ML17320A945

DC Cook Unit 2,Cycle 5 Sar.

ML17320A945
Person / Time
Site:	Cook
Issue date:	10/24/1983
From:	Skogen F, Williamson H, Wimpy P SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML17320A942	List: ML17320A941 ML17320A944 ML17320A945 ML17320A946 ML17334A516
References
XN-NF-83-85, NUDOCS 8403080220
Download: ML17320A945 (38)

Text

XN-NF-83-85 Issue Date: 10/24/83 D. C. COOK UNIT 2, CYCLE 5 SAFETY ANALYSIS REPORT Written by:

P. D. Wi , Engineer PWR Neutronics Reviewed by:

F. B. Skogen, Manager

.PWR Neutronics

~ ~ /

Prepared by:

H. E. Williamson, Manager Neutronics and Fue Management Prepared by: - Ps R. B. Stout, Manager Licensing and Safety Engineering Approved by: (g/W4 P)

G. J. Busselman, Manager Fuel Design Approved by:

G. A. Sofer., Manager Fuel Engineering an Technical Services Concurred by:

J. 1P Morg'an, Manager Proposals and Customer Services Engineering csk E@CZM NUCLEAR VVMPARV,lac.

8403080220 840302 PDR ADQCK P.

050003ih PDR

NUCLEAR REGULATORY COMMISSION DISCLAIMER IMPORTANT NOTiCE REGARDINQ CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY This technical report was derived through research and development programs sponsored by Exxon Nuclear Company, Inc. It is being sub-mined by Exxon Nuclear to the USNRC as part of a technical contri-bution to facilitate safety analyses by licensees of the USNRC which utilize Exxon Nuclear fabricated reload fuel or other technical services provided by Exxon Nuclear for licht water power reactors and it is true and conect to the best of Exxon Nuclear's knowledge, information, and belief. The information contained herein may be used by the USNRC in its review of this report, and by licensees or applicants before the USNRC which are customers of Exxon Nuclear in their demonstration of compliance with the USNRC's reguladons.

Without derogating from the foregoing, neither Exxon Nuclear nor any person acting nn its behalf:

A. Makes any warranty, express or implied, with respect to the accuracy, completeness. or usefulness of the infor-mation contained in this document, or that the use of any information, apparatus, method, or process disclosed in this document will not infringe privately owned rights; or B. Assumes any liabilities with respect to the use of, or for dan ages resulting from the use of, any information, ap.

paratus, method, or process disclosed in this document.

XN- NF- FOO, 766

XN-NF-83-85 TABLE OF CONTENTS Section ~Pa e

1.0 INTRODUCTION

2.0

SUMMARY

. ~ ~ 2 3.0 OPERATING HISTORY OF THE REFERENCE CYCLE. . 4 4.0 GENERAL DESCRIPTION . . 7 5.0 FUEL SYSTEM DESIGN. . 12 6.0 NUCLEAR CORE DESIGN .  ; 13 6.1 PHYSICS CHARACTERISTICS. 14 6.1. 1. Power Distribution Considerations . . 15 6.1.2 Control Rod Reactivity Requirements . . 16 6.1.3 Moderator Temperature Coefficient Considerations. . . 17 6;2 ANALYTICAL METHODOLOGY . . 17 7.0 THERMAL-HYDRAULIC DESIGN ANALYSIS . . 23 8.0 ACCIDENT AND TRANSIENT ANALYSES . . 24

8. 1 PLANT TRANSIENT ANALYSIS . . 24 8.2 ECCS ANALYSIS. . 24 8.3 ROD EJECTION ANALYSIS. . 24

9.0 REFERENCES

. ~ 28

XN-NF-83-85 LIST OF TABLES Table ~Pa e 4.1 0. C. Cook Unit 2, Principal Characteristics for Nuclear Analysis of Cycle 5 Fuel . ~ ~ ~ 9 6.1 0. C. Cook Unit 2 Neutronics Characteristics of Cycle 5 Compared with Cycle 4 Data . . . . . . . . . . . . . . . . . . . 18 6.2 0. C. Cook Unit 2 Control Rod Shutdown Margins and Requirements of Cycle 5 Compared to Cycle 4. . . . . . . . . . . 19 8.1 0. C. Cook Unit 2 Cycle 5, Ejected Rod Analysis, HFP . . . . . . 26 8.2 0. C. Cook Unit 2 Cycle 5, Ejected Rod Analysis, HZP . . . . . . 27 LIST OF FIGURES

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3. 1 0. C. Cook Unit 2, Cycle 4 Boron Letdown Curve . 5 3.2 0. C. Cook Unit 2 Cycle 4, Power Distribution Comparison to Map 204-46, 1005 Power, Bank 0 9220 Steps, 7752 MWD/MT. 6 4.1 0. C. Cook Unit 2, Cycle 5 Full Core Loading Pattern . 10 4.2 0. C. Cook Unit 2, Cycle 5 Loading Pattern and BOC Exposure Distribution. 11

6. 1 0. C. Cook Unit 2, Cycle 5 Boron Letdown Curve . 20 6.2 D. C. Cook Unit 2, Cycle 5 Relative Power Distribution 100 MWD/MT, 1149 ppm, 3411 MWt, ARO. 21 6.3 D. C. Cook Unit 2, Cycle 5 Relative Power Distribution 17,900 MWD/MT, 10 ppm, 3411 MWt, ARO . 22

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XN-NF-83-85 D. C. COOK UNIT 2 CYCLE 5 SAFETY ANALYSIS REPORT PROLOGUE This report is the fourth in a series of five reports which address the neutronics characteristics of the Cycle 5 core and provides the safety evaluation for Cycle 5. Preliminary analyses were performed in response to the Tentative Scheduled Delivery Date (TSDD) notice and were provided in letter report PWR:41:82. Subsequently, a final reload was established in response to the Final Scheduled Delivery Date (FSDD) notice and was documented in letter report PWR:04:83. The Fuel Cycle Design Report (XN-NF-83-75(P)), which provides the Reference Design for the safety evalu-ation was issued in September, 1983. This Safety Analysis Report will be followed by a Cycle 5 Startup and Operations Report.

XN-NF-83-85

0. C. COOK UNIT 2 CYCLE 5 SAFETY ANALYSIS REPORT 1.0 INTROOUCTION The results of the Safety Analysis for Cycle 5 of the 0. C. Cook Unit 2 nuclear plant are presented in this report. The topics addressed include operating history of, the reference cycle, power distribution considerations, control rod reactivity requirements, temperature co-efficient considerations, and control rod ejection accident analysis.

XN-NF-83-85 2.0

SUMMARY

The 0. C. Cook Unit 2 nuclear plant is scheduled to operate in Cycle 5 beginning in April of 1984 with ninety-two (92) fresh assemblies (Reload Batch XN-2) supplied by Exxon Nuclear Company (ENC). The composition of the core during Cycle 5 will be ninety-two (92) fresh ENC assemblies in Region 7, seventy-two (72) once-burnt ENC assemblies in Region 6, and twenty-nine (29) twice-burnt Westinghouse assemblies in Region 5. The Cycle 5 design alsa utilizes 1,040 fresh A1203-B4C burnable absorber rods, each containing 0.026 gm/in of B-10. The burnable absorber rods are distributed among seventy-two (72) of the fresh assemblies.

The characteristics of the fuel and the reloaded core are in con-formance with existing Technical Specification limits regarding shutdown margin provisions and thermal limits. The ENC fuel design'is presented in Reference 1. The Plant Transient Analysis, the thermal-hydraulic analysis, and the LOCA-ECCS analysis will be presented under separate cover. The results of the Control Rod Ejection Analysis are provided herein and are derived from a combination of the generic parameters and results described in Reference 2 and specific analyses performed for Cycle 5.

The neutronics characteristics of Cycle 5 are similar to those of Cycle 4. The minimum excess shutdown margin above that required for safe operation is calculated to be 721 pcm at EOC. A postulated control rod ejection event is conservatively calculated to result in an energy deposition of less than 170 cal/gm.

XN-NF-83-85 N

At hot full power equilibrium xenon conditions, the peak F is calculated to be 1.64 and occurs at BOC in an assembly supplied by ENC.

N The peak F

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for Westinghouse (W) supplied fuel is calculated to be 1.40 at hot full power equilibrium xenon conditions, and also occurs at BOC5.

Including a 3X engineering factor, a 5X measurement uncertainty, K(Z) considerations, and an 11% POC-II allowance (for a +5K target band on T

axial flux difference), the total peaking factor, F , during Cycle 5 is calculated to be 1.97 in ENC supplied fuel and 1.68 in Westinghouse N

supplied fuel. The maximum relative pin power, F<H, during the cycle is calculated to be 1.38 in ENC supplied fuel and 1.16 in Westinghouse supplied fuel and occurs at 15,000 MWD/MT, and 500 MWD/MT, respectively.

T Throughout the cycle, both F and F<H are expected to remain within the allowable limits which will be defined by transient and accident analyses and presented under separate cover.

-4 XN-NF-83-85 3.0 OPERATING HISTORY OF THE REFERENCE CYCLE D. C. Cook Unit 2 Cycle 4 has been chosen as the reference cycle with respect to Cycle 5 due to the close resemblance of the neutronic characteristics between these two cycles. The Cycle 4 operations began in January, 1983, and as of the end of September, 1983, the core has accrued about 9,000 MWD/MT exposure. The Cycle 4 core loading consisted of one hundred twenty one (121) Westinghouse assemblies and seventy-two (72)

ENC assemblies.

The measured power peaking factors at hot-full-power, equilibrium xenon conditions, have remained below the Technical Specification limits T

throughout Cycle 4. The total peaking factor, F , and the radial pin N

peaking factors, F<H, have remained below 2.04 and 1.49, respectively. The Cycle 4 operation has typically been rod free with the 0 control rod bank positioned in the range of 218 to 225 steps, 228 steps being fully

- withdrawn. It is anticipated that similar control rod bank insertions will be used in Cycle 5 operations..

The cri'tical boron concentration as calculated by ENC for Cycle 4 has agreed to within about 30 ppm with the measured values (see Figure 3.1).

Also the power distribution calculated by ENC has generally agreed to within +5 percent of the measured values (see Figure 3.2 for a comparison at 7,752 MWD/MT).

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10.0 g CYCLE EXPOSURE (GAD/I1T)

Figure 3.1 D.C. Cook Un) t 2, Cycle 4 Boron Letdown Curve

XN-NF-83-85' G E 0 ,C

.856 .985 .975 1.046 .971 1.069 1.008 .901

..848 .966 .964 1.042 .983 1.086 1.019 .859

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.968 1.064 1'.186 1.069 1.206 1.002 1.125 .737

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.974 1.220 1.081 1.073 1.104 1.234 1.020 .862

.972 1.187 1.079 1.080 1.123 1.243 1.039 .835

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1.049 1.076 1.095 1.093 1.246 1.024 1.107 .554 1.041 1.065 1.072 1.099 1.249 1.058 1.126 .564

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.970 1.219 1.105 1.246 .990 1.175 ;758

.983 1.196 1.118 1.240 1.030 1.195 .766

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1. 068 .980 1.227 1.023 1.173 1.019 .396 1.059 .986 1.221 1.051 1.190 1.031 .401 W.9 -0. 6 +0. 5 2~7

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-1.2 -1.2 1.007 1.124 .999 1.102 .755 .395 Calculated (XTGPWR) 1.010 1.109 1.030 1.105 .757 .395 Measured Assembly Power

-0.3 +1.4 -3.0 -0.3 -0.3 0.0 C-M M

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.903 .748 .857 .551

.852 .735 .822 .558 Calculated Measured X Oi ff.

+6.0 +1.8 '+4. 3 -1.3 N 1.354 1.343 +0.8 F 1.565 1.557 +0.5 q

Figure 3.2 O.C. Cook Unit 2 Cycle 4, Power Oistribution Comparison to Map 204-46, 100K Power, Bank 0 8220 Steps, 7,752 MWO/MT

XN-NF-83-85 4.0 GENERAL DESCRIPTION The D. C. Cook Unit 2 reactor consists of one hundred ninety three (193) assemblies, each having a 17x17 fuel rod array. Each assembly contains two hundred sixty four (264) fuel rods, twenty-four (24) RCC guide tubes, and one (1) instrumentation tube. The fuel rods consist of slightly enriched U02 pellets inserted into zircaloy tubes. The RCC guide tubes and the instrumentation tube are also made of zircaloy. Each ENC assembly contains eight zircaloy spacers with Inconel springs; seven of the spacers are located within the active fuel region.

The Cycle 5 loading pattern is shown in Figure 4. 1 with assemblies identified by their Cycle 4 location and Fabrication ID. The fresh fuel is not assigned a Fabrication ID but the burnable absorber configuration is noted. The initial enrichment of the various regions are listed in Table 4.1. The calculated BOC5 exposures, based on an EOC4 exposure of 13,400 MWD/MT, are shown >n a quarter core representation in Figure 4.2 along with the quarter core fuel shuffle simulation. The core consists of ninety-two (92) fresh ENC assemblies at an average enrichment of 3.64 w/o U-235, seventy-two (72) once-burnt ENC assemblies, and twenty-nine (29) twice-burnt Westinghouse assemblies. A low radial leakage fuel management plan has been developed and results in the scatter-loading of the fresh fuel t

throughout the core with the fresh assemblies in the core interior containing A1203-84C burnable absorber rods. The exposed fuel is also scatter-loaded in the center in a manner to control the power peaking. The

XN-NF-83-85 Al203 84C burnable absorber rods contain 0.026 gm/in of 8-10 and 1,040 of these rods are distributed among seventy-two (72) fresh assemblies loaded in the core interior. Pertinent fuel assembly parameters for the Cycle 5 are depicted in Table 4.1.

'uel

XN-NF-83-85 Table 4. 1 O.C Cook Unit 2, Principal Characteristics for Nuclear Analysis of Cycle 5 Fuel

~Re ion 5 ~Re ion 6 ~Re ion 7 Nominal Enrichment (w/o) 3.40 3. 65 3. 64 Nominal Density (X TO) 95 94 94

'3030 Pellet 00 (in) .3225 .3030 Clad OD (in) .374 . 360 . 360 Diametral Gap (in) .0065 .0070 .0070 Clad Thickness (in) .0225 .0250 .0250 Rod Pitch (in) .496 .496 .496 Spacer Material Inconel Bi-Metallic Bi-Metallic Fuel Supplier ENC ENC Fuel Stack Height Nominal (in) 144 144 144 Number of Assemblies 29 72 92 Regionwise Loading (MTU) 13.286 29.077 37.154 Exposure (MWO/MT)

BOC5 24,069 16,368 0 EOC5 34,866 35,410 19,546 Incremental 10,797 19,042 19,546

XN-NF-83-85 R P N M L K J H G F E 0 C B A M2 02 R47 R8 J7 F15 013 Jl M13 K15 H3 R92 S03 S27 R73 S13 501 R46 N8 L2 L4 Hl E4 E2 G7 R19 S21 S51 S06 S46 S57 R70 A10 P7 K3 J12 F3 B7 R10 S10 S45 S66 S37 532 S23 S08 P4 P5 J6 E15 L15 G6 85 B4 R78 S63 S48 R6 R65 S41 S61 R23 C12 N6 K7 G4 F7 C6 N12 519 530 S31 S52 S35 553 S17 M5 All' d J2 N13 G2 Rll 05 "+

S28 R37 S39 S54 S42 R89 S43 R7 R8 07 M9 C13 J15 N3 09 M7 A9 R4 S07 S72 S34 558 R54 S20. S15 S49 R9 Mll A5 J14 C3 G14 R5 011 S65 R2 S68 S56 S69 R57 S70 C4 N10 K9 G12 F9 C10 N4 512 S25 S64 S22 S38 S33 S50 P12 P11 J10 E1 Ll G10 B11 B12 R60 S60 S47 R33 R81 526 S62 R62 A6 P9 K13 J4 F13 B9 b R6 502 529 S36 S71 S44 S40 S09 J9 L14 L12 H15 E12 c E14 'B Rl S59 S67 S24 S14 S55 R52 H13 Fl 03 G15 M3 Kl G9 R42 S11 518 R36 S16 S04 R71 M14 014 Previous Core Location R49 R3 Fabrication IO

+ Fresh Fuel Assembly, No BA Pins, a Fresh Fuel Assembly, 4 BA Pins b Fresh Fuel Assembly, 12 BA Pins c Fresh Fuel Assembly, 16 BA Pins d Fresh Fuel Assembly, 20 BA Pins figure 4. 1 O.C. Cook Unit 2, Cycle 5 Full Core Loading Pattern

XN-NF-83-85 E

G15 C13 09 09 AS 90 270 180 A9'3,843 24,353 16,089 17,611 17,973 12,945 0 C13 G14 All 011 90 180 13,487 16,177 19,856 18, 183 G12 F.9 C10 C12

'80 17,973 17,865 17,906 17,142 E15 G10 811 812 180 19,790 17,883 0 16,105 23,028 G12 F13 89 A10 180 180 17,611 '0 17,776 16,265 12,404 H15 E12 E14 CS v

12,972 18,190 15,989 0 0 28,483 G15 013 F15 G9 180 180 24,023 17,109 12,295 30,235 014 Core Location in Previous Cycle Rotation (degrees) 23,083 Assembly .Average Exposure (MWD/NT)

+ Fresh Fuel Assembly, No BA Pins a Fresh Fuel Assembly, 4 BA Pins b Fresh Fuel Assembly, 12 BA Pins c Fresh Fuel Assembly, 16 BA Pins d Fresh Fuel Assembly, 20 BA Pins Figure 4.2 D. C. Cook Unit 2, Cycle 5, Loading Pattern and BOC Exposure Distribution

XN-NF-83-85 5.0 FUEL SYSTEM OESIGN A description of the Exxon Nuclear supplied fuel design and design methods is contained in Reference 1. This fuel has been specifically designed to be compatible with the resident fuel supplied by Westinghouse.

XN-NF-83-85 6.0 NUCLEAR CORE DESIGN The neutronic characteristics of the projected Cycle 5 core are similar to those of the Cycle 4 core (see Section 6.1).

The nuclear design bases for the Cycle 5 core are as follows:

1. The design shall permit operation within the Technical Specifi-cation for D. C. Cook Unit 2 nuclear plant.

2. The length of Cycle 5 shall be determined on the basis of a Cycle 4 energy of 1133.2 GWD (13,400 MWD/MT exposure).

3. The Cycle 5 loading pattern shall be designed to achieve power distributions and control rod reactivity worths according to the following constraints:

T N a) The peak F and the peak F<H shall not exceed the Technical

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Specification limits in any single ENC fuel rod through the cycle,.under nominal full power operating conditions.

b) The scram worth of all rods minus the most reactive rod shall exceed BOC and EOC shutdown requirements.

The neutronic design methods utilized to ensure the above re-quirements are consistent with those described in References 3, 4, and 5.

The Cycle 5 loading contains 1,040 A1203-B4C burnable absorber rods distributed among seventy-two (72) of the ninety-two (92) fresh ENC supplied assemblies. In sixteen (16) of these assemblies there are twenty (20) burnable absorber rods per assembly. Another thirty-six (36)

'N-NF-83-85 assemblies will each contain sixteen (16) A1203-84C rods, eight (8) assemblies will each contain twelve (12) A1203-84C rods, and twelve (12) assemblies will each contain four (4) A1203-84C rods. The A1203 84C burnable absorber rods each contain 0.026 gm/in of 8-10. The core loading pattern has been designed to achieve a desirable power distribution while maximizing the benefit of assemblies with burnable absorbers to reduce the beginning of cycle (BOC) boron concentration. The BOC worth of the 1,040 A1203-84C absorber rods is calculated to be equivalent to the worth of 717 ppm soluble boron.

6.1 PHYSICS CHARACTERISTICS The neutronics characteristics of the Cycle 5 core are compared with those of Cycle 4 and are presented in Table 6.1. The data presented in the table indicates the neutronic similarity between Cycles 4 and 5.

The reactivity coefficients of the Cycle 5 core are bounded by the coefficients used in the safety analysis. The safety analysis for Cycle 5 is applicable for Cycle 4 burnup of +1,000 MWD/MT and -1,000 MWD/MT about the nominal burnup of 13,400 MWO/MT.

The boron letdown curve for Cycle 5 is shown in Figure 6. 1. The BOC5 xenon free critical boron concentration is calculated to be 1,491 ppm.

At 100 MWO/MT, equilibrium xenon, the critical boron concentration is 1,149 ppm. The Cycle 5 length is projected to be 17,900 MWO/MT at a core power of 3411 MWt with 10 ppm soluble boron remaining.

XN-NF-83-85

6. 1. 1 Power Distribution Considerations Representative calculated power maps for Cycle 5 are shown in Figures 6.2 and 6.3 for BOC, (equilbrium xenon), and EOC con-ditions, respectively. The power distributions were obtained from a three-dimensional quarter core XTG (6) model with moderator density and Doppler feedback effects incorporated. As shown in Figure 6.2, for the design Cycle 5 loading pattern, the calculated BOC, hot-full-power, N

equilibrium xenon nuclear power peaking factors, F

~, and F<H are 1.64, and N

1.32, respectively. At EOC conditions the corresponding values of F and N

F<H are 1.54 and 1.37, respectively for the limiting first cycle fuel. The N

BOC, HFP, equilibrium xenon F value of 1.64 is compared to the measured

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Cycle 4 value of 1.59 in Table 6.1.

N At hot full power, equilibrium conditions, the peak F during the cycle is calculated to be 1.64. Including a 3X.engineering factor, a 5X measurement uncertainty, K(Z) considerations, and an 11K allowance for PDC-II, (for a +5% target band on axial flux difference) the T N expected total peak, F

~, is 1.97. The maximum relative pin power, F<H, is T N calculated to be 1.38 at 15,000 MWD/MT. Both F and F<H are expected to

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remain within the allowable limits throughout the cycle.

The control of the core power distribution is accom-plished by following the procedures for "Exxon Nuclear Power Distribution Control for Pressurized Water Reactors Phase II"( ' . The results reported in those documents provide the means for projecting the maximum

XN-NF-83-85 F (Z) distribution anticipated during operation under the PDC-II Q

procedure taking into account the incore measured equilibrium power distribution data. A comparison of this distribution with the Technical Specification limit curve assures that the Technical Specification limit will not be exceeded while operating with the PDC-II procedures. The T

PDC-II'documents describe the maximum possible variation in F Q(Z) which can occur during operation when following the outlined procedures. The "T

bounding variation in F Q(Z) represents the maximum variation when the, axial offset is maintained within the allowable range.

6.1.2 Control Rod Reactivit Re uirements Detailed calculations of shutdown margins for Cycle 5 are compared with Cycle 4 data in Table 6.2. The D. C. Cook Unit 2 nuclear plant Technical Specifications require a minimum required shutdown margin of 1600 pcm at BOC and EOC. The Cycle 5 analysis indicates excess shutdown margin of 1,008 pcm at BOC and 721 at the EOC. The Cycle 4 analysis indicated an excess shutdown margin of 722 pcm at BOC and 734 pcm at EOC.

The reactivity allowance for control rod insertion and power, defect at BOC and EOC conservatively bound the most adverse combination of power level and rod insertion to the power dependent insertion limit.

The control rod groups and insertion limits for Cycle 5 will remain unchanged from Cycle 4. With these limits the nominal worth of the control bank, D-Bank, inserted to the insertion limits at HFP is 149

XN-NF-83-85 pcm at BOC and 272 pcm at EOC. The control rod shutdown requirements allow for a HFP D-Bank insertion equivalent to 400 pcm and 500 pcm at BOC and EOC, respectively.

6.1.3 Moderator Tem erature Coefficient Considerations The Technical Specifications require that the moderator temperature coefficient be less than or equal to +5 pcm/ 0 F below 70K of 0

rated power and less than or equal to 0 pcm/ F at or above 70K power. The HZP, ARO moderator temperature coefficient is calculated to be +3.0+2.

pcm/ F and meets the Technical Specification limit below 705 power. The moderator temperature coefficient at or above 70K rated power is cal-culated to be less than 0 pcm/ F and also meets the Technical Specifi-cations.

6.2 ANALYTICAL METHODOLOGY The methods used in the Cycle 5 core analysis are described in References 3, 4, and 5. In summary, the reference neutronic design of the reload core performed using the (6) reactor analysis was XTG simulator code. The input isotopics data were based on quarter core depletion calculations performed for Cycle 4 using the XTG code. The fuel shuffling between cycles was accounted for in the calculations.

N Calculated values of F~ and F<H were determined with the XTG reactor model. The calculational thermal-hydraulic feedback 'and axial exposure distribution effects on power shapes, rod worths, and cycle lifetime are explicitly included in the analysis.

XN-NF-83-85 Table 6.1 D.C. Cook Unit 2, Neutronics Characteristics of Cycle 5 Compared With Cycle 4 Oata C cle4 C cle5 BOC EOC BOC EOC Critical Boron HFP, ARO, Eq. Xenon (ppm) 989(b) ]0(b) 1 149 10 HZP, ARO, No Xenon (ppm) 1,465 ( a) -------- 1,569 Moderator Temperature Coefficient HFP, (pcm/oF) 4 0 (b) -27.5(b) -2.1 -26. 3 HZP, (pcm/oF) -0.97(') -21 9( ) +3 0 -21.1 Isothermal Temperature Coefficient HFP, (pcm/oF) -5.4 (b) -29.2(b). -3.4 -27.8 HZP, (pcm/oF) -2.S6(a) -23.6(b) +1.3 -23.0 Ooppler Coefficient (pcm/oF) -1.4 -1. 6 -1.3 -1.5 Boron Worth, (pcm/ppm)

HFP -7.7 (b) -S.7 (b) -S.O -9. 6 HZP -S.95(a) -1O.9(b) -9.4 -11.7 Total Nuclear Peaking Factor N

F

~, HFP, Equilibrium Xenon 1.59 (a) 1.55 (b) 1.64 1.54 Oelayed Neutron Fraction .0057 .0051 .0062 .0051 Control Rod Worth of All Rods In Minus Most Reactive Rod, HZP, (pcm) 5,525 6,093 6,301 6,172 Excess Shutdown Margin, (pcm)(c) 722 734 1,008 721 (a) Measured data (b) ENC calcul ated (c) Shutdown margin evaluation based on the most adverse combination of power level and rod insertion

XN-NF-83-85 Table 6.2 D.C. Cook Unit 2, Control Rod Shutdown Margins and Requirements of Cycle 5 Compared to Cycle 4 C cle 4 Cycle 5 BOC EOC, BOC EOC Control Rod Worth (HZP), cm All Rods Inserted (ARI) 6,348 6,888 7,065 7,279 ARI Less Most Reactive (N-1) 5,525 6,093 6,065 6,079 N-1 Less 10K Allowance

[(N-l)*.9)] 4,972 5,484 5,458 5,471 Reactivit Insertion, cm(a)

Power Oefect (Moderator+Oopplar) 400 500 400 500 Flux Redistribution 600 600 600 600 Void 50 50 50 50 Sum of the Above Three 1,050 1,150 1,050 1,150 Rod Insertion Allowance 1,600 2,000 1,800 2,000 Total Requirements 2,650 3,150 2,850 3,150 Shutdown Margin (N-l)*.9-Total Requirements 2 322 2,334 2,608 2,321 Required Shutdown Margin 1600( b) 1600(b) 1600(b) 1600(b)

Excess Shutdown Margin 722 734 1,008 721 (a) The reactivity insertion allowance assumes the most adverse combination of power level and rod insertion. The BOC shutdown margin is increased at HFP conditions and the EOC shutdowm margin remains unaffected at HFP conditions.

(b) Technical Specification limit.

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~ P

• P I ~ I I ~ ~

I ~ ~

l ~ ~

• ~

I" ".

o 1000 I~ ~ I~ ~ I P

I~ I I

~

~ I:

800 C

O O P 600 ill O

I I

l 400 200 I

~ I I I I I 0

0 2000 4000 6000 8000 10000 12000 14000 16000 18000 Cycle Exposure (NWD/HT)

Figure 6.1 0. C. Cook Unit 2, Cycle 5, Boron Letdown Curve

I XN-NF-83-85 1.049 1.169 1.119 1.1'59 1.136 1.177 .955 .993 1.206 1.164 1.042 1.175 1.112 1.098 .982 1.116 1.110 1.099 1.154 1.102 1.148 .856 1.160 1.044 1.156 1.090 1.084 1.072 1.054 .409 1.141 1.106 1.085 1.050 .681

1. 177 1. 113 1.151 1.075 1.050 .886 .316

.954 1.099 1.023 1.056 .682 .309 Assembly Relative Power

.993 .982 .857 .410 Peak Assembly = 1.206 (H9)

Pin F~ = 1.323 (H9)

N Peak F~ = 1.644 (G15)

Figure 6.2 0. C. Cook Unit 2, Cycle 5, Relative Power Oistribution, 100 MWD/MT, 1149 ppm, 3411 MWt, ARO

XN-NF-83-85 C, B

.904 1.002 1.075 1.233 1.072 1.035 .894, .841 1.025 1.054 1.216 1.094 1.221 1.036 . 1.077 .850 1.073 1.217 1.119 1.259 1.108 1.190 .954 .777 1.233 1.095 1.260 1.127 1.235 1.057 .997 .434 1.075 1.222 1.110 1.235';160 I:122 .732 1.035 1.036 1.190 1.058 1.121 .955 .396

.893 1.077 .954 .997 .732 .387 Assembly Relative Power

.841 .850 .777 .434 Peak Assembly = 1.260 (Fll)

Pin F~H

= 1.369 (Fll)

Peak F = 1.536 (fll)

Figure 6.3 D. C. Cook Unit 2, Cycle 5, Relative Power Distribution, 17;900 MWD/MT, 10 ppm, 3411 MWD/MT, ARO,

XN-NF-83-85 7.0 THERMAL-HYORAULIC OESIGN ANALYSIS Thermal-hydraulic design analyses for ENC fuel that is being placed in O. C. Cook Unit 2 for this cycle will be provided under separate cover.

XN-NF-83-85 8.0 ACCIDENT AND TRANSIENT ANALYSES 8.1 PLANT TRANSIENT ANALYSIS Plant transient analyses for the ENC fuel that is being placed in D. C. Cook Unit 2 this cycle will be provided under separate cover.

8.2 ECCS ANALYSIS The LOCA-ECCS analysis for ENC fuel at D. C. Cook Unit 2 will be provided under separate cover.

8.3 ROD EJECTION ANALYSIS A Control Rod Ejection Accident is defined as the mechanical failure of a control rod mechanism pressure housing, resulting in the ejection of a Rod Cluster Assembly (RCCA) and drive shaft. The consequence of this mechanical failure is a rapid reactivity insertion together with an adverse core power distribution, possibly leading to localized fuel rod damage.

The rod ejection accident has been evaluated with the procedures developed in the ENC Generic Rod Ejection Analysis . The ejected rod worths and hot pellet peaking factors were calculated, using the XTG code.

No credit was taken for the power flattening effects of Doppler or moderator feedback in the calculation of ejected rod worths or resultant peaking factors. The calculations made for Cycle 5 using a full core XTGPWR model were two-dimensional with appropriate axial buckling cor-

XN-NF-83-85 T

rection. The total peaking factor, F~, were determined as the product of radial peaking (as calculated using XTG) and a conservative axial peaking factor. The pellet energy deposition resulting from an ejected rod was conservatively evaluated explicitly for BOC and EOC conditions. The HFP pellet energy deposited was calculated to be 161.9 cal/gm at BOC and 159.2 cal/gm at EOC. The HZP pellet energy deposition was calculated to be less than 55 cal/gm for both BOC and EOC conditions. The rod ejection accident was found to result in an energy deposition of less than the 280 cal/gm limit as stated in Regulatory Guide 1.77. The significant parameters for the analyses, along with the results, are summarized in Tables 8.1 and 8.2.

Table 8.1 D. C. Cook Unit 2 Cycle 5, Ejected Rod Analysis, HFP BOC EOC Contribution(>) to Contribution(a) to Energy Deposition, Energy Deposition, Value (cal/ m) Value cal/ m)

A. Initial Fuel Enthaply (cal/gm) 66. 5 68."2 B. Generic Initial Fuel Enthalpy (cal/gm) 40.& 40.8 C. Delta Initial Fuel Enthalpy (cal/gm) 25.7 25.7 27.4 27.4 D. Maximum Control Rod Worth (pcm) 179 130 194 143 E. Doppler Coefficient (pcm/oF) -1.0(e) 1.04(b) -1.40(e) 0.89(b)

F. Delayed Neutron Fraction, 5 .0062 1.00(b) .0051 1.05(b)

G. Power Peaking Factor 2.6 4.1 H. Power Peaking Factor Used(<) 6.0 7.5 161.9(d) 159.2(d)

(a) The contribution to the total pellet energy deposition is a function of initial fuel enthalpy, maximum control rod worth, Doppler coefficient, and delayed neutron fraction. The energy de-position contribution values and factors are derived from data calculated in the "Generic Analysis of the Control Rod Ejection Transient...." document.

(b) These values are multiplication factors applied to (C+D).

(c) The energy deposition due to maximum control rod worth is a function of the power peaking factor.

(d) Total pellet energy deposition (cal/gm) calculated by the equation-Total (cal/gm) = (C+D) (E) (F)

(e) For this Doppler coefficient conservative values of -1.0 and -1.40 were assumed at BOC and EOC, respectively.

Table 8. 2 D. C. Cook Unit 2 Cycle 5, Ejected Rod Analysis, HZP BOC EOC Contribution(a) to Contribution(a) to Energy Deposition, Energy Deposition, Value (cal/ m Value cal/ m A. Initial Fuel Enthalpy (cal/gm) 16. 7 16.7 B. Generic Initial Fuel Enthalpy (cal/gm) 16. 7 16.7 C. Delta Initial Fuel Enthalpy (cal/gm) 0.0 0.0 0.0 0.0 D.- Maximum -Control-Rod Worth (pcm) 427 20 667 60 E. Doppler Coefficient, (pcm/oF) -1.0(e) 1.03(b -1.5(e) .73(b)

F. Delayed Neutron Fraction, B .0062 1.00(b) .0051 1.20(b)

G. Power Peaking Factor 5.8 11.4 Power Peaking Factor Used(c) 13.0 13.0 TOTAL 20.6(d) 52. 6(d)

(a) The contribution to the total pellet energy deposition is a function of initial fuel enthalpy, maximum control rod worth, Doppler coefficient, and delayed neutron fraction.. The energy de-position contribution values and factors are derived from data calculated in the "Generic Analysis of the Control Rod Ejection Transient...." document.

(b) These values are multiplication factors applied to (C+D).

(c) The energy deposition due to maximum control rod worth is a function of the power peaking factor.

(d) Total pellet energy deposition (cal/gm) calculated by the equation-Total (cal/gm) = (C+D) (E) (F)

(e) For this Doppler coefficient conservative values of -1.0 and -1.50 were assumed at BOC and EOC, respectively.

XN-NF-83-85

9.0 REFERENCES

1. XN-NF-82-25(A), "Generic Mechanical Oesign Report, Exxon 17x17 Fuel Assembly", Exxon Nuclear Company, April 1982.

2. XN-NF-78-44(A), "A Generic Analysis of The Control Rod Ejection Tran-sient for Pressurized Water Reactors", Exxon Nuclear Company, January 1979.

3. XN-75-27(A), "Exxon Nuclear Neutronics Oesign Methods for Pressurized Water Reactors", Exxon Nuclear company, June 1975.

4. XN-75-27(A), Supplement 1, September 1976.

5. XN-75-27(A), Supplement 2, Oecember 1977.

6. XN-CC-28, Revision 5, "XTG - A Two Group Three-Oimensional Reactor Simulator Utilizing Coarse Mesh Spacing", Exxon Nuclear Company, July 1979.

7. XN-NF-77-57(A), "Exxon Nuclear Power Distribution Control for Pres-surized Water Reactors - Phase II", Exxon Nuclear Company, January 1978.

8. XN-NF-77-57(A), Supplement 1', June 1979.

9. XN-NF-77-57(A), Supplement 2, September 1981.

XN-NF-83-85 Issue Date: 10/24/8

0. C. COOK UNIT 2, CYCLE 5 SAFETY ANALYSIS REPORT DISTRIBUTION GJ BUSSELMAN JC CHANDLER RA COPELAND MR KILLGORE JN MORGAN GF OWSLEY RA PUGH HG SHAW FB SKOGEN GA SOFER RB STOUT T TAHVILI HE WILLIAMSON PD WIMPY DOCUMENT CONTROL (5)

AEP (5) / HG SHAW