ML17320A945
| 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: |
| 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: Reviewed by: Prepared by: Prepared by: Approved by: P.D.Wi , Engineer PWR Neutronics F.B.Skogen, Manager.PWR Neutronics
~~/H.E.Williamson, Manager Neutronics and Fue Management
-Ps R.B.Stout, Manager Licensing and Safety Engineering (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 050003ih PDR P.
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
1.0 INTRODUCTION
.
2.0
SUMMARY
.3.0 OPERATING HISTORY OF THE REFERENCE CYCLE.4.0 GENERAL DESCRIPTION
.5.0 FUEL SYSTEM DESIGN.6.0 NUCLEAR CORE DESIGN.6.1 PHYSICS CHARACTERISTICS.
6.1.1.Power Distribution Considerations
.6.1.2 Control Rod Reactivity Requirements
.~Pa e~~2.4.7.12;13 14.15.16 6.1.3 Moderator Temperature Coefficient Considerations.
..17 6;2 ANALYTICAL METHODOLOGY
.7.0 THERMAL-HYDRAULIC DESIGN ANALYSIS.8.0 ACCIDENT AND TRANSIENT ANALYSES.8.1 PLANT TRANSIENT ANALYSIS.8.2 ECCS ANALYSIS.8.3 ROD EJECTION ANALYSIS.
9.0 REFERENCES
.
.17.23.24.24.24.24~28 XN-NF-83-85 LIST OF TABLES Table 4.1 0.C.Cook Unit 2, Principal Characteristics for Nuclear Analysis of Cycle 5 Fuel.~Pa e~~~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~Fi are 3.1 0.C.Cook Unit 2, Cycle 4 Boron Letdown Curve.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.4.1 0.C.Cook Unit 2, Cycle 5 Full Core Loading Pattern.4.2 0.C.Cook Unit 2, Cycle 5 Loading Pattern and BOC Exposure Distribution.
6.1 0.C.Cook Unit 2, Cycle 5 Boron Letdown Curve.6.2 D.C.Cook Unit 2, Cycle 5 Relative Power Distribution 100 MWD/MT, 1149 ppm, 3411 MWt, ARO.6.3 D.C.Cook Unit 2, Cycle 5 Relative Power Distribution 17,900 MWD/MT, 10 ppm, 3411 MWt, ARO.~Pa e 5 6 10 11 20 21 22 h ,\
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 At hot full power equilibrium xenon conditions, the peak F is N calculated to be 1.64 and occurs at BOC in an assembly supplied by ENC.The peak F~for Westinghouse (W)supplied fuel is calculated to be 1.40 N 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 axial flux difference), the total peaking factor, F , during Cycle 5 is T calculated to be 1.97 in ENC supplied fuel and 1.68 in Westinghouse supplied fuel.The maximum relative pin power, F<H, during the cycle is N 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.
Throughout the cycle, both F and F<H are expected to remain within the T'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 throughout Cycle 4.The total peaking factor, F , and the radial pin T peaking factors, F<H, have remained below 2.04 and 1.49, respectively.
The N 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).
o o o o I I I I I I I I I I I J I I I I I I I I I I I I I I I I I~I I I I I I I I I I I I I I I I I I I+I'I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I 7 I I I I I I I I I I I I I I I I I I I I I I I I I I I I T I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I-"-"r"---L GEND,'FILC RTQ NEBSUREO I I T I I p CD~o Co 0 0 I I I I I I I I I I I I I I I I I+I I I I I I I I I I I I Z o'o Q~p CD CC, C3 LJ1 o o Q o Q r I I I I I I.I I I I I I I I I I I I I I I I I I I I I r I I I I I I+I l I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I 7 I I I I I I I I I r I I I I I I I I I I I I I I I I I I-----I--I I I I I I I I I I I I I I I I I I Q Q'p K o CD I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I T I I I I I I I I I I I I I I I-I I I CD o 3-0 S.o 5.0 6.0 7.0 B.O 9.0 10.0 11.0 CYCLE EXPOSURE (GAD/I1T)Figure 3.1 D.C.Cook Un)t 2, Cycle 4 Boron Letdown Curve 0.0 1.0 2.0 12.0 I I I I I I I I I OC I II I 13.0.10.0 g XN-NF-83-85' G E 0 ,C.856..848+0.9.982.968+1.4.974.972+0.2 1.049 1.041+0.8.970.983-1.3 1.068 1.059 W.9.985.966+2.0 1.079 1.064+1.4 1.220 1.187+2.8 1.076 1.065+1.0 1.219 1.196+1.9.980.986-0.6.975.964+1.1 1.218 1'.186+2 7 1.081 1.079+0.2 1.095 1.072'+2.1 1.105 1.118-1.2 1.227 1.221+0.5 1.046 1.042+0.4 1.071 1.069 W.2 1.073 1.080-0.6 1.093 1.099-0.5 1.246 1.240+0.5 1.023 1.051 2~7.971.983-1.2 1.218 1.206+1.0 1.104 1.123-1.7 1.246 1.249-0.2.990 1.030-3.9 1.173 1.190'1.4 1.069 1.086-1.6.997 1.002-0.5 1.234 1.243-0.7 1.024 1.058"312 1.175 1.195-1.7 1.019 1.031-1.2 1.008 1.019 1.128 1.125+0.3 1.020 1.039-1.8 1.107 1.126-1.7;758.766-1.0.396.401-1.2.901.859+4.9.742.737+0.7.862.835+3'.554.564-1.8 1.007 1.010-0.3.903 1.124 1.109+1.4.748.999 1.030-3.0.857 1.102 1.105-0.3.551.755.757-0.3.395.395 0.0 Calculated (XTGPWR)Measured Assembly Power-x 100 C-M M.852+6.0.735+1.8.822'+4.3.558-1.3 N F q Calculated Measured X Oi ff.1.354 1.343+0.8 1.565 1.557+0.5 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'uel are depicted in Table 4.1. XN-NF-83-85 Table 4.1 O.C Cook Unit 2, Principal Characteristics for Nuclear Analysis of Cycle 5 Fuel Nominal Enrichment (w/o)Nominal Density (X TO)Pellet 00 (in)Clad OD (in)Diametral Gap (in)Clad Thickness (in)Rod Pitch (in)Spacer Material Fuel Supplier Fuel Stack Height Nominal (in)Number of Assemblies Regionwise Loading (MTU)Exposure (MWO/MT)BOC5 EOC5 Incremental
~Re ion 5 3.40 95.3225.374.0065.0225.496 Inconel 144 29 13.286 24,069 34,866 10,797~Re ion 6 3.65 94.3030.360.0070.0250.496 Bi-Metallic ENC 144 72 29.077 16,368 35,410 19,042~Re ion 7 3.64 94'3030.360.0070.0250.496 Bi-Metallic ENC 144 92 37.154 0 19,546 19,546 XN-NF-83-85 R P N M L K J M2 R47 H G F E 0 C B A 02 R8 P4 R78 N8 R19 A10 S10 C12 519 J7 R92 P5 S63 F15 S03 P7 S45 N6 530 L2 S21 J6 S48 013 S27 K3 S66 K7 S31 L4 S51 E15 R6 Jl R73 Hl S06 J12 S37 G4 S52 E4 S46 L15 R65 M13 S13 F3 532 F7 S35 E2 S57 G6 S41 K15 501 B7 S23 C6 553 H3 R46 85 S61 G7 R70 R10 S08 N12 S17 B4 R23 M5 S28 All'd R37 J2 S39 N13 S54 G2 S42 Rll R89 05 S43"+P12 R60 R7 R4 C4 512 R8 S07 Mll S65 P11 S60 07 S72 N10 S25 A5 R2 J10 S47 M9 C13 S34 558 J14 S68 K9 S64 E1 R33 J15 R54 C3 S56 G12 S22 N3 S20.G14 S69 Ll R81 09 S15 F9 S38 R5 R57 G10 526 M7 S49 C10 S33 011 S70 B11 S62 A9 R9 N4 S50 B12 R62 A6 502 J9 Rl H13 R42 P9 529 Fl S11 L14 S59 K13 S36 03 518 L12 S67 J4 S71 H15 S24 G15 R36 E12 S14 F13 S44 c M3 S16 E14 S55 B9 S40 Kl S04 b R6 S09'B R52 G9 R71 M14 R49+Fresh Fuel Assembly, a Fresh Fuel Assembly, b Fresh Fuel Assembly, c Fresh Fuel Assembly, d Fresh Fuel Assembly, 014 R3 No BA Pins, 4 BA Pins 12 BA Pins 16 BA Pins 20 BA Pins Previous Core Location Fabrication IO figure 4.1 O.C.Cook Unit 2, Cycle 5 Full Core Loading Pattern XN-NF-83-85 E G15 90 24,353 C13 90 13,487 G12 17,973 C13 270 16,089 G14 16,177 09 17,611 F.9 17,865 All 180 19,856 09 180 17,973 C10 17,906 AS 12,945 011 18, 183 A9'3,843 0 C12'80 17,142 G12 180 17,611 H15 E15 180 19,790'0 E12 F13 17,776 G10 17,883 0 89 16,265 E14 811 16,105 A10 180 12,404 CS 812 23,028 12,972 G15 24,023 18,190 013 180 17,109 15,989 0 F15 180 12,295 0 v G9 30,235 28,483 014 23,083 Core Location in Previous Cycle Rotation (degrees)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:
a)The peak F~and the peak F<H shall not exceed the Technical T N 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 model with moderator density and (6)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 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 N Cycle 4 value of 1.59 in Table 6.1.At hot full power, equilibrium conditions, the peak F N 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 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 T N 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/F below 70K of 0 rated power and less than or equal to 0 pcm/F at or above 70K power.The 0 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 analysis of the reload core was performed using the XTG reactor (6)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 BOC C cle4 EOC BOC C cle5 EOC Critical Boron HFP, ARO, Eq.Xenon (ppm)HZP, ARO, No Xenon (ppm)989(b)]0(b)1 149 1,465 (a)--------1,569 10 Moderator Temperature Coefficient HFP, (pcm/oF)4 0 (b)-27.5(b)-2.1 HZP, (pcm/oF)-0.97(')-21 9()+3 0-26.3-21.1 Ooppler Coefficient (pcm/oF)-1.4 Isothermal Temperature Coefficient HFP, (pcm/oF)-5.4 (b)HZP, (pcm/oF)-2.S6(a)-1.6-1.3-29.2(b).-3.4-23.6(b)+1.3-27.8-23.0-1.5 Boron Worth, (pcm/ppm)HFP HZP Total Nuclear Peaking Factor F~, HFP, Equilibrium Xenon N Oelayed Neutron Fraction 1.59 (a)1.55 (b)1.64.0057.0051.0062-7.7 (b)-S.7 (b)-S.O-S.95(a)-1O.9(b)-9.4-9.6-11.7 1.54.0051 Control Rod Worth of All Rods In Minus Most Reactive Rod, HZP, (pcm)5,525 Excess Shutdown Margin, (pcm)(c)722 6,093 734 6,301 1,008 6,172 721 (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 Control Rod Worth (HZP), cm BOC C cle 4 Cycle 5 EOC, BOC EOC All Rods Inserted (ARI)ARI Less Most Reactive (N-1)N-1 Less 10K Allowance[(N-l)*.9)]
Reactivit Insertion, cm(a)6,348 5,525 4,972 6,888 6,093 5,484 7,065 6,065 5,458 7,279 6,079 5,471 Power Oefect (Moderator+Oopplar) 400 Flux Redistribution 600 Void 50 Sum of the Above Three 1,050 Rod Insertion Allowance 1,600 500 600 50 1,150 2,000 400 600 50 1,050 1,800 500 600 50 1,150 2,000 Total Requirements 2,650 3,150 2,850 3,150 Shutdown Margin (N-l)*.9-Total Requirements Required Shutdown Margin Excess Shutdown Margin 2 322 1600(b)722 2,334 1600(b)734 2,608 1600(b)2,321 1600(b)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.
1600 I l Pl l~il~~~I~~,.i I'~1400~~~I l~'l~~i~~I 1200-~l~i*~~i P P*~I I*~I~~l I"".~~~I~o 1000 800 C O$-O 600 ill O I I l 400 200 0 0~I~I P~I~I'I~~I'~I: P I~I I I I I 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=1.644 (G15)Peak Assembly=1.206 (H9)Pin F~=1.323 (H9)Peak F~N 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 rection.The total peaking factor, F~, were determined as the product of T 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 Value BOC Contribution(>)
to Energy Deposition, (cal/m)Value EOC Contribution(a) to Energy Deposition, cal/m)A.B.C.D.E.F.G.H.Initial Fuel Enthaply (cal/gm)Generic Initial Fuel Enthalpy (cal/gm)Delta Initial Fuel Enthalpy (cal/gm)Maximum Control Rod Worth (pcm)Doppler Coefficient (pcm/oF)Delayed Neutron Fraction, 5 Power Peaking Factor Power Peaking Factor Used(<)66.5 40.&25.7 179-1.0(e).0062 2.6 6.0 25.7 130 1.04(b)1.00(b)161.9(d)68."2 40.8 27.4 194-1.40(e).0051 4.1 7.5 27.4 143 0.89(b)1.05(b)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 A.B.C.-D.-E.F.G.Initial Fuel Enthalpy (cal/gm)Generic Initial Fuel Enthalpy (cal/gm)Delta Initial Fuel Enthalpy (cal/gm)Maximum-Control-Rod Worth (pcm)Doppler Coefficient, (pcm/oF)Delayed Neutron Fraction, B Power Peaking Factor Power Peaking Factor Used(c)Value 16.7 16.7 0.0 427-1.0(e).0062 5.8 13.0 Contribution(a) to Energy Deposition, (cal/m 0.0 20 1.03(b 1.00(b)Value 16.7 16.7 0.0 667-1.5(e).0051 11.4 13.0 Contribution(a) to Energy Deposition, cal/m 0.0 60.73(b)1.20(b)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 10/24/8 Issue Date: 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