

• F

_4 .T.;, "'.* ".. ;'!. 'O ~'O 7 Tai . m.. u.... 2-7

FIGURE 2-2 CONTROL ROD BANK LCCATIONS A P N M L K J

• H G

F E D C B A 1 l I t_ _i e A D A 3 S S I 4 C B P B C = i , l 5 S lS 6 A B D C D B A 7 S s s s 8 D P C P C. P D 9 'S S S S i l 1 10 A B D C l g.Dl B A 1 r-r r l 11 s s 12 C B P B C 13 S S 14 A D A 15 l CONTROL ROD ASSEMBLY BANKS l Function Number of Assemblies Cancrol Bank D 8 Control Bank C 8 Control Bank B 8 l Control Bank A 8 Shutdown (S) 16 Part Length (P) 55 SOURCE ASSDiBLY LOCATIONS 2-8

FIGURE 2-3 'SURRY UNI *IS 1 AND 2 - CYCLE 1 FUEL LOADING 08 09 10 11 12 13 14 15 1 2 3 2 'l 2 1 3 H 0 0 0 0 0 0 0 0 Fresh Fresh Fresh Fresh Fresh Fresh Fresh Fresh 2 1 2 1 2 1 3 3 y 0 0 0 0 0 0 0 0 Fresh Fresh Fresh Fresh Fresh Fresh. Fresh Fresh i 1 2 1 2 1 2 3 7 0 0 0 0 0 0 0 Fresh Fresh Fresh Fresh Fresh Fresh Fresh' 2 1 2 1 2 3 3 0 0 0 0 0 0 0 Fresh Fresh Fresh-Fresh' Fresh

Fresh, Fresh 1

2 1 2 Il 3 g O O O O O O Fresh Fresh Fresh Fresh Fresh Fresh 13iCi31 2 1 2 3 3 N Batch Enrich =ent - Density 0 0 0 0 0 (#F.A.'s) w/o U235 %TD ~ Fresh Frash Fresh Fresh Fresh 1 (53) 1.85 94 1 3 3 ~ 3 2 (52) 2.55 93 p 3 (52) 3.10 92 0 0 0 0 Fresh Fresh Fresh Fresh R ,O O Fresh Fresh. LEGEND xx - Batch No. U --Initial Burnup (MWD /MTU) --Previous Location (If applicable) 2-9

FIGURE 2-4 SURRY UNIT 1 - CYCLE 3 FUEL LOADING 08 09 10 11 12 13 14 15 i 43 1 43 2 45 i sc H 15358 0 12536 0 15298 0 14680 0 !!08 Fre si. H14 Fresh H13 Fresh H12 Fresh 4B 2 4A 2 2 2 4C 4C 0 16435 0 14299 16075 14378 0 0 F-peh LO8 Fresh Y1'- K11 L12 Fresh Fresh 1 4A 1 4A 2 1 4C K 12536 0 1126 0 15918 14191 0 l P08 __, Fresh M12 Fresh J12 X12 Fresh 4B 2 4A 2 2 4C I 4C L 0 14299 0 16755 16599 0 0 Fresh N10 Fresh J08 J10 Fresh Fre4h 2 2 2 2 4A 4C u 15298 16075 15918 16599 0 0

N08, L10 M09 K09 Fresh Fresh 43 2

1 4C 4C Initial Batch Enrichment Censity 0 14378 ~ 14191 0 0 f#E.A 's) w/o U235 gro Fresh M11 M10 Fresh Fresh 1 (21) l'85 94 1 4C 4C 4C 2 (52) 2.55-93 4A(20) 1.85 95 P 14680 0 0 0 4B (12) 2.60 95 4c(52) 3.35 9~a l

M06, Fresh Fresh Fresh 4C 4C 0

0 R Fresh Fresh LEGEND xx ---Bacch No. yy ---Initial Burnup (MWD /MTU) ---Previous Location (If applicable) zz 2-10

v FIGURE 2-5 SURRY UNIT 2 - CYCLE 2, FUEL LOADING 08 09 10 11 12 13 14 15 C 1 3 3 2 4A 2 2 4B I H 16693 15422 11414 17992 0 18295 16805 0 JO9 G14 HIS M11 Fresh H09 C08 Fresh 3 4A 2 4A 2 3 4B AB 15422 0 17635 0 '.5755 14725 0 0 P09 Fresh K11 l Fresh K13 M11 Fresh Fresh 1 3 2 3 2

• A 4
• o 11414 17635 15422 18151 0

'17470 0 K R08 L10 J14 J10 Fresh J12 Fresh 2 4A 2 4A 2 4B 4B L 17992 0 18151 0 15860 0 0 LO8 Fresh K09 Fresh L12 Fresh Fresh 4A 2 4A 2 4A 4B 0 15755 0 15860 0 0 Fresh NIO Fresh Mll __, fresh Fresh 2 3 2 4B 43 Batch Enrichment Density (JF.A.'s) w/o U235 ?.TD ' N 18295 14.725 17470 -O' 0 J08 L13 M09 Fresh Fresh *j' 1 (1) 1.85 94 2 (52) 2.55 93 2 4B 43 4B 3 (20) 3.10 92 4A(32) 2.60 94 P 16805 0 0 0 4B(52) 3.10 95 H13 Fresh Fresh Fresh 4B 4B R 0 0 Fresh Fresh LEGEND xx ---Batch No. yy ---Initial Burnup (MWD /MTU) ---Previous Location (If applicable) zz 2-11

FZGURE 2-6 SURRY UNIT 1 - CYCLE 3, FUEL LOADING 08 09 10 11 12 13 14 15 1 3 1 3 3 3 3 4C H 15236 10420 14182 14034 14034 12224 8387 '439 cyt w cy1 unt c,. i n :, c.. i ets cyi vi s ey! xis cy1 i_is i cy, uni 3 4B 3 4A 3 Il 4C 4C J 10420 8568 8973 7946 13334 13859 7142 4903 Cv1 A06 Ov2 L0f Cy1 C12 Cy2 J10 Cy1 203 Cy1 J13 Cy2 F14 Cy2 G01 1 3 4A 5 5 3 4G I K ~ 14182 8973 6462 0 0 8174 6967 Cvl LIJ Cv1 MOS Cv2 M12 Fresh Fresh Cvl JIS Cv2 CO2 3 4A 5 3 4A 'C 4C 4 L 14034 7946 0 12224 7486 7601 5667 Cvl P0c Cv2 KOc Fresh._ 'Cv l P10 Cv2 K11 Ov2 E03 Cv2 D03 3 3 5 4A 3 l 'C 4 M 14034 13334 0 7486 8377 5230 Cy1 P0-Cy1 C05 Fresh Cy2 L10 Cy1 305 Cy2 E02 Initial d 1 3 3 Batch Enrichme'.t Density I (#F.A.'s) w/o U2.L TD N 12224 13859 8174 7601 5230 Cvl P06 Cv1 NOC cvl R0o Cv2 C05 Cv2 B05_ l 3 4C 4C 4C 4A(20) 1.85 95 p 837/ 7142 6967 5667 ) 6) 2.10 M Cy1 P05 Cy2 P0t Cy2 B07 Cy2 C04 4C 4C R 6239 4903 Cv2 A0E Cv2 AO: LEGEND xx ---Batch No, yy ---Initial Burnup (MID/MTU) ---Previous Location (If applicable) z: Previous cycle-- aa 2-12 n

FIGURE 2-7 SURRY UNIT 2 - CYCLE 3 FUEL LOADING 08 09 10 11 12 13 14 15 'l 43 4A 3 1 48 1 43 ~ H 16635 10342 10943 12414 15214 10342 13761 6985 Cv1 vna c._. ' f i r. c9 cnol r.i vis E n ry? cia c..e uti . :,,, uni 4B 1 4B 4A 4B 4A 43 5 J 10342 15214 10076 11093 8505 11101 5739 0 Cy2 P07 Cy1 N09 Oy2 L13

y2 K12 Cy2 K14 Cy2 J11 Cy2 J15 Fresh 4A 4B 4A 3

1 3

• s 7

K 10943 10076 10916 9228 15586 8978 5972 Cv2 J0c Cv2 Nil Cv2 M08

vl Ll4 Cvl K12 Cvl J'S cv2 LIA 3

4B 3 1 3 4B 5 L 13414 11093 9228 16377 9896 6806 Cy1 POE Cy2 M1C Cy1 Pil Ty1 LO9 Cy1 M13 Cy2 M13 Fresh 1 4B 1 3 3 5 M 15214 8505 15586 9896 13414 0 Cy1 N07 Cy2 PIC Cy1 M10

y1 N12 Cy1 P10 Fresh 4B 4A 3

4B 5 Initial Latch Enrichment Density 10342 11101 3973 6806 0 p (#F.A.'s) w/o U235 %TD Cy2 P05 Cy2 LOS Cy1 R09 Cy2 N12 Fresh i (25) 1.85 94 3 (32) 3.10 92 1 4B 43 5 4A(24) 2.60 94 P 13761 5739 5972 C 4B(52) 3.10 95 j 5 (24) 3.10 95 Cy1 POE Cy2 ROS, Cy2 P11 Fresh Io 5 R 6985 0 Cy2 A08 Fresh LEGEND xx ---Batch No. yy ---Initial Burnup (M'JD/MTU) zz ---Previous Location (If applicable) Previous Cycle--- nn 2-13 e

T FIGURE 2-8 SURRY UNIT 1 - CYCLE 4 FUEL LOADING 08 09 10 11 12 13 14 15 1 6A 5 6A AB 5 4B 6C 1 H 15236 0 10552 0 8760 10552 8615 0 Cy1 H06_ l Fresh Cv3 F1: Fresh Cv2 H09 Cy3 E10 CY2 H13 Fresh 6A 4C 6A 4C 4C 4C 6C 0 6C 0 17296 0 10529 11355 1495-_ 0 Fresh Cy3 N11 Fresh Cy3 A09 Cy3 N04 Cy3 F02 h Fresh Fresh -n- - 3 6A 2/4A 6A 4C 6B 6C K 10552 0 9890 0 16875 0 0 Cy3 L10 Fresh S2C2 M12 Fresh Cy3 J14 Fresh Fresh 6A 4C 6A 4C 4C 6C 6C L 0 10529 0 17287 11267 0 0 Fresh Cy3 J01 Fresh Cy3 L13 Cy3 L14 Fresh Fresh 4B 4C 4C 4C 4C 6C M 8760 11355 16875 11267 13240 0 Cy2 J08 Cy3 D13 Cy3 P09 Cy3 P11 Cy3 R08 Fresh 5 II4C 6B 6C 6C I"iCI"1 Eatch Enrichment Density g 10552 14951 0 0 0 (#F.A.'s), w/o U235 %TD Cy3 K11 Cy3 B06 Fresh Fresh Fresh 1 (1) 1.85 94 2/4A (4) 2.60 95 4B 6C 6C 6C 48 (8) 2.60 95 P 8615 0 0 0 4C (52) 3.35 95 5 (8) 2.10 95 Cy2 N08 Fresh Fresh Fresh 6A (24) 2.60 95 6B (8) 2.60 95 6C 6C 6C (52) 2.90 95 I o o g Fresh Fresh LEGEND xx ---Batch No. yy

---

Initial Burnup (MWD /MTU) zz Previous Location (If applicable) 2-14

FIGURE 2-9 SURRY UNIT 2 - CYCLE 4 FUEL LOADING 08 09 10 11 12 13 14 15 1 6A 4B 4B 4B 4B 4B 6B H 16635 0 21458 13392 21240 20813 17227 0 Cy1 F08 Fresh Cy3 G10 Cy3 HIS Cy3 H09 Cy3 H13 Cy3 E13

Fresh, 6A 4B 6A 5

5 4B 6B 6B J 0 19642 0 5849 6678 14467 0 0 Fresh Cv3 M09 Fresh Cy3 J15 Cy3 L14 Cy3 K14 Fresh Fresh 4B 6A 4B 5 4B 6B 6A K 21458 0 19630 7340 15745 0 0 Cy3 K09 Fresh J12 Cy3 M13 Cy3 J14 Fresh Fresh 4 4B 5 5 4B 6A 63 6B L 13392 5849 7340 21464 0 0 0 Cy3 R08 Cy3 R09 Cy3 N12 Cy3 J10 Fresh Fresh Fresh 4B l5 bI 6A 4B 6B 21240 6678 15745 0 17235 0 Cv3 J08 Cv3 Pil Cv3 P09 Fresh Cy3 L13 Fresh 7 4B 4B 6B 63 6B Batch -Enrichment Density (#F.A.'s) w/o U235 %TD 20813 14467 0 0 0 fB(52) 3$10 95 cv3 N08 Cv3 P10 Fresh Fresh Fresh 4B 6B 6A 6B 5 (24) 3.10 95 6A (28) 2.90 95 P 17227 0 0 0 6B (52) 3.20 95 Cy3 N11 Fresh Fresh Fresh 6B 63 0 0 R Fresh Fresh Leeend xx Batch No. yy ----Initial Burnup (5'U/FrIU)

---

Previous Location (If applicable) zz 2 - 15

...~. _ _ - - -

FIGURE 2-10 SURRY UNITS 1 AND 2 - CYCLE 1 3urnable icison Rod Loading 08 09 10 11 12 13 14 15 1 2 1 2 1 2 11 3 H 12 12 12 I J 2 1 2 1 2 1 3 3 12 12 12 12 I O O l 11 2 1 2 1 2 3 12 12 12 = 2 1 2 1 2 3 3 L 12 12 12 12 0 1 1 2 1 2 1 3 M i I 2 1 2 3 3 .N 12 12 12 1 3 3 3 12 3 3 r R LEGEND xx Batch No. yy No. of Fresh Burnable Poison Rods 2 - 16

FIGURE 2-11 SURRY UNIT i -- CYCLE 2 BURNABLE POISON ROD LOADING 08 09 10 11 12 13 14 15 1 4B 1 4B 2 4B 1 4C H 8 F. 12 4B 2 4A 2 2 2 4C 4C 8 20 ~ n l 4A 1 4A 2 1 4C K ..I 43 2 4A 2 2 4C 4C 8 12 2 2 2 2 4A 4C M 4B 2 1 4C 4C 12 12 ~ 1 4C 4C 4C 20 4C 4C R Batch.No. xx Ey No. of Fresh Iornable Poison Rods z:t. g,---No. of Depicted Burnable Poison Rods 2-17

FIGURE 2-12 SURRY WIT 2 - CYCLE 2 BURNABLE POISON ROD LOADING 08 09 10 11 12 13 14 15 1 3 3 2 4A 2 2 4B H 12 3 4A 2 4A 2 3 43 4B J 3 2 3 2 l 4A 2 I 4B K = 2 4A 2 4A 2 4B 4B L I 12 4A 2 4A 2 4A l 4B l M 2 3 2 4B 43 5 12 12 2 4B 4B 4B P 4B 4B i LEGEND. xx --Batch No. z - No. of Depleted Burnable Poison Reds 2-18

FIGURE 2-13 SURRY UNIT 1 - CYCLE 3 BURNABLE POISON ROD LOADING 08 09 10 11 12 13 14 15 3 1 3 3 J 3 4C H 12 3 4B 3 4A 3 1 4C 4C J A ~ 3 4A 5 5 3 4C K 12 3 4A 5 3 4A 4C 4C L 3 3 5 4A 3 4C M 3 1 3 4C 4C N 12 12 3 4C 4C 4C P 4C 4C R ? ynnm xx ~ Batch No. zz --No. of Depleted Bumable Poise 1 Rods 2-19 . ~. _ _..... _ _ _. _..

I FIGURE 2-14 SURRY 1,3IT 2 - CYCLE 3 BtJRNABLE POISON ROD LOADING 08 09 10 11 12 13 14 15 4 1 4B 4A 3 1

• n 1

43

• H 12 4B 1

4B 4A 4B 4A 4F 5 J 12 12 4A 4B l 4A 3 1 3 4B ~ 12 12 12 3 4A 3 1 3 4B 5 L 12 12 mesi.: h 4B 1 3 3 15 a a 12 12 4B 4A 3 4B 5 't 12 12 1 AB 4B 5 'd f r 2

• B 5

t LEGEND xx - Batch No. zz - No. of Pepleted Burnable Poison Rods 2-20

FIGURE 2-15 SURRY UNIT 1 - CYCLE 4 l BURNABLE POISON ROD LOADING I 08 09 10 11 12 13 14 15 f 1 6A 3 6A 4B 3 4B 6C H 8 12 12 6A 4C 6A 4C 4C 4C 6C 6C J 8 12 8 12 8 1 i 5 6A 2/4A 6A 4C 6B 6C l K 12 12 12 t f r 6A 4C 6/. 4C 4C 6C 6C 12 8 12 8 8 I 4B 4C 4C 4C 4C f.C M 12 8 5-4C 6B 6C 6C 98 12 8 12 4B 6C 6C 6C P 8 ( 6C 6C l R xx

---

Batch No.

yy ----No. of Fresh Burnable Poison Reds zz

---

No. of Depleted Burnable Poison Rods 2-21

FIGL3E 2-16 SURRY UNIT 2 - CYCLE 4 BURNABLE POISON P D LOArING 08 09 10 11 12 13 14 15 1 6A 4B 4B 4B 4B 4B 6B H 12 12 6A 48 6A 5 5 4B 6B 6B J 12 12 12 8 12 4B 6A 4B 5 43 6B 6A K 12 8 20 4B 5 5 4B 6A 63 6B L 12 8 12 16 4B 5 4B 6A 4B 6B M 8 12 4B 4B 6B 6B 6B N 20 16 12 4B 6B 6A 6B P 12 t l 6B 6B R xx -Batch No. yy

---

No. of Fresh Burnable Poison Rods 2::

No. of Depleted Burnable Poison Rods 7 ?? - ~ - -.

SECTIOS 3 - MODEL DESCRIPTION 3.1 Introduction The Vapco FLAME Podel is used to calculate nodal power densities and core reactivity for three-dimensional geometries in which each fuel assembly is represented by one radial node and by up to 32 axial nodes. The method used by the Vepco FLAME Model to perform these calculations is based on the FLAREi9) technique which is derived from modified, one energy group diffusion theory. Effects of nonuniform moderator density and fuel temperature are accounted for by thermal-hydraulic feedback. The Vepco FLAME Model perfor=s calculations in several steps. First, a fine-group neutron flux spectrum and the appropriate cross sections as a function of neutron energy are calculated for each material composition in the fuel assembly by a cross section generating code (1,e., the NULIF(1) computer code). Then the fine-group flux spectrum is used to spectrum weight and collapse the fine-group cross sections into the one energy group parameters required (i.e., K. and M )' by the FLAME 3(3) computer 2 code. These parameters, as well as any appropriate normalization factors, are then used by FLAME 3 to perform an iterative, modified diffusion theory calculation for the neutron production rate density as a function of position. The method of solution comprises two levels of iteration: source (or nuclear) and thermal-hydraulic feedback. The source calculation is performed first based on an initial guess for the thermal-hydraulic parameters. Then a new set af thermal-hydraulic parameters are calculated. Using these new thermal-hydraulic parameters, another source c lculation is performed. This process is continued until both the source and thermal-hydraulic convergence criteria are satisfied. 3-1

Several interrelated computer codes are used to perform the calculations outlined above. The computer codes comprising the Vepco FLAME Model and.. air interrelationships are presented in the flow chart in Figure ~ 3-1. T*a FLAME 3 computer code is the principal analytical tool in the Vepco FLAM' Model. FLAME 3 is used to perform the three-dimensional, one group, modified diffusion theory calculations. The other codes proviie either input data or data manipulation. As indicated in Figure 3-l, the NULIF code is used to generate the required one group data for the non-rodded fuel assembly. The FLAFIT(3) code formats these data for use by FLAME 3. Data for burnable poison and control rods are generated with'the Vepco PDQ07 Discrete Model(7) using a quarter assembly discrete representation. The in-house codes EDITQAR(3) and PICCOLO (5) format the data from the quarter assembly PD007 calculations for use by FLAME 3. The FLMSHUFLO) code is a data manipulation code that shuffles the appropriate end-of-cycle burnup distributions to duplicate the movement of fuel assemblies during refueling. l rhe shuffled burnup distributions from FLMSHUFL are input to FLAME 3 to be;;in i reload cycla analyses. l 1he remainder of this chapter desc.ribes in greater detail the input _ to aed functioning of the computer codes used in the Vepco FLAME Model. 3.2 Input Preparation The Vepco FLAME Model uses the FL/.RE neutron source option for all calculations with the FLAME 3 code. Two physics parameters are required at each node by the FLARE (9) option: the infinite multiplication factor (K.) and the migration area (M ). At peripheral nodes,'it may also be necessary 2 to input a leakage parameter (or albedo); ' The infinite multiplication factor, migration area, and albedoes for - each node are usually functions of one or more variables. Among these are initial 3. ---

^* ^ c hROSSSECTIOtt DEPLETION FUEI. ASSEtiSLY CROSS SECTIO 1 1/4-ASSET!BLY DEPLETIO3 GEI!ERAL l LIPP.'iRY ll3PUT s DESCRIPTION TAP.IESETS CE0!!ETitY INPUT PROSLEll l I I I I I 4 ).. 14ULIF PDQ07 1 m W O. s n. FLAFIT EDITQAl: f Id n o @A 9 0 .,,, tes u

• Oou O 30 FICC01.C p

H K., DATA FOR TitC tlOIl-2 RCDDED ASSZ:: GLY & tl o f O I FOR EC&CTOR NORit4LIZATION CENF.RAL Ef{flL E! pSORIPTIO!! FACT 03S PR07.1. Cit ,a l I J BUltdUP FLAttE3 1 DISTP.IBUTInti "L!!S!!UFi AT TOC n AXIAL POWER ROD Weitris .iXI4L BURdVP DISTellt.UTION SiiAPES DISTR Irdr(IO:'

enrichment, burnup, soluble boron concentration, xenon concentration, fuel and mouerator t>mperatures, and the presence or absence of burnable poison or control rods. To represent the functional dependencies, K. and M2 are calculated for various combinations of the appropriate variables. The resulting 2 values of K. (or M ) are then tabulated for use in the FLAME 3 cede by using the FLATIT and PICCOLO codes. The FLAME 3 code uses a Lagrangian interpolation routine for selecting the appropriate input data from the tables constructed by FLAFIT and PICCOLO. This routine allows any input variable (e.g., K.) to be fit as a function of one, two, or three independent variables. Due to this constraint, it has been necessary to use several tables to properly represent all the functional dependencies of K,. The leakage parameters cannot be input in tabular form. However, the leakage parameters may be changed to reflect changing reactor conditions. This is discussed further in Section 4.1. The processes by which tabular input for K. and M2 are developed are described in this section. 3.2.1 Method of Calculating Input Data for the Non-Rodded Fuel Assembly The NULIF code is used to generate the non-rodded (i.e., no control or burnable poison rods are present) fuel assembly input data for the FLAME 3 code. The NULIF calculations are performed using the supercell option. l A supercell is defined as a group of unit cells cceprising, for example, a fuel assembly. For the supercell group, the fuel rod unit cell is designated as the central cell. Any other unit cell types (e.g., control rod guide tube cell) present in an assembly are designated as subregion cells (subcells). The supercell option is used to represent the fuel assembly in the Vepco FLAME Midel because only homogenized groups of unit cells can be represented due the larger-than-discrete mesh description ieplicit in a nodal representation of the reactor core. The homogenization is performed in a manner that 3-4

results in ene node representing each fuel assembly. i The calculetion of the neutron energy spectrum and the spectrum-weighted two-group cross sections for each supercell is described in detail in Reference 1. Those aspects which involve calculations using the supercell option are described below. With the supercell option of the NULIF code, the material compositions of the central cell and the various subcells must be homogenized together before the fine-group neutron flux is calculated, since the NULIF code does not perform a spatial calculation for the various subregions. Because of this, a methcd must be employed to represent the heterogeneous nature of the supe rcell. This is done by inputting appropriate thermal flux depression 1 factors for each subcell relative to the supercell. These flux depression factors are generated by a detailed spacial calculation (i.e., a quarter assembly discrete PDQ07 calculation where each fuel rod, thimble cell, and water channel associated with the fuel assembly is explicitly represented). 4 From this detailed spatial calculation, the ratios of the thermal flux in-the average fuel cell, enirble cell, and water gap relative to the thermal flux d in the entire assembly a e determined. The above flux de;ression factors are combined in NULIF with thost. rormally calculated by NULIF for the central ~ i cell (i.e., the flux distribution in the fuel pellet, clad, and moderator regions of the central cell) to give the overall flux depression factors co be applied over each of the 80 thermal fine-groups for each nuclide in the supercell. i' NULIF calculates the neutron flux in the supercell for each of 31 fast and 80 thetmai energy fine groups. The macroscopic, one-energy group parameters needed as input to the FLAME 3 code are then determined from the neutron' flux and cross sections-for each fine group. 3-5 i. ~~ y, m... . ~. _. _, - _ =. _.. -.,...

2 Assenbly (supercell) caletdations of K. and M sich the NULIF code are performed as a function of: 1. Initini enrich =ent 2. Burnup 3. Saluble boron concentration 4. Moderator specific volume 5. Fuel temperature 6. Xenon concentration An assembly of each enrichment used in the core is depleted with NULIF while maintaining constant soluble boron and xenon concentrations, moderator speci-fic volume and fuel temperature. Recovery cases are then perfor=ed at selected burnups where these parameters are varied. Care has been taken to e vary each parameter over the range of values expected to occur during the types of operation to be modeled with the Vepco FLAME Model. 4+ 3.2.1.1 K-Infinity The Vepco FLAME Model uses for each enrichment a basic K. which is a function of burnup, soluble boron concentration, and xenon concentra-tion. This combination of independent variables was chosen for tvc reasons. First, these variables have the largest effects on K. for the non-rodded assembly (aside from enrichment which, as stated above, is represented by generating separa:e tablesets for each initial enrichment). Second, the effects of these variables are relatively insensitive to changes in other parameters. For these reasons, the functional dependencies of K. on burnup, boron, and xenon are represented together in the basic K.. Since K. cannot be adequately represented as a function of only three independent variables (the limit for any table used in the FLAME 3 code), other tables (i.e., K. multipliers) are developed to allow the incorporation of addit tonal functicnal dependencies. The K, multiplier tables contain

l factors ubich are used to modify the values contained in the basic Km tables. Th::se facters are arrived at by taking the ratio of K. at scme nominal condition to the K, generated by the NULIF code after a change in one or two independent variables. For the non-codded fuel assembly, two K. multipliers are used to represent the effects of fuel and moderator temparatures. The first K. multiplier represents the effects on K. of changing moderator camperature. Changes in moderator temperature result in two important phenomena: 1) change in moderation (neutron spectrum) and 2) change in parasitic absorption in the moderator. For these reasons, the first K. multiplier is fit as a function of moderator specific volume, burnup, and soluble baron concentration. The second K, multiplier incoctorates the dupendence of K. on fuel temperature. Changing fuel temperature causes a change in resonance abso rp tion. This effect is senaitive only to burnup. Therefore, the second K. multiplier is fit as a function of fuel temperature and burnup. 3.2.1.2 Migration Area The migration area, M, is used by the FLAME 3 code as a measure of 5 the leakage from an assembly. As such, M'is sensitive only to moderator 2 specific volu=e, burnup, and enrichment. Therefore, M can be adequately represented by a single table fit as a function of these three parameters. 2 In addition, the FLAME 3 code permits the use of a M multiplier as 2 a normalization f actor. The use of M multipliers is discussed in Section 4.1. 3.2.1.3 Input Parameters for Calculating Xenon Concentration The FLARE nectron sourts option does not allow the explicit calcu-lation of any material concentrations. However, to accurauely represent K. it is necessary to know the xenon concentration at each node in the core. 2nce the nodal xenon concentrations are found, then K for each node may be obtained from the appropriate tables. 3-7

i To find the nodal xenon concentrations, the FLAME 3 code solves the differential equations for xenon and iodine based on the assumption of constant power throughout a burnup step (see Section 3.4 ). This solution recuires that several parameters be input to the FLAME 3 code in tubular form as functions of the appropriate independent variables. These input parameters are listed below: 2 f2 2)(1/05) 0 A = (K E 0 1 f1 7+ K I fl l+ YIlf 20 )(1/02) 0 2 YI = (3gI YX = (yx g1 t+ yx f 2 2)(1/02) I 0 I 0 c,Xe = thermal microscopic absorption cross section of xenon, Xe SIGM = multiplier for o2 a where: K = energy released per fission %g = macroscopic fission cross cection 0 = neutron flux Y I = yield per fission of Iodine-135 Yx = yield per fission of Xenon-135 and subscripts 1 and 2 refer to the fast and thermal energy groups respectively. The parameters are all generated by the NULIF code and are. tabulated by FLAFIT for use in the FLAME 3 code. The parameters A, YI, and YX are strongly dependent on burnup and enrichment. Howevar, variations in other ind "ndent variables do not signi-ficantly influence these parameters. For this reason, A, YI, and YX are fit as a function of burnup and enrichment only. The thermal microscopic absorption cross section of xenon is dependent upon the neutron energy spectrum in the fuel pellet. The neutron spectrum is in curr. dependent upon the amounts of absorbing and fissile material in the fuel and the moderation which occurs outside the fuel pellet. X Therefore, c j is fit as a function of burnup, xenon concentration and enrich-ment. To account for the spectral effects of the moderator, the multiplier

for e Xe', (SIGM), la fit as a function of soluble boron concentration and a2 modelator specific volume. 3.2.2 Method of Calculating Input Data for the Burnable Poison Rodded Assembly 3.2.2.1 Int;oduction The input data for an assembly containing burnable poison (BP) rods are calculated with the Vepco PDQ07 Discrete Model using a 2-D, quarter-assembly representation. Theoretically, these data could be obtained by NULIF supercell calculations (as described in Sectic.n 3.2.3 above) in which one of the subcells contains the appropriate amount of BP. However, the fact that the hetergeneous effect of the various subcells are accounted for by applying ther=al flux depres-sion factors determined from quarter assembly discrete calculations must be con-sidered. In general, for assemblies containing no BP, these flux depression factors do not change significantly as a function of burnup, so that the flux depression factors calculcted at zero burnup are adequate at any stage of deple-tion. However, this is not the case for subcells containing BP because the BP depletes rapidly with increasing burnup causing the flux depression factors to vary significantly with burnup. Since a large number of quarter-assembly discrete calculations would have to be' made to calculate chese factors for input to NULIF, it has been decided to use the quarter assembly runs directly to generate K. multipliers to account far BP reactivity effects. These multipliers (in conjunction with the K. data for the non-rodded fuel assembly discussed in Section 3.2.1) are applied to nodes which contain (or have contained) BP rods. Four K. multipliers are used in the Vepco FLAME Model to represent Be rods. Two multipliers represent the presence of fresh BP in a fuel assembly. The third cultiplier accounts for the presence of depleted BP. The fourth K. 3-9

multiplier'is used to represent the reactivity "after-effect" resulting from the depletion of a fuel assembly containing fresh BP rods. 3.2.2.2 K-Infinity Multipliers For Fuel Cantaining Fresh 1P } The presence of fresh BP rods in a fuel assembly changes that i assembly's characteristics in two ways. First, a large concentration of non-i fissile, neutron absorbing material is added to the assembly. Second, moderator j is displaced from the asse=bly. The reactivity effects associated with these changes depend on initial assembly enrichment, cumulative buraup of the assembly while BP rods are present, soluble boron concentration, and moderator temperature. To represent these effects, two K, multipliers are used. I The first K. multiplier table for BP accounts for the dependence on 4 enrichment, cu=mulative burnup, and soluble boren concentration. To generate I this table, assemblies of various enrich =ents are depleted at nominal moderator-and fuel temperatures with and without BP rods using a quarter assembly repre-sentation of the Vepco PDQ07 Discrete Model. At a number of burnup points, recovery cases are performed during which the soluble boron concentration is changed. The values of the first K. multiplier are then obtained by racioing K= in the rodded state to K-in the non-rodded state. The PICCOLO code is used to tabulate these values as a function of cumulative burnup, initial enrichment, and soluble boron concentration. This procedure is followed to construct separate tables for configurations of 8, 12, 16. and 20 BP rods per assembly. i The second K multiplier accounts for the variation in the reactivity effect of fresh BP with changes in moderator temperature. Sensitivity calcula-r tions have shown that this multiplier is not significant1y' affected by enrichment or number of BP rods per assembly. (10) Therefore,.the second K, multiplier is generated for an assembly of average enrichment (i.e.,-2.90 w/o U235) containing i 1-10

. _ _.. =_ m. i 1 the most co==only used configuration of BP rods (i.e.,12 fresh BP rods per assembly). Again using the Vepco PDQ07 Discrete Model, a 2.90 w/o enriched assembly is first depleted with 12 fresh BP rods present and then depleted without the presence of any BP. At several burnup points during the above q two depletions, additional cases are performed in which moderator te2per:ture and soluble boron concentration are varied simultaneously. The values of the second K. multiplier are obtained in two steps. First, for each set of conditions, the value of K, obtained f rom the deple 1 case with BP rods is divided by the value of K. obtained from the depletion case without BP 1 rods. Then the ratios determined in the first step cre divided by the ratio 1 of K. with BP rods to K. without BP rods obtained for the nominal moderator temperature (i.e., 566 F). The PICCOLO code is then used to tabulate these values in a form acceptabic to the FLAME 3 code as a function of cumulative burnup, soluble boron concentration, and moderator temperature. 3.2.2.3 K-Infinity Multipliers for Fuel Containing Depleted BP To account for the presence of depleted BP in a fuel assembly, a third K. multiplier is used. As the BP rods in an assembly are depleted,' the boron concentration decreases rapidly. At the end of one cycle of burnup, j the boron concentration in the BP reds may be assumed to be cero. When t these depleted BP rods are inserted in other assemblies, their principal effects are the displacement of water and absorption of neutrons in stainless j steel. Sensitivity calculations have shown that this multiplier is not sig-nificantly affected by enrichment.(10) The only configuration of depleted BP rods used in Cycles 1 through 4 of Surry Units 1 and 2 is 12 rods per assem-Sly. Therefore, this multiplier is generated for an assembly cf average enrichment (i.e., 2.90 w/o U235) containing 12 depleted BP rods. Using the Vepco PDQ07 ' screte Model, a 2.90 w/o enriched assembly is depleted both j with and withou. 12 depleted BP rods. At several burnup points additional 3-11 ,--,->-.v,wy ,y wrv w.-o-y e--, r-y--y-m.v, ,e.,.. =_w-, wme< ~%. .m.m,, r-m.r v-1 w'w-mm -6

cases are perfor=ed for which moderator temperatur: and voluble boren concen-tration are varied simultaneously. The values of the third K. multiplier are obtained by dividing K, with depleted BP by K, without BP. The PICCOLO code is then used to tabulate thete values for input to the FLAME 3 code as a function of total burnup, moderator temperature, and soluble boron concentration. 3.2.2.4 K-Infinity Multiplier for the "Af ter Effect" of Fresh BP The presence of fresh BP rods in a fuel assemoly luring depletion also affects the reactivity in that assembly after the BP rods are removed. This BP "After-effect" arises from the change in isotopics accompanying depletion of the fuel assembly in the harder neutron spectrum resulting from the presence of fresh BP rods. Once the BP rods are removed, the after-effect diminishes with continued depletion of the assembly. To represent this effect, assemblies of various enrichments are depleted in two ways using a quarter assembly representation of the Vepco PDQ07 Discrete Model. First, an assembly is depleted without BP rods. Second, an assembly of the same enrich-ment containing 12 BP rods is depleted to various burnups. Then, the BP rods are removed, and the assembly is depleted further. The values of the fourth K, multiplier for BP are obtained by dividing K, from the assembly which has had BP by K, of the assembly which has not had BP. The PICCOLO code is used to tabulate these values as a function of cumulative burnup experienced by the asse-bly with BP rods present, total burnup, and enrichment. 3.2.3 Method of Calculating Input Data for the Control Rodded Assembly

3. 2. 3.1 Introduction The input dcre for assemblies centaining control rods are generated in a mannet similar to that used to generate the BP data (Section 3.2.2).

Again, a quarter-assembly represer.tation of the Vepco PDQ07 Discrete Model is used to perform all the basic calculations, while the PICCOLO code is 1_19

t used to construct the tables for input to the FLAME 3 code. Four K. multipliers are used to the Vepco FLi.ME Model to represent control rods. Three multipliers represent the presence of control rods in a fuel assembly. The fourth K. cultiplier is used to ac;ount for the reactivity "after-effect" resulting from the depletica of a fuel assembly containing control rods. j i 3.2.3.2 K-Infinity Multipliers for Fuel Containitg Control Rods i The presence of control rods in a fuel assemSly has much the same ef fect as the presence of fresh BP in an assembly. First, a large con-centration of non-fissile, neutron absorbing =aterial is added to the assembly. Second, moderator is displaced from the assembly. The reactivity effects f associated with the presence of control rods depend on initial assembly enrich-ment. cu=ulative burnup of the assembly while the rods are present, total l assembly burnup, soluble boron concentration, and moderator temperature. To represent these effects, two K. =ultipliers are used. The first K. multiplier accounts for the dependence on enrichment, moderator temperature, and soluble boron concentration. To generate the values for this multiplier, calculations are performed for assemblies of yarious anrichments with and 9tithout control rods using the Vepco PDQ07 Discrete Model. These calculations are done at zero burnup and with different, values of soluble boron concentretion and moderator temperature. The values of the first K. multiplier are obtained by raticing K. in the rodded condition to K. in the non-rodded condition. The second K. multiplier becounts for the variation in the eactivity effect of control rods with changes in assembly burnup. The values of the second K. multiplier are generated by depleting assemblies of various enrichments at constant soluble boron concentration and =oderator temperature usi:tg the Vepco PDQ07 Discrete Model. At various burnup points, separate recovery cases 1_12

are performed during which control rods are inserted. To calculate the values of the multiplier, the ratio of K, with the control rods to K, without control i rods at each burnup point is divided by the ratio of K,with control rods to j K. without control rods at zero burnup. The third K. multiplier accounts for the dependence on cumulative l burnup of fuel while control rods are present. The change in isotopics accom - panying depletion of an assembly containing control rods can result in signi-ficant reactivity variations while the control rods are still in the assembly. While the input data for BP include this effect in the first K, multiplier for BP (Section 3.2.2.2), a separate multiplier is used to represent this effect for control rods. To generate the values for this multiplier, assemblies of various enrich =ents are deplete:d with coc. tant moderator temperature and soluble boron concentration in the fo11 cuing manner: j

1) an assembly containing control rods is depleted to various b urnup s,

T

2) at these burnups, the control rods are removed and the assenbly is depleted further,

3) recovery cases are performed at various burnup points in

+ the non-rodded portion of the depletion for which the control rods are reinserted. ? The values of the third K multiplier for control rods are obtained in two steps. First, for a particular combination of cumulative and total burnup, the value of K, with control rods is divided by the valuc of K, without centrol rods. Second, these values are divided by the ratio of K, with and without control rods obtained.at zero cumulative and zero cotal burnup. 3.2.3.3 K-Infinity Multiplier for the "Af ter Effect" of Control Rods The method by which the fourth K, multiplier for control rods.is 4 generated is the sa=c as the method described in Section 3.2.1.4 for the-fourth i 3-14 . ~. - -,y-7--,y wy ...,y ~%. r, vym_,- me,.,._,.pg,,.9 .-.-.,-n--,,, -, err ,-..n m..

.~. - K. multiplier for BP. First, an assembly is depleted without control rods using the Vepco PDQ07 Discrete Model. Second, an assembly of the same enrich-I ment containing control rods is depleted to various burnups. Then the rods are removed, and the assembly is depleted further. The values of the fourth K. multiplier for control rods are obtained by dividing K, from the assembly which has had control rods by K,of the assembly which has not had control f rods present. 1 3.3 Thermal-Hvdraulic Feedback Parameters Thermal-hydraulic feedback effects are represented in the 7epco FLAME Model in. order to more accurately calculate the pcwer distribution. j The thermal-hydraulic feedback model incorporated in the FLAME 3 code is the same as the feedback model used in the PDQ07 code portion of the Vepco j PDQ07 Discrete Model.II) The input required consists of:

1) Coolant inlet enthalpy, i

2) Heated peri =eter per unit area of flow, 3)

Flow area of the fuel assembly per total cross-sectional i~ area of the assembly, 4) System pressure, i

5) Difference between average fuel temperature and moderator tempera-I ture as a function of relative power density.

I The feedback calculation is performed in the following manner.

First, an initial estimate of the fuel and moderator temperatures is made for each coolant channel. Based on this estimate and the function tables supplied, the FLAME 3 code calculates the K-infinity and the migration area for each node. These parameters are then used to calculate the power density in each node. This power distribution is used to compute new fuel and coderator temperatures for each channel.

In turn, the new fuel and =oderator temperatures are used to i 3-15

i calculate new K-infinities and migration areas for another power distribution calculation. This process is continued until the power density for each node th in the Nth iteration differs from the power density in the N-I iteration.bj' less than the specified convergence criterion. 3.4 Xenon Concentration Calculation The Vepco Fill!E Model represents the xenon reactivity effect as a functional dependence of the basic K. en xenon concentration (refer to Section 3.2.1.1). The xenon and iodine concentrations at each node are calculated using the parameters described in Section 3.2.1.3 and the assumption of constant flux and power throughout the duration of a time /burnup step. With the assumption of constant power and flux, the differential equations which describe the dependence of xenon and iodine concentration are solved by integration over the time step interval, At. The nodal iodine concentration, I (t) (atoms / barn-cm), is cal-g culated with Equation 3.1: YI *G1 YI G1 g g I (t) = - I (t-at) exp (- A at) (3.1) g g 7 I I where SP 10" 0 g {^1' core j t=t-at A, YI are given in Section 3.2.1.3. g g P = core power (watts), V = core volume (cm ), S = relative power density q; noc= r, g A = decay constant of Iodine-135. 7 3-16

The nodal xecun concentration, X (t) (atoms / barn-cm), is cal-g culated with Equation 3.2: P (YI +YM ) G1 A YI G1 g g 7 g X (t) = - I (t-at) exp(-A at) g g y G2 G2-A A-7 7 (YI +YX ) G1 g g X (t-at) g G2 A YI G1 7 g - I (t-at)

• exp (-G2 at)

(3.2) g G2-A A I I ~ where G2 = A + o "(t-at) G1, Xe Xe ai a21' I 1 = thermal microscopic absorption cross section c of Xenon-135 (barns), A = decay constant of Xenon-135, x YX, a 21' SIOMt g are given in Section 3.2.1.3, and other para-meters ara as previously defined for Equation 3.1'. s

SECTION 4 - CALCULATIONAL TECHNIQUES 4.1 Fover Distribution Normalization The FLAME 3 code provides two means by which the predicted power distribution may be normalized to the results of either measurements or another calculational model (e.g., Vepco PDQ07 Discrete Mcdel). The'two normalization factors which may be used are 1) leakage parameters (radial and/or axial albedoes) and 2) migration area multipliers. For FLAME 3 calculations, all the nodes modeled are inside the core. The presence of the reflector is represented by leakaga parameters (albedoes) which may be input for peripheral nodes. The migration area multipliar is intended to correct a limi ation i in the FLARE neutron source option. This limitation arises from the inability of the FLARE option to account for spectral effects in assemblies containing Sr;aable poison or control rods. The result of this limitation is that the power may be somewhat underpredicted in these assemblies. The migration area multiplier may also be used to " fine-tune" the radial power distribution of interior located assemblies that do not necessarily contain burnable j poison. 4.1.1 Radial Power Distribution The redial power distribution calculated with the Vepco FLAME Model is normalized to the'~ radial power distribution calculeted with' the Vepco PDQ07 Discrete Model. This normalization is generally accomplished by using a combination of. radial albedoes and migration area multipliers. Radial albedoes are chesen so the,t the desired in-out power sharing and peripheral powers are given by the FLAME 3 code. TPc migration area multipliers are used to " fine tune" the radial paver distribution at interior assembly locations. m N

t I i I The radial power distr.bution is normalized for two sets of condi-tions. The first set of conditions consists of cycle depletion calculations where normalization is performed for the all rods'out (ARO), hot full power (HFP) condi:1on. The design objective is to maintain less than a 5% dif-ference betweer ;he aasemblywise radial powers computed by the Vepco FLAME .and PDQ07 Discrete Models throughout each cycle. The reason for this ST. limit is to insure that acceptable burnup distributions will be calculated during depletion. For "shcrt" (e.g., annual) cycles in which relatively little fresh BP is reed, the normalization f actors for beginning of cycle (BOC) generally yield acceptable radial power distributions throughout the cycle. However for "long" cycles (i.e., 18 month cycles) which contain large numbers of fresh fuel assemblies and BP rods, it is necessary to renormalize the in-out power sharing (1.e., change radial albedoes) at approximately the middle of cycle (MDC) to maintain accepte.ble agreement in radial power distribution. The second set of conditions at which the radial power distribution is normalized is BOC, HZP with control banks C and D inserted in the core. The. primary purpose for this normalization is to find the proper M multiplier for control rodded assemblies. For cases at less than full power and with control rods inserted, the radial albedoes for HFP, ARO provide acceptable peripheral powers so that changing the peripheral albedoes at HZP is usually not required. The acceptance criterion of 5% difference between the Vepco FLAME and ?DQ07 Discrete Nbdels is relaxed for rodded cases in which no depletion is required (e.g., HZP and HFP rod worths), because variations in radial power distribution have been demonstrated to have little affect on the axial information calculated with the FLAME model.( 4-2

d 4.1.2 Axial Power Distribution } Normalization of the axial power distribution calculated by the FLAME model is required only for tero or part power cases of annual cycles containing little fresh BP. In these cases, a generic value of the axial albedo gives close agreement between measured and calculated axial power distributions for HZP. For part power situations, the exial albedo is taken to be a percentage of the RZP value. l 4.2 Differential and Integral Control Rod Worths as a Function of Bank Position Differential and integral control rod worths are calculated with the Vepco FLAME Model in two steps. First, a series of cases is run with the FLAME 3 code in which all reactor parameters are held constant except for the position of the rod bank (s) whose worth is to be determined. For example, if the differential and integral wort hs of control bank U are required, the following set of cases is analyzed: i

1) all rods out 2)

D bank inserted in the top node of the appropriate assemblies [

3) D bank inserted in the top 2 nodes of the appropriate assemblies n+1)

D bank inserted in the top n nodes of the appropriate p assemblies last) D bank fully inserted into the core. The change in core reactivity resulting from each movement of the rod bank (s) ^ is a direct measure of the control rod bank (s) differential worth. The second step in the process of calculating bank worths is normalization to results from the Vepco PDQ07 Discrete Model. The discrete model is the production calculational model for determining total integral control rod worths. There-4-3

l fore, a normalization factor is applied to the results obtained with the FLAME 3 code to assure that the total integral worth predicted by the Vepco FLAME Model is the same as that predicted by the Vepco PDQ07 Discrete-Model. Based on the methodology outlined above,.the following equations are used to compute the red worths: i-1 i 5 Differential Worth at k -k 10 -(4,y) k-1xki SPN

• Node 'i' (pcm/ step) i i

Integral Worth at k -k 5 (4.2)- x 10 x3 Mode 'l' (pcm) i k Xk where k = eigenvalue given by FLAME 3 for the control bank out, k = eigenvalue given by FLAME 3 for the control bank inserted th in the i

node, j

SPN = number of steps of control rod movement per node, i N = total integral worth from Vepco PDQ07 Discrete Model divided-f by total integral worth from Vepco FLAME Model. r b t 4-4

SECTION 5 - RESULTS 5.1 Introduction The. purpose of this section is to demonstrate the predictive capability of the Vepco FLAME Model fot calculations of axial power dis-tributions, axial offsets, and differential and integral control rod worths. As explained in the previous section, the radial power distribution calculated with the Vepco FLAME Model is normalized to a 2-D (x, y) dis-crete power distribution calculated with the Vepco PDQ07 Discrete Model prior to initiating axially dependent calculations. Therefore, this section presents 1) typical radial power distribution comparisons between the Vepco FLAME Model (af ter normalization) and the PDQ07 Discrete Model, and 2) comparisons to measured data taken at the Surry Nuclear Power Stations for-axial power distribution, axial offset, and differential and integral rod worths. The specific types of results compared are presented in Table 5-1. 5.2 Radial Power Distribution As discussed in Section 4.1.1, radial power distributions calculated with the Vepco FLAME Model are normalized to 2-D (x, y) discrete model power distributions calculated at the ARO, HFP condition as a function of b urnup. The power distribution comparisons given in Figures 5-1 through 5-4 for Surry 1, Cycle 1 and in Figures 5-5 through 5-8 for Surry 1, Cycle 4 are representative of the agreement obtained between the Vepco FLAME Model (af ter normalization) and the Vepco PDQ07 Discrete Model. Normally, it is desirable to maintain the assemblywise agreement in radial power distribution to less than a 5% difference, but this "ad=inistra-tive" limit is usually relaxed if violations only occur infrequently in a small number of fuel assemblies. The reason for this limit is to ensure that the accumulated assemblywise burnup obtained during a cycle depletion will be acceptable. This limit may also be relaxed for power distributions associated 5-1

with non-depletion types of calculations (e.g., rod sorths), since it has been demonstrated to have little impact based on sensitivity studies. (11) 5.3 Axial Power Distribution Axial power distribution and axial orfset comparisons between the Vepco FLAME Model and measurements are presented in this section for core conditions ranging from beginning-of-cycle (BOC) to end-of-cycle (EOC) at hot zero power (HZP) and hot full power (HFP). In general, the Vepco Flame Model predictions attempted to simulate the actual core conditions in terms of power level, core burnup, and control rod position. However, for reasons of increased ca".cula-tional efficiency, the Vepco FLAME Model does no: explicitly represent the effect of spacer grids, nor can the control rod position be exactly specified (due to the limitations of the axial mesh spacing). However, the accuracy that is compromised for the increased calculational effi-ciency is not significant. Representative axial power distribution comparisons are presented for Surry 1, Cycle 1, Surry 2, Cycle 3, Surry 1, Cycle 4 2nd Surry 2, Cycle 4. Figures 5-9 and 5-10 give the Surry 1, Cycle 1 core average axial power distribution at BOC, H P for D-Bank partially inserted and l near fully inserted, respectively. The axial power distribution of an individual assembly (B-8) containing a partially inserted control tod (same core conditions as for Figure 5-10) is provided in Figure 5-11. Surry 2, Cycle 3 axial power distribution comparisons are presented in Figures 5-12 snd 5-13 at BOC, HZP conditions for ARO and D-bank inserted. Additionally, comparisons at HFP, ARO at approximately BOC, MOC, and EOC conditions are given in F'igures 5-14 through 5-16. Similar comparisons for Surry 1 and 2 Cycles 4 are provided in Figures 5-17 through 5-20 and 5-21 through 5-23, respectively. Measured and pr'edicted axial 5-2

offset comparisons are also indicated on Figures 5-9 through 5-23, as appropriate. 5.4 Differential And Integral Rod Worths Results of the Vepco FLAME Model predittions and startup physics measurements for differential and integral control rod bank worths for f the first four cycles of Surry Units No. 1 and 2 are presented in this section. It should be noted that Surry 2, Cycle 1 results are not given, since they are essentially identical to Surry 1, Cycle 1. Figures 5-24 through 5-27 provide the Surry 1, Cycle 1 indivi-dual differential and integral bank worths for Banks D through A, respec-tively. Also worths for Banks D through B moving in 100 step overlap are given i.n Figure 5-28. For Cycles 2 and 3 of both Surry Units 1 and 2, only banks D and C control rod worthn were measured during startup physics testing. These results for Surry 1, Cycle 2 and Surry 2, Cycle 3 are compared to Vepco FLAME Model predictions in Figures 5-29 through 5-32. Measurements taken for Banks D through A during the startup of Surry 1, l Cycle 4 are compared with Vepco FLAME Model predictions in Figures 5-33 through 5-36 and in Figure 5-37 for the 100 step overlap mode. Similar-comparisons for Surry 2,. Cycle 4 are displayed in Figutes 5-38 through 5-42. v t 5-3

l l TABLE 5-1

SUMMARY

OF COMPARISONS Reactor Condition At Reference To Tvpe of Comoarison Which Comparision Is Made Figure Radial Power Distribution Unit 1, Cycle 1, HZP, ARO, BOC 5-1 Unit 1, Cycle 1, HFP, ARO, BOC 5-2 Unit 1, Cycle 1, HFP, ARO, MOC 5-3 Unit 1, Cycle 1, HFP, ARO, EOC 5-4 Unit 1, Cycle 4, HZP, ARO, BOC 5-5 Unit 1, Cycle 4, RF?, ARO, BOC 5-6 Unit 1, Cycle 4, HFP, AR0, MOC 5-7 Unit 1, Cycle 4, HFP, ARO, EOC 5-8 Axial Power Distribution Unit 1, Cycle 1, HZP, DO 204,~70 HWD/MTU 5-9 Unit 1, Cycle 1, HZP, D@ 26, 70 MWD /MTU 5-10 Unit 1, Cycle 1, HZP, D@ 26, 70 MWD /NEU 5-11 Assembly B-8 Unit 2, Cycle 3, HZP, DG 202, O MWD /MTU 5-12 Unit 2, Cycle 3, HZP, D in, O MWD /MIU 5-13 Unit 2, Cycle 3, HFP, DO 219,150 MWD /MTU 5-14 Unit 2, Cycle 3, HFP, D@ 221, 3920 MWD /MTU 5-15 Unit 2, Cycle 3, HFP, ARO, 7740 MWD /MTU 5-16 Unit 1, Cycle 4, HZP, D@ 220, O MWD /MTU 5-17 Unit 1, Cycle 4, HFP, D@ 215,150 MWD /MTU 5-18 Unit 1, Cycle 4, HFP, D@ 206, 7715 MWD /MTU 5-19 Unit 1, Cycle 4, HFP, D3 220, 12585 MWD /MTU 5-20 1 Unit 2, Cycle 4, HZP, AR0, O MWD /MTU 5-21 Unit 2, Cycle 4, RFP, ARO, 204 MWD /MIU 5-22 Unit 2, Cycle 4, RFP, DG 210, 6986 MWD /MTU 5-23 Differential and Integral Rod Worthz Unit 1, Cycle 1, HZP, BOC, D-Bank 5-24 l Unit 1, Cycle 1, HZP, BOC, C-Bank 5-25 Unit 1, Cycle 1, HZP, BOC, B-Bank 5-26 Unit 1, Cycle 1, HZP, BOC, A-Bank 5-27 Unit 1, Cycle 1, HZP, BOC, B through -3 5-28 in 100 step overlap mode Unit 1, C ele 2, HZP, BOC, D-Bank 5-29 Unit 1, Cycle 2, HZP, BOC, C-Bank 5-30 Unit 2, Cycle 3, HZP, BOC, D-Bank 5-31 Unit 2, Cycle 3, HZP, BOC, C-Bank 5-32 Unit 1, Cycle 4, HZP, BOC, D-Bank 5-33 Unit 1, Cycle 4, HZP, BOC, C-Bank 5-34 Unit 1, Cycle 4, HZP, BOC, B-Bank 5-35 Unit 1, Cycle 4, HZP, BOC, A-Bank 5-36 5-4

. ~. _ . ~..... I I TABLE 5-1

SUMMARY

OF COMPARISONS (Continued) Reactor dondition At Reference To. Type of Comnarison Which Comparison is Made Figures J 4-Dif ferential and Integral Rod Worths (cont.) Unit l', Cycle 4, HZP, BOC, A through D 5-37 i in 100 step overlap mode Unit 2 Cycle 4, HIP, BOC, D-Bank 5-38 ' Unit 2, Cycle 4, KZP. BOC, C-Bank 5-39 Unit 2, Cycle 4,.HZP, BOC, B-Bank 5-40 Unit 2, Cycle 4, HZP, BOC, A-Bank 5-41 i Unit 2, Cycle 4, H7.P, BOC, A through D .5-42 in 100 step overlap mode 4 4 i S 4 i 1 l 5-5 ew r e.-- ..~,,,-,vm-

FIGURE 5-1 RADIAL POWER DISTRIBUTION COMPARISON FOR SURRY 1, CYCLE 1 HZP, ARO, BOC Gi 09 ~ 10 11 12 - 13 14 15 1.158 H 1.1895 +2.7 l I 1.243 1.144 J 1.2971 1.1708 +4.4 +2.3 J 1.134

1. 214 1.098 1.1550 1.2563 1.1098 i

+1.9 +3.5 +1.1 1 i 1.197 1.097 1.143 0.998 1.2279 1.1031 1.1640 0.9896 g +2.6 +0.6 +1.8 -0.8 a 1 1.084 1.154 1.019 0.990 0.774 1.074St 1.1630 1.0027 0.9941 0.7630 -0.8 +0.8 -1.6 +0.4 +1.4 it l 2 e 1 1.144 1.047 1.047 C.943 0.626

1. 342 1.0208 1.0525 0.9230 0.6109 l

-0.9 -2.5 +0.5 -2.1 -2.4 i l 1 0.995 1.103 0.975 3.625 j p' O.9809 1.0929 0.9580 0.6108 -1.4 -0.9 -1.7 -2.3 I I l 0.872 0.657 R 0.8564 0.6445 ~ -1.8 -1.9 PD107 FLAME 4% 5-6

FICURE 5-2 RADIAL POWER DISTRIBUTION COMPARISON FOR SURRY 1, CYCLE 1, HFP,. ARO, BOC 08 09 10 11 12 13 14 15 l l.1641 i 1.1690 +0.5 1.2625 1.1519 'J 1.2630 1.1558 +1.6 +0.3 l Il 1.1417 1.2158 1.109C 1.1434 1.2323 1.107 n] +0. 2 +1.4 -0.1 pw_ u 1.1982 1.1057 1.1485

1. 01 '. '

( 1.2092 1.1015 1.1589 1.0067 +0.9

-0.4 40.3

-0.7 1.0878 1.1522 1.027e 0.9981 0.7937 l.0751 1.1539 1.0143 1.0113 0.7988 -1.2 +0.1 -1.3 +1.3 +0.6 I 1.1322 1.0413 1.0356 0.9393 0.6414 1.1216 1.0212 1.0499 0.9355 0.6442 -0.8 -1.9 +1.4 -0.4 +0.4 0.9835 1.0783 0.9578 0.6270 P 0.9767 1.0696 0.9453 0.6293 -0.7 -0.8 -1.3 +0.4 .;e 0.8605 0.6526 0.8537 0.6553 -0.8 +0.4 I PDQ07 FLAME h% 5-7

FIGURE 5-3 RADIAL POWER DISTRIBUTION C0" PARIS 0N FOR SURRY 1, CYCLE 1, HFP, ARO, MOC 08 09 10 11 12 13 14 15 L 1.1290 I 1.1330 +0.4 l 1 i p 1.2249 1.1273 1.2288 1.1320 +0.3 +0.4 1: 1.1248 1.2186 1.1155 1.1301 1.2238 1.1219 +0.5 H).4 +0.6 l 1.2095 1.1115 ~. 1901 1.0646 1.2149 1.1174 1.1956 1.0710 +0.6 +0.5 +0.5 +0.6 i ll i is i. 1.0920 1.1773 1.0639 1.0772 1 0.8549 1.0961 1.1801 1.0647 1.0859 0.8639 f' +0.4 +0.2 +0.1 +0.8 +1.1 l.1295 1.0332 1.06i9 0.9825 0.6601 1.1271 1.0323 1.0733 0.9630 0.6571 -0.2 -0.1 +0.8 -2.0 -0.5 l l 0.9288 1.0140 0.8825 0.6152 I' O.9309 0.9963 0.8747 0.6070 i +0.2 -1. 7 -0.9 -0.9 i ll 0.7427 0.5884 0.734A 0.5865 i -0.3 i B PDQ07 FLAME a% a 5-8

FIGURE 5-4 RADIAL POWER DISTRIBUTION COMPARISON FOR SURRY 1, CYCLE 1, HFP, ARD, EOC l 4 08 0? 10 11 12 13 14 15 1.0613 i 1.0585 -0.3

• ' -a, 1.1635

! 1.0659 J 1.1636 1.0626 +0.0 -0.3 J l 1.0695 1.1722 ! 1.0757 ,g 1.0659 1.1710, 1.0714 -0.1 -0.1 -0.4 l t ! 1.1761 1.0763 1.1782 1.0650 1.1745 1.0720 1.1749 1.0621 -0.1 -0.4 -0.3 -0.3 ll Il si si g 1.0717 1.1723 1.0660 1.1193 0.8894 1.0681 1.1692

1. 05r,1 1.1216 0.9005 l

-0.3 -0.3 -0.1 +0.1 +1.2 1.1391 1.0354 1.1003 1.0346 0.6880 1 1.1389 1.0346 1.1064 1.0166 0.6947 -G.0 -0.1 +0.6 -1.7 +1.d 0.9363 1.0323 0.8856 0.6351 P 0.9494 1.0257 0.8871 0.6388 +1.4 -0.6 +0.2 +0.6 0.7410 0.6005 0.7560 0.6155 4 +2.0 +2.5 PDQ07 FLUE _ 4% 5-9

FIGURE 5-5 i l RADIAL POWER DISTRIBUTION COMPARISON FOR SURRY 1, CYCLE 4, HZP, ARO, 30C 08 09 10 11 12 13 14 15 0.783 f 0.7635 -2.- 1.039 1.088 J 1.0 722 1.1422 +3.2 +5.0 0.907 l 1.028 1.096 0.8911 1.0454 1.0651 -1.8 +1.7 -2.8 1.036 1.101 1.083

1. 15 7 i

1.0229 1.1211 1.0873 1.1509 -1.3 +1.8 +0.4 -0.5 gi si -i. 1.113 1.078 1.153 1.108 1.0 29 1.0890 1.0606 1.1539 1.1145 1.0391 -2.2 -1.6 +0.l' +0.6 +1.0 i 0.954 l 1.208 1.078 1.030 0.689 0.9242 ~ 1.1702 1.0577 1.0339 0.6967 - 3.1 -3.1 -1.9 +0.4 +1.0 1.075 1.091 0.967 0.634 P 1.0302 1.1194 0.9837 0.6421 -4.2 +2.6 +1.7 +1.3 0.810 0.627 0.8057 0.6268 -0.5 -0.0 PDQ07 FLAME A% 5-10 )

~ FIGURE 5-6 RADIAL POWER DISTRIBUTION. COMPARISON FOR SURRY 1, CYCLE 4, HFP, ARO, BOC 08 09 10 11 12 13 14 15 1 0.820 d 0.8139 -0.7 7 J 1.067 1.108 1.0962 1.1489 +2.7 +3.7 s 0.930 1.040 1.097 i. 0.9300 1.0616 1.0737 0.0 +2.1 -2.1 1.037

1. 093 1.071 1.132

i, 1.0347 1.1097 1.0785 1.1194

-0.2 +1.5 +0.7 -1.1 16 is is i. 1.103 1.061 1.130 1.085 1.018 "g 1.0886 1.0478 1.1246 1.0837 1.0197 -1.3 -1.2 -0.5 -0.1 +0.2 0.954 1.190 1.068 1.027 0.701 1 0.9419 'I 1.1470 1.0494 1.0242 0.7132 -1.3 -3.6 -1.7 -0.3 +1.7 ' I l 1.082 1.093 0.976 0.652 p 1.0394 1.1087 0.9839 0.6656 -3.9 +1.4 +0.8 +2.1 [ 3 --- 1 l 0.837 0.653 0.8379 0.6604 +0.1 +1.1 'e PDQ07 l \\ FLAME M 5-11 I

FIGURE 5-7 RADIAL PO'4ER DISTRIBUTION COMPARISON FOR SURRY 1, CYCLE 4, HFP, ARO, MOC 08 09 10 t5 12 13 14 15 l 0.937 4 0.9228 -1.5 1 .198 1.233 !.1 J 1.1850 1.2102 -2.9 +1.0 1.037 1.238 1.175 1.0261 1.1947 1.1356 -2.9 -3.5 -3.4 l 1.212 1.212 1.202 1.117 1.1649 1.2233 1.1796 1.1374 -3.9 +0.9 -1.9 +1.8 il il 7: 1.085 1.103 1.106 1.067 0.938 1.0793 1.1046 1.1332 1.0890 0.9728 -0.5 +0.1 +2.5 +2.1 +3.7 0.877' 1.069 1.070 1.001 0.662 1 ! 0.8779 1.0868 1.0641 1.0087 0.6742 I +0.1 +1.7 -0.6 +0.8 +1.8 0.913 0.979 0.868 0.613 p 0.9004 0.9940 0.8812 0.6169 -1.4 +1.5 +1.5 +0.6 1 0.701 0.567 0.7012 0.5640 +0.0 -0.5 PDQ07 FLLE 6% h 5-12

FIGURE 5-8 RADIAL POWER DISTRIBUTION COMPARISON FOR SURRY 1, CYCLE 4, HFP, ARO, EOC 08 09 10 11 12 13 14 15 i 0.9 06 H 0.8966 -1.3 Il 1.179 1.132 I J 1.1291 1.1400 j -4.2 +0.7 1 1.037 1.227 1.140 t 1.0136 1.1757 1.1028 -2.3 -4.2 -3.3 1.231 1.225 1.214 1.100 1.1708 1.2222 1.1638 1.0939 -4.9 -0.2 -4.1 -0.6 il li it 1.078

1. 13 7 1.100 1.083 0.934 l'

I l.0678 1.1179 1.0981 l1.0838 0.9584 -0.9 -1.7 -0.2 +0.1 +2.6 L 0.878 1.054 1.096 1.021 0.669 1 0.8793 1.0604 1.0789 1.0319 0.6932 +0.1 +0.6 -1.6 +1.1 +3.*6 0.899 0.982 0.858 0.621 P 0.9220 1.0189 0.8905 0.6490 +2.6 +3.8 +3.8 +4.5 1 0.696 0.573 0.7454 0.6171 +7.1 +7.7 PDQ07 FLAME A% 5-13

..= Figure 5-9 Core Average Axial Power Distribution Comparison for Surry 1, Cycle 1, BOC, EZP, ARO o o .o. o o. a / .O o e.o .N ) o o.o* to tu. Io= o e-. o. ng ~r c o e r co -8 Z k.1 N O 1 p.g oc N om z o o

c o

C 7 o - c o cm n up "J R' o E n_ a"* 8 C _o .-4 w n en x ""U C C.c = 5 o En22 o. -u o E-T T - -5 a o ca n 5?"5 -m c c < C SL*t OS*t S3* t 00*t SL*O OS*O SZ' O 00* f 83 mod 3AI18738 5-14

1 l FIGURE 5-10 l Core Average Axial Power Distribution Comparison for Surry 1, Cycle 1, BOC, HZP, D Bank In o e f.D. C O e O -I o O C o -S m tu I CJ Z o D e -g z e somes pm n i E o o o H 7 O V 4 (fy On@ O <~ n i n., d 8 J C n an o e-. Y Hk U$ 83 $G'8 8 no a E*23 aba4 Ck st t os t sa t co t st o os o st o co f 83M0d.3AI18738 5-15 .~....

FIGURE 5-11 Assembly B-8 Axial Power Distribution Coc:parison Surry 1, Cycle 1, BOC, HZP, D Bank In e o -5 o e. e o ~ o 3 o LO ta I O Z-h o 9 e E N -8 z 5R"2 z

• e y

e - O I ^ e-o b .o m v e 8 J C e ~C O U. E C %c e 2 o 0Q* o E"5I E"a -5

3 a a. a o

st. t os t sa t co t sc o os o sz o 00 cP M3M0d 3AI18733 5-16

FIGURE 5-12 Core Average Axial Power Distribution Comparison Surry 2, Cycle 3, BOC, HZP, ARO o o .fD w O O. O o .N C9 O (D to I () Z o M o N, Z ~ me<~m V3 OM l N Z I N O o l o l-- = A zeemM o ON L d S J C ^ c^ .O W l i D CM T X h s,a 5 C s mam-e vN e Q Q. v.M 3 k CM .O E * *"' 1 8 =m a. a < Ck SL'I OS*I SZ't 00*I SL'O 09 0 93*0 00* f 83M0d 3AI18738 5-17 y -,. ~,,., .,_4 .,..__,---,,_m -.e-

FIGURE 5-13 Core Average Axial Power Distribution Comparisen Surry 2, Cycle 3, BOC, HZP, D Bank In O o .to. e O* Ov ee O C _N C 3 9e 'w (O tu l I () Z g^v o Q M Q N, 2 I

s o e o m I

W N r.c Z Z O e o H ^ E O d 8 J C s .8. G o x p .: t cr s $d E - 2. % e X5a 9 =we- 'o E*a3 w =3 x = ma< O st/ t os t sa't 00 t st. o os o sa o 00 cP l 83M0d 3AI18738 5 18 l

~ FIGUR$ 5 ! Core Average Axial Power Distribution Comparison Surry 2, Cycle 3, BOC, HFP, ARO o o .t,0 m C 3 C e C e .N ~ t t o Q I C -S e tu I h, Z o c M"8E .g _= =--~ lC s c z O o e C Q e. m H i

4v co o ao m lC o

m i O 'g 4 C M g --m g m 'bC g g m en 9C C e on L .o I* wv 1 W g I wu C w w c E e n. d o w -u e o Nan .o g!"1 e= a. m < C h st t os t sa t 00 t st o os o sz o ' 00 - (f 83M0d 3AI18738 5-19

l, NM "YC n na 5 ' y,. NWa eO4r,NEaO8oD5 a 2mn w- @S" e M8- $. gO o c.o h t 0 0 o 4 1 o o.a h i o o. o b i S E H R C N o Q)( f I o. 0 D b N E0011 e R902 - I U212 [ S3 N AE O M o l o. I T lf ) 6 b I x s ( 0082 S E901 - O M212 A3 P L D F o L o. A b I ) n) 4 5~ X U o% T i( A M t t / ie D ss W of M o M)Pf o. (% O ~ p(k urnl b neaa 2 rwBi uo x BPDA 0 0 ? om. a o ~ mCe 0*O 'd .$z w>~FEJwE yO l t(, 1 li l 1 l l

,~ aM ue$ 92 i.lr -h~ .F1 Oe$nw rOU nQuNd8a O D ( na

• n r" ".

n* x o o c.o 's t 0 0 g 0 4'1 0 0 0 2 't 0 0 m-0 T. 0 't S E k. H m k C ) N x o Q( ^ 4 c. I D E0082 o R402 - 'e N U712 I S7 AE N M O o I 6 c. T )x 'o I ( 0083 s E402 - S M712 A7 O L P F 0 L ) n) 0 A U o% 'o I 4 T i( X M t t R / i e D ss k W of M) Pf 0 ( % O p(k 0 urnl neaa o '2 r wBi uo x BPDA 0 0 w N.~ ~ om mu O-mQo @.'O mN.O o9 % J>*~&rIJ* gW3O LL L c L u,N l .l 4 .i i i! i ;

i FIGUR5 5-17 Core Average Axial Power Distribution Comparison Surry 1, Cycle 4, BOC, HZP, ARO o9 O _e O 9 C C 9 O .N C9e -S e ta=uz a D o e Se<eA -8 z x NN l3 N a z s o e o H ^ N _o %' O < CO b D . :d M N O 5 o_ a J C 9 8^ -E ~ s e z ua N W @ a em i! s t N 8 l w a l C.v M o 3 64 cM g C C g@B"1 =m e. m < C9 st t Os t sz t 00 t sc o os o 93 0 00 d' 83M0d 3AI18738 2 5-22 l

FIGURE 5-18 Core Average Axial Power Distribution ~ Comparison for Surry 1, Cycle 4, BOC, HFP, ARO o9e .w e o o we S o .w e o eo m ta= U z v o jf ~ m9 o. m SceeJ z i Mcc-L m O m *=* N mg z C z e o l l o 9 5 n' .5 [ v e, = = ~ m til e o - I e6 o z--s 1 0. C a J o. C .o ~ n en D o 42 X Y EE C s -o e ea E,SU o' ww o it 'd - cona N umme 3 C M ecc< o9 SL* t OS*t SU't 00*t SL*O OS*O SZ'O 00* @ 83M0d 2AI1873S 5-23

e FIGURE 3-19 Core Average Axial Power Distribution Comparison for Surry 1, Cycle 4. SiOC, HFP, ARO o o.o .r.o. eo o -r e o.o ,N ) o c. o .o U; La I o' u z O^ '(x o. e e =.ae o e e, .o c 2: -Co l 7 DN M N W mN d .c z x

8.. a

~ n o E c e m r4 -8 =a - C e i u) ZNmN Q $^ c. i '

n.

8 ) er n an .o 3 ON w w X 3: ua c s y = mx 3 O% U .nrw a vn o o. c v.:d uc-o r n.n. u e am. w 3 o x mmc < .,7 o$ SL'I 09'i S3'I 00*I SL'O OS*O S3 0 00*f i M3M0d 2AI.18738 5-24 t s k

iG FIGURE 5-20 Core Average Axial Power Distribution Comparison for 3erry 1, Cycle 4, EOC, HFP, ARO i O C. O .m. O O ese O O ( .N. l O c ~O g to o I U $v Z e Q W c M^CC- "o Z X 03 C N m 3 -N o 4-a Z l C C ]: O ] l== o v c c co m lC _O Mmo-w h O r 1 0 C CC n an .O e-o b hU n' ct = ma o w -* e L C tn to O a vH C O

c. v.2 i

3 6 c-

.O c 3 c c3 N 3=ee< O O. SL*t OS*t SE*t 00*t SL*O OS*O SZ'O 00*f 83M0d 3AI18738 e 5-25

1 FIGURE 5-21 Core Aversge Axial Power Distribution Comparison for Surry 2 Cycle 4, BOC, HZP, ARO o9e .o o 4o e4 o o.o N o i o.o r[ -S w ta u A,, y r e 4 o M S!o<=A .a nm D N W j z O o m an a m LeacA a~ y a mm g m m Q p* Q g g 8g C

?

-o s zy x t C a am 3 ou = - a. - tts ~ 3 ue~ g a Esa2 .o 3 O N =a=A oo st.* t 09 t sz t co t suo os o sa o ao f 83M0d 3AI18738 5-26

~ i FIGURE S-22 Core Average Axial Power Distribution Comparison for Surry 2, Cycle 4, BOC, HFP, ARO E ~ o o .o o 4.- /' e o O, o .ca e o c "e D tu I m u . 'E S

a e o e N moon

_o 2 gN-N o aC E z 2 8 S ~4er. ldCCN 1 .o m k D e -g a x a. E $C H -v .o k mN X g ga a-5G"o

c. v u o

3 Le cM o ( 0 (11 e O

• J3m-

.o Qk o st t os t sa t co t st o Os o sa o cotf 83M0d 3AI18738 5-27 4

FIGURE 5-23 Core Average Axial Power Distribution Cociparison for Surry 2, Cycle 4, MOC, HFP, ARO e I e .W O e~ we O3o .N. C C "b (O w n I v u bcooE Z o o %%C* I Q W wemn a* -8 z a .c m 2 5ec-A o o cc m o,-* i o e-a ]e H .O"N .g ax coa c_ m e-B 3" 8 a Ei 1 -o a d 2 O% X =^'8 1

c. v.:d. e

3 6 c C 1J Q Q o

W 3 23

• o 52a4

_f N O O st t os t sc t oo t sc o as o sz o' 00 5 83M0d 3AI18738 5-28

FIGURE 5-24 Control Rod Worth Comparison for Surry 1, Cycle 1, BOC, HZP, D Bank 16

ii- ^:::E::.i:m= F.= =!======== ; 4 1600

=..

~ :- ~ V 7:~ I==n , = - - - =. = EE9:4= =~ Tpi 11=iE=E JE =- = =-4 ~

== -- =. n...c.i. -...j.=. 7..hF,=.'__; =. =a =.. - - . ~. - - - FLAME t 14 - f ~~" ' ~ f -~ =}1525 f:_ :--- _g_ 1 URF.D - ~~=.NEi . iH N J i 1400 y _ g. ;.=g.g ip= W h f=:f \\ ;=.:Ju:i=? IEEii4jE.1Er -f at:: =.i c rm "isli~~= M V IEss ' li-EE~:!!==i =!=55:ti"~ :Mn "H .EEEi:Edid EE u\\ g=~ f==1 iij4=ii-[i 3d={p.i... -. 2 _g ;= g4g g=_4 g zg; g tj==_3;g==. j=3 = ; 1200 g = -E:=-[i===E, 'lX \\E5= l - E: g4 - y=i = jes-EE=E EEEcpsE n liffi--!~ -I' $f d\\ EEPi n T!M42=6i_i'--i.~+J:f1-4:F ~ b =r _ g = t - 2. g. 4 y_.33.g = :- =. p_, ; g

a - =p = =2 g

=; ':iE=t'=lj__; 1000g h --l-EE.f- : =ilt :~=Pi \\ ~IE W ii F=Ei!:i+E \\ =

===F==i[ lie!E- \\\\l= T!!Elg., JEEinEiiiE=i r=Z= i?'_i;=}i _.= - =!#-l ~ =% EhEM\\=E=iiisM =t=--!: =c!5Mt!5 si& =i g = =

== i=_== 1, :EE! -- - \\\\==:j =p -iEi==iEdis.i==Ei.E 800 $

3 TL--jEE5fElEE+ t' \\::- 5 -EEi l 'UMI5lEi52!5_ Eis=~ 3 3

s

+E=i=:= /@%f:;.:gJf \\\\= F;=E g =i=i =i 9t.= =- j - ; j b ZE---lEE:d.==--!- 9?Ml'i~ T=IIl4\\ kE9EEEEi-E"- ~'~ 5 i=.---[p_. _ r; 7- :j="i \\\\+:= =1= :] q =h= = i =%-. 600 y 6 a2 - - -- /. F=== t==i... _. \\\\. .. s 1.{1 = =r:- -c i - g 3 j"_= EL 8 i'EEiil ? "il=E= "!\\\\i R Z-Zi ' d !' it -9u-

L

i= /l hed=f=-i-ir.=

: \\\\=i==:i=4 b ~="-+ ~ ~~"=+

-==t--dl FZ:Z{= Zr_T--i =-$===;i=-E i.Qs= i=E"' = i '-- ! 5 4 00 M4 /jI---IF.5I"-f[14.5EE -h=_il:5S5-id "htI':' :. _ _ Y t....f =.. =.3. =_ g g __ m..._.t...._l.\\\\ i...._ g =_. -.1-- 2 =; z t / f.., -. _... _...-.,_.-._.t. = N (p - _o==_ =__3 d - :. h. ! "E YIWEE-- r---- -h-t =E---Ck"EIE:"E~liDE~ "i? 2 200

==/f==t=__=r- _. y _ __ L; N y -.. _ 3.m. - : =__ a = ,=;=,

_ a_

t- _._..3 ~~f E=-k ~fb t __ \\

= /E t=T:t-el E i f b e =~Ji-- Zi= = =! = iEEi a =;f i m i'

JEv==-=l==-t= = =J -izr3 j = ;== 14._ A 0 ~ 0 0 40 80 120 160 200 228 BANK P0HTDN (STEPS) 5-29

FIGUPI 5-25 Control P.od Worth Comparison for Surrf 1, Cycle 1, 30C, HZP, C Bank 14 1400

g=~

g-a y= \\. - - - FLAME Z ~-- 5 5 ,.__k._ _J$--)N-N. _ _ _ 1200 EM~ i 12 h(' _2fb5 T (r-tg-yk i--j \\g.- _:= 10 5\\=0 4 \\* ' ~ 1000 _ _E E \\.- % =_. c, e m 5 Mi s\\ k-E O ji- _g\\ {\\ 800 5 ~ M =8 5 ~ h, _. _k k C._g\\_ _ _. j } \\_=_ w. - I rA g Q t- \\-- m 11 - 3.-- g _._C h\\- ' O__-._- h I Z w -] \\ ~ 6j y g gy 6* s T .. _i= f \\Q-s ' Q_~~~ 4 N 400 6. s 9 m w -- - = //

• A T=L 2

if EV-200

==//

w %= l,'.-~ '~ w -NN 's =._= W g + 0 o 0 40 80 120 160 200 228 BANK POSITION (STEPS) 5-30

... b,,'. .g FIGURE 5-26 =.* 4 .,y. ~ -tf;c i,. Control Rod Worth Comparison i for Surry 1, Cycle 1, BOC, HZF L Bank I 3 s - 4 2 2 h __ 7._ 2200 20 T2"~- 2000

3. ~ ~

t_..__. %. - +. _._1 ___pg ,m 9_f. = T-4 I-EASURED r-i_ A ^^ & =% k{ )$ ~ bl6 ~ _.~ 1600 4 3 gy 3 p

a;;=yr

-a f., , + 3 c f, is -c Le = 5 12 \\' ? 1200 E 3 /- 5 's a 1 5 ,7 \\\\ . :5_(

=

- - g-r-. ~T

a g,

= ~~~ a: k_._. c. o N 8, 800 T j g 4 ol 1 t t f. \\ s f-; -\\ '( ---.hl ik k-f;.' ^ 400 -{9g/ _ --,=: v [/----/ +g,# "k__ Q 7, .__j ,( _j +-- 0-- 0 6 - 0 40 80 120 150 200 228 ?' BAb POSITION (STEPS) ~ + 5-31 [ y n. y

FIGURE 5-27 Centrol Rod Worth Comparison for Surry 1, Cycle 1, BOC, HZP, A Bank 1500 13. L A. ~ 1900 14 = %-4 _o [=K E =\\ \\=[.- _ m-e\\ =; .-l_=TQ_ - - - rtAxE gg MEASURED --- 1200 12 ~ hl_ k ^ ~ J!=A 1 {' '\\. ] , _m i g T-1000 10 ..h k,E\\ N. ~~~ c e, _--p,_ = p l 'Cl ( -j=_g j = -n T=. c w i^-I v 800 8 f Y-! 'g "%\\- E x c ! 9.I e ih \\

1

~ H p v E 56 ,I_ .\\ 1 600 $ d I, ~ f '\\ h r \\ L n 4 -JI: A T ~ 400- [=!L ^K K 74 t th - 1. w .s- ~\\ W i Ar! \\ ![ - 3'A 200 2 /.) N k1-+--- l g A g: y/; N wi

d S ;=> -

0 0 0 40 80 120 160 200 228 BANK POSITION (STEPS) =_w

) pI 4-i' 't 4" l j 3200 16 l ,l . g I s L.. j \\ - - - FLAME j j N e l HEASURED n {12 's h, }- - f g g 2400 u i e u .m =l h a t; r I! i . a1 . 'i_ n 'y/7 I e e e =nne: l p i m gs a y 1 r h 0 08 - h8 <ej 1 - l 1600 - g .I Y E ?" $ ' 2 fr Y ~ I, \\ d auo no t3 - { h . w .-I . i Q Ona$* SNii l 3 E[rj h \\ j. ll ET " S N 800

• E'd 4

t t E 5 l-l [ I i l' tl pili.i.% hhi +' l l l t i 4.M 0 O -- " 150 180 210 228 n Bank d 30 50 90 150 150 150 21O 228 C Ilanh 0 30 60 90 120 150 180 210 228 D Bank BAllK POSITION (STEPS) ~ ' ~ ~ ' ' ' ^ ' ' ^ '~' Y.- . { _ Y.). L 'l +... 3,>.,, [ '. "[ 1 p f. 7 _ t ^ 1 .j : - -;4

• g..

4

n. g

..y 4

+

V FIGURE 5-29 Control Rod Worth Cor::parison for Surry 1, Cycle 2, BOC, HZP, D Bank 12 1200 3 i i.. - ~ ~ ; R., q :. =__. j -. . =.._: _ I. 10 __. : IM.-_I- __ - -. _. [ .-__--..E__ 1000 N ! i.. - - - FLAME ]- _.y__....t. 3 MEASURED ,s.. .) i e 5 3 g g t - i. CI" ~ ! ~

• .,..s"

'm i 9 - ~ _..=__.:.._....s. _. _ _... -. \\ l- \\ \\ g t :.. _.. j - _ _; _ 800 M_ .. ;J.__. u -l ' _. - l-g J:i < N i-c g

gy

_._,3 ; _.._ _ _.;., e /-

j.. gg

.] ' ' 4 p

.__&:2 J :

_= _d .l' T '. Ll_a : -g._h _..y._:_._._._.....

N s.;.g.j;

} j; 5 .j ll. /:i [ g 600 m x6 /.. .. - -- V.. !. .-._ d _ru=_ _j - ; /... -.;r j

aj-2 o

m.. '

a i ,l-,. p:__ J: s-d.- _.-_ s u._ ;. _. 3 q_.. - 4 t N' 2 cf.m. .t __ s 'C* - \\-- $.ll H g f _.. ; - -,L . c, i... z H w o N;- = e ~ f /- w i 400 x4 1-m N ,I..._._,4...s' .l N ._....,___../__!__,- c i l s i, i e .I i .t

i.,

I f t. s. g. i l l'. N I

_; L.../

a..u; 4 r... .L_.2.._i .i 1, I: /. r i i 9..

--i-- ;

-\\,' 200 2 /-, . ; /, t _;._ ].i_.' _ j_ ::{ j;_.;_.j..l. '.N ._ \\... i } i /. l l

i. l l

t i I \\ \\. y _.% ~ l 'I l l t. _..._.,\\ 7,.. !...'.. -..i t _. __.. _ 9._. _s. ..g I s,,,, t I e 0 0 0 40 80 120 160 200 228. E BANK POSITION (STEPS) 5-34

FIGURE 5-30 Cc-trol Rod Worth Comparison for Surry Cycle 2, BOC, HZP, d Bank ~ 1!:00 14 _ !._ _ q.. _. _.. h.....,... X-- i l ]. 1 3 -l., F j .i.... i .rn-* .t N.

c 1200 12

- s.i i i.... p .g. I;__..'_-_ L _ ;- N4 _ _!. _ - ;;. nLO!. i.;_, S[ u-E

2- "

l - \\ p _.._-..__--..g. l-n- - - - FLAltg

\\

'.1 MEASURED 1000 10 t ..: : e

t:.- :--- * -- ---

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FIGURE 5-31 Control Rod Worth Comparison for Surry 2, Cycle 3, BOC, HZP, D Bank 1200 12 1 . -. :.l ... L-.-. .l i ....:......n.. ..:.....--.......-.-.-......----..L. 3 = - =.= % f 4 .g 10 K -. 1..-- /1100 - - - FLAE i . l- .; :s' .r.. - -- ' EASM ...[., l 1 I ..,,.{ f l' 1. j

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IF ( R FIGURE 5-32 Control Rod Worth Comparison for Surry 2, Cycle 3, BOC, HZP, C Bank 13 1300 6 j. 12 .... _Qy = f; l yl r _n_ _';. : -2 1200 = _:.. J =~ ._: _ z.. g.. ._L.g..--

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Figure 5-33 Control Rod Worth' Comparison f or Surry 1, Cycle 4, BOC, HZP, D 3ank 12 1200 f

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1 1 Figure 5-34 Control Rod Worth Ccmparison for Surry 1, Cycle 4, EOC, HZP, C Bank 12 1200 '.__~;.. ..i .[.- . ' /."'. ; - j r "M. .l..--"-J_--*.1..r.,.,- ~ - - - - - * - ' ~.'.d......= : _ '.j.' _... ...I . s....

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Figure 5-35 Control Rod Worth Co=parison for Surry 1, Cycle 4, BOC, HZP, B Bank 20 2000 7!-- ..; t : :!R ! *2:_.. ' -- ' '. ^ - 2fd fi -~ C ;'i 2'ri F. I :[.~~ ' ; -! _%.":2 i E .. E : i l i - u.." } - ._......_..:[. [ J . 5 ! - ;k.5b.t.b _..b.. b.'.c. -~ N..__..3. .h db....*.. f.. L L. b

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Figure 5-36 Centrol Rod Wortn Comparison for Surry 1, Cycle 4, BOC, HZP, A Bank 2000 20 .t .: _:., : --.i_.. . U - ' -- I _' ..-~t .~2T' ..T I I 1. -- ' - f. ;.. : ! _.r..-. t-t - -- 9 .. _ _ _. _. _ _,,, ~ ..... _.....,. _..... _ _.... _ _ _._._.._...,..__.,...J., _ 4,. ) _. _......s..._.. i l,u m... .. m 1600 .._......_..y....._..,......_. % ::r 2,......_

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Figure 5-38 Control Rod b' orth Comparison for Surry 2, Cycle 4, BOC, HZP, D Bank 1200 12 . _ _ _.__,----r__ - - _ - - - - - + -

---

x---------=._-._

- =.. _ -- -- ~~ g. =:__. _. m --- -. . = - - - -- -r _ s u ---t =:- 2 :-- - ^ . _ _ _ _ - - _ _. _ _ _ = 10 _E

==

1. ===

1000 f___.r----:-Y_._~;\\R -- J - t-~ - =~ --~= - - - ME E~ E~~-+'i~5 3- __.i_w_9._e-_ p 3 ~~~ -. l-..[ 7 ,-e- _.%-,..I _4 b__ 7 . _ _ _ _., _ _ e!---- ^ k,r ' f. ?b._----._=__. E3 _q_.- /=:--- \\ r. i t-g c f ^ =. =.. =w e ~ g.

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• l [ :- *-

$\\=A I'_ a w _\\ \\.~. : - ;:_ .x-t. N _ ';\\ ----- - ---- ;\\.. _ y_ W_ __ --- C- ,__,___a - : /,I----+ \\_1. a M _ = n _ __..__ _ _ _,_,,., p ^ .z . -- _.. - l s -~\\ = 7 L=._.K - 2 ..3.._.___.. 400 _f \\_s. g Sa 3 -x-_._-_---_=_ g=-

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a ~ 200 -/ ^ ~ .-Q. - k =.. - _. 1. _ _..... ' y .. 21 '-'"~ - /.- -i r- +-

__ Y

--- f ... j-+ - - - - - + 1/ - r 0 .= -..=-.r 0 0 40 80 120 160 200 228 BANK POSITION (STEPS) 5-43 % m

'I E Figure 5-39 Control Rod Worth Comparison for Su m/ 2, Cycle 4, BOC, HZP, C Bank 1000 10 < __.__.__n=2= -__=__: .s-N. 1 s _:zN_ .dx=~ . _ _ = -~~ 800 8 =Mgg@=E __ ws _ g p g q -- g v.r.aSuazo - - - a o.; _ .qt y /:=_ i + 1 - r:- K

'm 5

1= '6 600 - S 6 gggf \\ gg. q 1 =1 x -m ~ = e =y +C f r f* X \\\\ N-5 j i 'N ~ = 3 fC_ [_C \\; k.' _ o3 ~ +- N Y - ~ _ N 3 j ,%El g 4 _i x = x; -- g_ - 400 ;a g =jr s N -3_: -N- . c. g K-AT %=eh-- ~ g / =' e ____ _._ CT-_=; N =g =_ \\ -- - - -

a.

/ 1 N-{ s- - - - r /7

-N--s

\\__t - 2; 200 . gf_ g ._ygs m m.___.; _._3 1== w _=v ~ -- - - _ w.t. rr , w. 0 40 80 120 160 200 228 BANK POSITION (STEPS) m s 5-44

ligure 5-40 Control Rod Worth Comparison for Surry 2, Cycle 4, BOC, EZP, B Bank 18 1800 = -t- =1 r== = t== t= = Ez.5 ai 4 =+.= ;-E:; ;;.; _i_z:.j== -p=-E---pt =+i==. E_.-====. c u : 1- +_- .%Ni!=M_--==-I T=#I~_-5$=E -Ib' = ~i ' i 2i-~='- = Ei=>= r==E~11.._ ~ r rs=-E==f nr===i:= '=+=P

i=-ti--

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=

~~ N ~ = --Pi=_ =\\+/ I h-E \\4W=F =M" --- FLAME ~ 1400 14

EEEhi-E M 1 l1 L # M =~J = =f-E k MEASURED r

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===t- = P l-\\T=d - =- %- =. -- - -

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E==I f.__..._ _a=\\= =g; \\22 = V=

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==9 ", g' _ _=. _ __==r==j= = =\\,- A7,.._. -N== t. rd-.. __.- _ l = p- - t r- _ N; -- =..=..--a ,1 _.. \\-..... r --- w _,_ (= i T- ' 400 4 - w %x.+== t_.. =a.=. Q. ; \\==r-

=_..=.. \\'==,f= ---

t ._a_.__ +_= =x

==v =v =- t= - _ n. -- E elf /i 3EEEEt= 7 j'=-~~L= _M' 5 $ i\\'ic);;~'i'~7 M M ~ x gfy,:.g.r_ _._ -. - _.. _. p _. __.__.. _q= =r ug. p ; .._.:n =.-%N=m--t=J: \\W Eila'= ~ -. --.u 200 2 4 - =p. ---,---- : -- -- - :_h.v. - --_ _. -..__.,_ _g_=u 1 + .. t-s g _.. _ ; _ _.. _._ ..___p_,.._q_.- / t _.. _a - u _t= -. j n =- i _..... N 7._. [ =g ....g

• -$' ^' ^ 2 ' = = g= = + = :

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Figure 5-41 Control Rod Worth Comparison for Surry 2, Cycle 4, BOC, HZP, A Bank 17 1700 .g y _.. gg ,;. __. g_ _;.

==q= p.=a. ;u f. p g 7 __ _ a== g _ ;. g ;;= =1.;. _. -- 1600 16 ~-~ ~ ~ -'-~- - - - - ~ ~ M=4= l i-9 \\ EiElE4 5 L=+ ===t= L i L +=4*i - - _.-..W:qs fe+5a q==sj===;=ss:- - 4=i =i=s=g 3.- Ai= b z-is= di =ris-i= wate= A== ie=:=. lo=-!- = =4 =.= :\\: =' }=.3 = t- _tu x=. + sis E=t== = = + i-es = - =ih.: :=G_. r~ ~ : n.. 1400 J _: T :: - --- N,b =:r.~: 14 - i-e .+ = nm

~~:-!?=iiNW "i Li #1-i i~ i"i-i=i=i 1E.ASURED jft EliEiE i~ - ti=~iM *: sis-i'li: 9="F

==-",= r-- -- ~._..lh=E-E j = if '. !_T;?_sli-i=i!'E xi'=;'-03i_ua;i = r_'=.i 'EkbhE. hildI# \\1 = .=E+k"-H=--ib '-hib=!=_ ~ - -

-E p

a = p = = F- "7.+.Q =i _+.\\ Wi 1711=i= izzg=E =-?;EEmi=i::_2p m b =i=R-E$1![ _.c Li=ih \\ F4! E =--i==i=i ilri" : h, c_ . i ;==4 yr:== :;g13_.; _3.;;;; d .: t r=.== 4 L.=a. r-- 1000 - 0 10 ii=it"tij =i= L 2 ' ',= ""- k \\= *-

== E='= iE=?: - F-i:

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i'=3 31fK=:h 31@-\\E --r:- - -- !=%i!==-i-E:

U-3 h 3 8 ' = "--I i "IN'\\5f ='hI' \\~ ~~'" =- I ! ~ " E~ -5'" 5 800 _~~'"!.--g isihii!:%\\l.=d!-f_ OdM lEEWir-=.-il =J i~ s y ._ :. 13-4 l ---, - Es--2:f =iN= g n= \\t E" q=--tE

= =.g

7 -

2 S 35= 1EET[E,' s-~= 1== -\\i \\Qdi=t= C=i=?iE=Ji=i= a=- 2 M i=Iti=.% =_W i N= H N N +=F+, g6 .,_y_j __ j_, =g;_ : j=g;;g.\\g; \\j = __ .i\\_ , _ ; _= 600 =iE~i 7 !2l =iE-EEF'-i E!"_:-\\ \\.-iE4= N _ Ni 4 i i

_ 2 f

. j us:t - - - i=. - -t =_1k.. \\_;. _. _.

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p___,

'"2._.-- 400 4 =g f 1.. _.. _ ___ u__ _ _ m_ r-- \\ \\ =. -- -_., y+\\

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_-iz /.s_.u,t as;-- rE s_st a=__=_=__=__=ga-

_m_\\_ \\;__.A. =_.. :_.1\\ _;. _.

__r___ 4 =E'Y !I=~-I = i -- -- I -- f T ~~~'~~*'Y='I' ' ~ =I 200 = 2 n==. _. t__.-. - __ r r--- =_= an = __ z _- <==.-/ :-- ; -_ 5 i t-. Yg !=--- t :: l

}

( m. - -/ : g-t: ___.____-.._c-" ,__..-.._y-_-y.. +_- =- m---_- t.. _ _. f :--- =. 1 [/.:bb- - -.___. b 2:$b. _ bb5I-~ ~b-i~dINi;. Y H_ b _-__;;_._.=______._y_.-._;__. ~ ~ - - - ~ = - ~; ~ ~ ~ ~ ~ ~ ~ ~ " 0 0 0 - 40 80 120 160 200 228 BANK POSITION (STEPS) 5-46 m .w

Figure 5-42 I Control Rod Worth Comparison for Surry 2, Cycle 4, 30C, HZP, Banks A - D moving in 100 Step Overlap x 1 C INTEGRAL WORTH (PCM) 5 C C C C C C O O C C C C C C C C C 3 i,t1 4 -M N CC eN I. _ - ;_ __ _ ; u 1 N ^-?_ _. 4 r_. '. -^ is je=-i~ =stseet - __;===. rnp: :, u e

=4

g

g. [rc._ r4;- ;= ;=f: -- (=:_: {j--] 1.. ~ - j,-g ff ).((.-

1-

=2"! ~Js

@ "= 5 ii s ?-! M = i = E--i E :8 / #i" * / ~ ] ,O ~ u ---ME =

E-t # "= rul~1_J- =2 Fr e / ~"-tii

//= =EEj==:j =, = t ~ "= = 1_ _E} = : j :Qu-i.7. =;i.[/.2ic_ r._. n Z' y

-s-;==

1 g-;ic. ; 4 4 - g gi a ei: ef==. = = M -i=r= I =.GiE=E-iMev ~=4 e li:,'/m-ti=: - ll e = = bi-s s ! EE-15 =;- =gEt=. T = r p j;i.p = =:----

• 3

Ei=s=:,Li: b=istEI -T=EBi=Eni

-iE-! r_a -:G .e G - m:j _,2.; = n- + n_ m:_ _.;g : g-g,]7- - 1 p _ ;:. r _ : ; g .__..-.._. _ _ ~.__ ,m P 9*#.. -, [i -- ---@E= r 21 ~ -~=- ; i. _ _ _ _. _. __d ZE-I_r_-- i t _. t_ _.. N e 23-t-zH#. _~ T===iEE=iEE'EW i =/==vc eFE==i==i= = 515kEIf:---I;Ef.fE.2=!E

1. ~ [- 5 1:!'E~E-5 ' rI~~ 5 I

b --f-_. _-l:._ l j

a. _. _

r. rr r.... ;_ r. _-_ __. - _.__. 2. n__a- ---

.g_ t=n x t.. - -._i. _=.. =_ [ r- -_ m ,4 e +- e e M m T It r_:r-fi~--9.f~1=~ !b =I5H f ^ ~~ 2i"5 :- c ~ ~ i~E li=s-~E :.~ i EEE=~!: h /N)"_- E"E"i" E-iEifE g =. =-. ;.-. _, =y _ p...-. _,.. ,= ~ n a b

b. -f.

-. - _ bbb. ~ amp ._-*t..C*- L~- 7-~ , am ^ i s

--[rf=d.it

'[ f- ~ ~12..._$..i.. '--- t ' '- f 3 :n,---j,' [2 !.- ^b [-+iEi J!+_.Q.3 f ~==it==2f:ErifMiiM!=i=5 J1=iFJi += / E-e =._=n_t: _ ::_ ::;r /_ ;.rf m:tu_u= --

==a =+. u.: ~! = :r : y; n_;.- -- o 1 5N-----$-! Vik ~ ~~ $~ 'E-.: NP_25-i;5 N:i_I@ C 2EE=! EJ-f/_t M M.E_~1j M 2=F"=t=Eisss- ' iE !E=h=9'rtr st= '_. e ii 21 = en==E!==i-:=;--- .C ,_ru ?.,,, :- n_ co

---

 ; f r/_,.j =r.r-t -_ _. __ r = __.2_, w r j.mn= 1 - =. :.. _ _ r= 2. = 1 =r r1 n... _..

-=:

==: :- 2;--/I?_;-r4-A -.-_..I..-_; _T P = r - -- 1 -. m =.,.t = :r_. --,/.. i= - : : -. r. ~ <a----- _.___.-~"--T.'_-~._~~4_._q_-__-__i.-__.__.=QN_f,_,, }:_ -3_. _~;-'- {. _'_~...~__~ I. -. l };; i...

}= =_il-J E = [~-~~~ E '~

C f l=E~1 l--~ ' --~}~--_.q_ p = "O C N CC 4 C N .e (d31S/H3d) H130M TfIIN3H333IG 5-47

Section 6 - Summary and Conclusions The Vepco FLAME Model is cperational at Vepco for the purpose of performing three-dimensional reactor physics analyses and supporting the evaluation of core performance. The model consists of the FLAME 3 code with ths NULIF, PDQ07, FLAFIT, EDITQAR, PICCOLO, and FLMSHUFL codes being used to providw either input or data manipulation. The accuracy of the Vepco FLAME Model has been established through extensive ecm-parision of calculations witn measursments frem the Surry Units No. 1 and 2 over eight cycles of operation. The results of these comparisiens indicate that the Vepco FLAME Model (which includes normalization to the Vepco PDQ07 Discrete Model) provides the capability to predict axial peaking factors a.d axial power distributicns as well as differential and integral control l rod worths. The comparisiens also indicate an acceptable capability to predict axial offset trends as a function of changing reactor core con-ditions. verification, as well as improvements to the Vepco FLAME Model, vill continue to be made as more experience is obtained through the application of the model to the uaits at the Surry and North Anna Nuclear Power Stations. t I 6-1

SECTION 7 - REFERENCES

1. W. A. Wittkopf, et. al., "NULIF - Neutron Spectru: Generator, f

Few Group Constant Calculator, and Fuel Depletion Code," BAW-10115, June 1976 (Babcock and Wilcox).

2. H. H. Ecssan, et. al., " Babcock and Wilcox Version of PDQ07 -

User's Manual," BAW-19117P, Dece=ber 1975 (Babcock and "ilcox).

3. C. W. Mays and M. Fartney, " FLAME 3 - A Thr#e-Di=ensional Nodal Code for Calculating Core Reactivity and Power Distributions,"

BAW-10124A,.^.ugust 1976 (Babcock and Wilcox).

4. Private correspondence from the Babcoci: and Wilco:: Company to the Virginia Electric and Power Company dated February 3, 1971, and October 6; 1971.

5. C. B. Franklin, "EDITQAR and PICCOLO - Auxiliary Fortran Computer Codes for Use with FLAME 3," NFE Technical Report No. 27, April 1977 (Virginia Electric and Power Co.).

6. W. C. Beck, "FLMSHUFL - An Auxiliary Fortran Computer Code for Use with FLAME 3," NFE Technical Report No. 55, October 1977 (Virginia E12ctric and Power Co.).

7. M. L. Smith, "The PDQ07 Discrete Model," VEP-FRD-19, July 1976 (Virginia Electric and Power Co.).

8. Final Safety Analysis Report - Surry Pre 9-Station Units 1 and 2, Virginia Electric and Power Company, Decencer 1969.

9. D. L. Delp, et. al., " FLARE - A Three-D'=ensional Boiling Water Reactor Si=ulator," GEAP-4598, July 1964 (General Electric).

10. C. B. Franklin, " Sensitivity Study of K-Infinity Multipliers for Depleted BP for the FLAME Model," NFE Calculational Note PM-4, July 1978 (Virginia Electric and Power Co.).

11. W. C. Beck, " Sensitivity of Control Rod Worth Shapes to Variation in Radial Power Distribution," NFE Calc"It tianal Note PM-3, July 1978 (Virginia Electric and Power Co.).

t 6 .) 7-1 ___}}