ML17319B411

DC Cook Unit 2,Cycle 4 Sar

ML17319B411
Person / Time
Site:	Cook
Issue date:	04/30/1982
From:	Evinay A, Skogen F, Wimpy P SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML17319B410	List: ML17319B409 ML17319B411 ML17319B412
References
XN-NF-82-37, NUDOCS 8207080395
Download: ML17319B411 (61)

Text

XN-NF-82-37 Issue Date: O4/3O/B2 D.

C.

COOK UNIT 2, CYCLE 4 SAFETY ANALYSIS REPORT Written by:

A. Evinay / P.

D. Wimpy PWR Neutronics Reviewed by:

F. B. Skog n, Manager PWR Neutronics Prepared by:

R. B. Stout, Manager Neutronics and Fuel Management

/V>

~P c~

Prepared by:

J.

N. Morg n, Manager Licensing and Safety Engineering Approved by:

G. J.

Buss an, Manager Fuel Design Approved by:

G.

. Sofer, Manager Fuel Engineering and Technical Services

/csk E@C03N MUCE EAIFR CC)IMlPAMV,Inn.

8207080395 820630

'PDR ADOCK 05000316 P

PDR

NUCLEAR REGULATORY COMMISSION DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY This technical report was rlerived through research and development programs sponsored by Exxon Nuclear Company, Inc.

It is being sub.

mitted 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 lioht water power reactors and it is true and correct 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 regulations.

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 8.

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-F00, 766

XN-NF-82-37 TABLE OF CONTENTS Section

1.0 INTRODUCTION

2.0

SUMMARY

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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 CHARACTERISTI(S.

6.1. 1 Power Distribution Considerations

6. 1.2 Control Rod Reactivity Requirements

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2 4

7 11 12 13 13 15

6. 1.3 Moderator Temperature Coefficient Considerations.

16 6.2 ANALYTICALMETHODOLOGY 7.0 THERMAL-HYDRAULIC DES IGN ANALYSIS

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8.0 ACCIDENT AND TRANSIENT ANALYSES 8.1 PLANT TRANSIENT ANALYSIS.

8.2 ECCS ANALYSIS.

8.3 ROD EJECTION ANALYSIS.

9.0 REFERENCES

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16 22 23 23 23 23 27

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XN-NF-82-37 LIST OF TABLES Table 4.1 D. C.

Cook Unit 2, Principal Characteristics for Nuclear Analysis of Cycle 4 Fuel................

6.1 D.

C.

Cook Unit 2 Neutronics Characteristics of Cycle 4

Compared with Cycle 3 Data 6.2 D.

C.

Cook Unit 2 Control Rod Shutdown Margins and Requirements of Cycle 4 Compared to Cycle 3.

8.1 D.

C.

Cook Unit 2 Cycle 4, Ejected Rod Analysis, HFP 8.2 D.

C.

Cook Unit 2 Cycle 4, Ejected Rod Analysis, HZP..

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8 17 18 25 26 LIST OF FIGURES

~Fi use 3.1 D.

C.

Cook Unit 2, Cycle 3 Boron Letdown Curve 3.2 D.

C.

Cook Unit 2 Cycle 3, Power Distribution Comparison to Map 203-50, 100K Power, Bank D 9222 Steps, 8533 MWD/MT.

4.1 D.

C.

Cook Unit 2, Cycle 4 Full Core Loading Pattern 4.2 D.

C.

Cook Unit 2, Cycle 4 Loading Pattern and BOC Exposure Distribution.

6.1 D.

C.

Cook Unit 2, Cycle 4 Boron Letdown Curve 6.2 D.

C.

Cook Unit 2, Cycle 4 Relative Power Distribution 100 MWD/MT, 1089

ppm, 3411 MWt, ARO.

6.3 D.

C.

Cook Unit 2, Cycle 4 Relative Power Distribution 14,150 MWD/MT, 10 ppm, 3411 MWt, ARO

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6 10 19 20 2 1

5

D.

C.

COOK UNIT 2 CYCLE 4 SAFETY ANALYSIS REPORT PROLOGUE This report is the fifth in a series of six reports which address the neutronics characteristics of the Cycle 4 core and provides the safety evaluation for Cycle 4.

Preliminary analyses were performed in response to the Tentative Scheduled Delivery Date (TSDD) notice and were provided in letter reports PWR:021:81 and PWR:025:81.

Subsequently a final reload design was established in response to the Final Scheduled Delivery Date (FSDD) notice and was documented in letter report PWR:034:81.

The Fuel Management

Report, (XN-NF-82-36 (P)), which provided the Reference Design for the safety evaluation was issued in April, 1982.

This Safety Analysis Report will be followed by a Cycle 4 Startup and Operations

\\

Report.

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XN-NF-82-37 D.

C.

COOK UNIT 2 CYCLE 4 SAFETY ANALYSIS REPORT

1.0 INTRODUCTION

The results of the Safety Analysis for Cycle 4 of the D.

C.

Cook r

$ 1 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 rod ejection accident analysis'he Cycle 4 design includes 72 Exxon Nuclear Company (ENC) Reload Batch XN-1 assemblies enriched to 3.65 w/o U-235.

The Cycle 4 design also utilizes 592 A1203-B4C burnable absorber

rods, each containing 0.026 gm/in of B-10.

The burnable absorber rods are distributed among 60 assemblies.

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.l

XN-NF-82-37 2.0 SNMARY The D.

C.

Cook Unit 2 nuclear plant is scheduled to operate in Cycle 4 beginning in November of 1982 with 72 fresh assemblies supplied by Exxon Nuclear Company (ENC)

(.Reload Batch XN-1).

The composition of the core during Cycle 4 will be 72 ENC assemblies in Region 6 and a

total of 121 Westinghouse assemblies; 16 in Region 3, 13 in Region 4, and 92 in Region 5, respectively.

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 and thermal-hydraulic analysis are presented in Reference 1.

The Plant Transient Analysis is documented in Reference 2 while the LOCA-ECCS Break Spectrum Analysis is presented in Reference 3.

The results of the Control Rod Ejection Analysis are provided herein and are derived in part from the generic analysis described in Reference 4.

The neutronics characteristics of Cycle 4 are similar to those of Cycle 3.

The range of kinetics coefficients reflected in the Plant I'ransient Analysis bound those expected in Cycle 4.

The minimum excess shutdown margin is calculated to be 630 pcm at EOC.

A postulated control rod ejection event is conservatively calculated to result in an energy deposition of less than 170 cal/gm.

At hot full power equilibrium xenon conditions, the peak F

, in-N eluding K(Z) considerations, is calculated to be 1.64+.08 and occurs in an assembly supplied by ENC at BOC4.

The peak Fq for Westinghouse (W)

N

I I

XN-NF-82-.,37 E

t fuel is calculated to be 1.48+.08 at hot full 'power equilibrium xe'non conditions, and also occurs at BOC4.

The Technical Specification Limit on the total peaking factor, F~, for ENC supplied fuel is anticipated T

to be at least 2.04 including uncertainties, and the allowable F~ for T

Westinghouse supplied fuel.,is,anticipated to be, 1.97 includinguncer.,;;.,';.;.;

t tainties.

Including a

3X engineering factor, a

5X "measurement uncer-

tainty, and an 11K PDC-II allowance, the total peaking factor, F, during Cycle 4 is calculated to be 1.97+.08 in ENC fuel and 1.78+;08*in 4

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N Westinghouse fuel.

The maximum relative pin power, F<H, during the cycle is calculated to be 1.40

+.07 and occurs at 500 MWD/MT, hot full N

power equilibrium xenon conditions.

The maximum allowable F~H is antici-pated to be 1.49 in ENC supplied fuel and 1.47 in Westinghouse supplied i

T fuel.

Both Fq and F~H are expected to remain within the anticipated limits throughout the cycle.

IL

XN-NF-82-37 3.0 OPERATING HISTORY OF THE REFERENCE CYCLE D.

C.

Cook Unit 2 Cycle 3 has been chosen as the reference cycle with respect to Cycle 4 due to the close resemblance of the neutronic I

characteristics between these two cycles.

The Cycle 3 operations began

/

in May 1981, and as'of the end of April, 1982, the core has accrued about 11,000 MWD/MT exposure.

The Cycle 3 core loading consisted of 193 Westinghouse assemblies.

The measured power peaking factors at hot-full-power, equilibrium xenon conditions, have remained below the Technical Specification limits throughout Cycle 3.

The total peaking factor, F, and the radial pin T

peaking factors, F

H, have remained below 1.99 and 1.49, respectively.

N The Cycle 3 operation has typically been rod free with the D control rod bank positioned in the range of 218 to 225 steps, 228 steps being fully P

withdrawn.

It is anticipated that similar control rod bank insertions wil.l, be used in Cycle 4 operations.

The critical boron concentration as calculated by ENC for Cycle 3

has agreed to within about 40 ppm with the measured values (see Figure 3.1)

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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 8,533 MWD/MT).

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..1600:

1400 1200 1000 O

I 800 o

600 400 200 0

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Meatsured CalIculated I-I I

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(XyG)

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I+

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0 12 14 4

6 8

10 Cycle Exposure (GWD/MT)

Figure 3.1 D. C.

Cook Unit 2, Cycle 3 Boron Letdown Curve

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XN-NF-82-37 A

r,

.855

.840

+1.8 1.034 1.039

-0.5 1.032 1.039

-0. 7 1.058 1.089

-0. 1 1.115 1.133

-1.6 1.238 1.241

-0.2 1.222 1.255

-2.6 1.136 1.124 1.108 1.135

-2.4 1.210 1.216

-0.5 1.060 1.064

-0.4 1.125 1.148

-2.0 1.030 1.051

-2.0 1.119 1.112

+0.6

.858

.855

+0.4

.849

.830

+2.3 1.116 1.131,

-1.3 1.223

l. 251

-2.2 1.108 1.134

-2.3 1.061 1.065

-'0;4

1. 239

1. 237

+0.2 1.136 1.124

+1.1

l. 211 1.211 0.0 1.125 1.148

-2.0

1. 141

1. 135

+0. 5 I 1.225 1.227

, -0.2 1.119 1.117

+0.2

1. 197

, 1.192

~ +0.4 1.225 1.227

-0.2

.948

.990

-4.2 1.144 1.161

-1.5 1.033 1.028

+0.5 1.120 1.115

+0.5

l. 146 1.161

-1.3 1.047 1.050

-0.3

.954

.950

+0.4 1.198 1~ 190

+0. 7 1.035 1.026

+0.9

.956

.946

+1.1

.893

.874

+2.2 1.021 1.032

.987

.993

-0.6

.796

.772

+3.1

.407

.398

+2.3

.758

.740

+2.4

.425

.425 0.0 1.030 1.036

-0.6 1.119 1.114

+0.5 1.021 1.029

-0.8

.986

.993

-0.7

. 795

.772

+3.0

.406

.396

+2.5 Calculated (XTGPWR)

Measured Assembly Power

(')x 100

.858

.855

+0.4

.849

.829

+2.4

.758

. 740

+2.4

.424

.422

+0.5 Calculated Measured X Diff.

1.536 F5H 1.331 1.328

+0.2 Fq 1.532

+0.3 Figure 3.2 D.

C.

Cook Unit 2, Cycle 3, Power Distribution Comparison to Map 203-50, 100K Power, Bank D 8222 Steps, 8,533 MWD/MT

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XN-NF-82-37 4.0 GENERAL DESCRIPTION The D. C.

Cook Unit 2 reactor consists of 193 assemblies, each having a 17x17 fuel rod array.

Each assembly contains 264 fuel rods, 24 RCC guide tubes, and 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 projected Cycle 4 loading pattern is shown in Figure 4. 1 with assemblies identified by their Cycle 3 location.

Fresh fuel is not assigned an ID but the burnable absorber configuration is noted.

The initial enrichment of the various regions are listed in Table 4. 1 The calculated BOC4 exposures, based on an EOC3 exposure of 16,730 NWD/MT, are shown in a quarter core representation in Figure 4.2 along with the quarter core fuel shuffle simulation.

The core consists of 72 fresh ENC assemblies at an average enrichment of 3.65 w/o U-235 and 121 exposed Westinghouse assemblies.

A low radial leakage fuel management plan has been developed and results in the scatter-loading of the fresh fu'el throughout the core with the fresh assemblies loaded in the core interior containing A1203-B4C burnable absorber rods.

The exposed fuel is also scatter-loaded in the center in a manner to control the power peaking.

The Al203-B4C burnable absorber rods contain 0.026 gm/in of 8-10 and 592 of these rods are distributed among 60 fresh assemblies loaded in the core interior.

Pertinent fuel assembly parameters for the Cycle 4 fuel are depicted in Table 4.1.

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XN-NF-82-37 Table 4.1 D. C.

Cook Unit 2, Principal Characteristics for Nuclear Analysis of Cycle 4 Fuel Region 3

Region 4

Region 5

Region 6

Nominal Enrichment (w/o)

Nominal Density (X TD)

Pellet OD (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 (MWD/MT)

BOC4 EOC4 Incremental

3. 10 95

.3225

.374

.0065

.0225

.496 Inconel 144 16 7.364 21,877 35, 293 13,416 3.40 95

.3225

.374

.0065

.0225

.496 Inconel 144 13 5.973 28,514 36,564 8,050 3.40 95

. 3225

.374

.0065

.0225

.496 Inconel 144 92 42.149 17,465 30,569 13, 104 3.65 94

.3030

.360

.0070

.0250

.496 Bi-Metallic ENC 144 72 29.077 17, 105 17, 105

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XN-NF-82-37 R

P N

M L

K J

H G

F E

P C

B A

F15 G15 J15 K15 B12 A10 N6 M3 P3 P14 K3 M4

• • • K5 L6 P5 L4 J4 J6 G2

• • • F3 L15 G4 Ll

• • • F5 H5 G6 E2 E]

• • • E4 C2 D5 B5 M14 P3 C4 P4

• • P]2

• • • B3

• • • C6 Rlo Ag

• • • K7

+** C13 J14 N13

• • • F7

• • • Pj R9 P9 AS A5 LS Rll 87 B 10 P 7 All ES R5 RS B9 Aj N10 A6 B4 N12 P13 M13

• • • Kg Llo P 1 1 L12

• • • Mll L14 M12 *** Kl1 N14 P2 K13 C3 Jlo J12 J2 N3 E15

• • • E12 H1 1 Glo E14 Ll F 11 Hl G12 G14

• • • F13 F9 Bll Dll C14

• • • Pg F10

• • • C]0

• • • B13 P12 P4 C12 M2 D13 R7 R6 10 12 13 14 Fl Gl Kl Location in Previous Cycle 15 fresh Fuel is Not Assigned 4 BA Pins per Assembly 8 BA Pins per Assembly

• • • 12 BA Pins per Assembly Figure 4. 1 D,

C.

Cook Unit 2, Cycle 4 Full Core Loading Pattern

4 l

XN-NF-82-37 Blo 4

27,146 G14+

5 18,780 E15 3

21,089 Hl1 5

20,349 E15+

3 23,630

. H15+

5 14,474 B9+

5'8,783 C13+

5 14,997 Glo 5

20, 749 G12 5

20, 136 All 3

21,094 E12 5

18,983 E14 5

16,560 F 1 1 5

20, 352 ES 5

20,336 F9 5

20,734 Bll 5

16,577 Dll 5

19,025

~ 6 C14 3

- 21,391 All+

3 23,634 E10 5

20,364 D12 4

29,659 A8+

5 14,475 D9 5

20,136 B13 3

21,395 B9 5

18,787 Clo 5

20,007 B12+

5 13,344 C12 4

28, 133 A9+

5 14,293 A10+

5 12, 717 G14 5

18,783 F13 5

19, 992 D14+

5 13,327 D13 4

28,090 0

G15+

6 14,291 F15+

5 12,713 Core Loc at ion in Pre viou s Cyc 1 e Fuel Region Number Assembly Average Exposure (MWD/NT)

Indicates 4 BA Pins per Assembly Indicates 8 BA Pins per Assembly

• • • Indicates 12 BA Pins per Assembly

+

Rotated 180o Degrees Figure 4.2 D. C.

Cook Unit 2, Cycle 4, Loading Pattern and BOC Exposure Distribution

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4' XN-NF-82-37 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.

I I

XN-NF-82-37 6.0 NUCLEAR CORE DESIGN The neutronic characteristics of the projected Cycle 4 core are quite similar to those of the Cycle 3 core (see Section

6. 1).

The nuclear design bases for the Cycle 4 core are as follows:

1.

The design shall permit operation within the Technical Specification for D. C.

Cook Unit 2 nuclear plant.

2.

The length of Cycle 4 shall be determined on the basis of an assumed Cycle 3 energy of 1481.8 GWD (16,730 NWD/HT exposure).

3.

The Cycle 4 loading pattern shall be designed to achieve power distributions and control rod reactivity worths according to the following constraints:

a)

The peak F

shall not exceed the limits shown in Reference I

3 and the peak F<H shall not exceed 1.55 (including uncertainties) in any single 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 5, 6, and 7.

The Cycle 4 loading contains 592 Al203-B4C burnable absorber rods distributed among 60 of the 72 fresh ENC supplied assemblies.

In forty (40) of these assemblies there are twelve (12) burnable absorber rods per assembly.

Another eight (8) assemblies, will each contain eight (8).

XN-NF-82-37 A1203-B4C rods and twelve (12) assemblies will each contain four (4)

A1203-B4C rods.

The A1203-B4C burnable absorber rods each contain 0.026 gm/in of B-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 592 A1203-B4C absorber rods is calculated to be equivalent to the worth of 458 ppm soluble boron.

6.1 PHYSICS CHARACTERISTICS The neutronics characteristics of the Cycle 4 core are compared with those of Cycle 3 and are presented in Table 6.1.

The data presented in the table indicates the neutronic similarity between Cycles 3 and 4.

The reactivity coefficients of the Cycle 4 core are bounded by the coefficients used in the safety analysis.

The safety analysis for Cycle 4 is applicable for Cycle 3 burnup of +1000 MWD/MT and

-1000 NWD/MT about the nominal burnup of 16,730 MWD/MT.

The boron letdown curve for Cycle 4 is shown in Figure 6.1.

The BOC4 xenon free critical boron concentration is calculated to be 1514 ppm.

At 100 NWD/MT, equilibrium xenon, the critical boron concen-tration is 1089 ppm.

The Cycle 4 length is projected to,be 14,150 NWD/MT +300 NWD/NT at a core power of 3411 NWt with 10 ppm soluble boron remaining.

6. 1. 1 Power Distribution Considerations Representative calculated power maps for Cycle 4 are shown in Figures 6.2 and 6.3 for BOC, (equi lbrium xenon),

and EOC con-ditions, respectively.

The power distributions were obtained from a three-dimensional XTG 'odel with moderator density and Doppler feed-(8)

XN-NF-82-37 back effects incorporated.

As shown in Figure 6.2, for the design Cycle 4 loading pattern, the calculated BOC, hot-full-power, equilibrium xenon nuclear power peaking factors, F, and F

H are 1.64, and 1.39, respectively.

N N

At EOC conditions the corresponding values of F~ and F>H are 1.56 and 1.36, respectively for the limiting first cycle fuel.

The

BOC, HFP, equi librium xenon F

value of 1.64 is compared to the measured Cycle 3

value of 1.61 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, and an llà allowance for PDC-II, the expected total peak, F~, is 1.97.

The maximum relative pin power, F<H, T

N T

N is calculated to be 1.40.

Both Fq and F~H are expected to remain within T

the allowable limits throughout the cycle.

The allowable limit on Fq is expected to'be at least 2.04 in ENC supplied fuel and 1.97 in Westinghouse N

supplied fuel.

The allowable F~H is expected to be 1.49 in ENC supplied fuel and 1.47 in Westinghouse supplied fuel.

The control of the core power distribution is accom-plished by following the procedures as discussed in the report, "Exxon Nuclear Power Distribution Control for Pressurized Water Reactors Phase II" '

The results reported in that document provide the (9, 10,11, 12) means for projecting the maximum F (Z) distribution anticipated during operation under the PDC-II procedure taking into account the incore measured equilibrium power distribution data.

A comparison of this

XN-NF-82-37 distribution with the Technical Specification limit curve assures that the Technical Specification limit will not be exceeded by the PDC-II procedure.

The PDC-II document describes the maximum possible variation in F (Z) which can occur during operation when following the outlined procedure.

This bounding variation, in F~(Z) represents the maximum T

variation when the axial offset is maintained within the range defined in the report

(+5% at full power condition).

6. 1.2 Control Rod Reactivit Re uirements Detailed calculations of shutdown margins for Cycle 4

are compared with Cycle 3 data in Table 6.2.

The D.

C.

Cook Unit 2 nuclear plant Technical Specifications require a minimum required shut-down margin of 1600 pcm at BOC and EOC.

The Cycle 4 analysis indicates excess shutdown margin of 656 pcm at BOC and 630 at the EOC.

The Cycle 3 analysis indicated an excess shutdown margin of 580 pcm at BOC and 100 pcm at EOC.

The control rod groups and insertion limits for Cycle 4

will remain unchanged from Cycle 3.

With these limits the nominal worth of the control bank, D-Bank, inserted to the insertion limits at HFP is 146 pcm at BOC and 234 pcm at EOC.

The control rod shutdown requirements in Table 6.2 allow for a HFP D-Bank insertion equivalent to 400 pcm and 500 pcm at BOC and EOC, respectively.

XN-NF-82-37

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 705 of 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 +2.

pcm/

F and meets the Technical Specification limit below 70K power.

The moderator temperature coefficient at or above 70%%d rated power is calculated to be less than 0 pcm/

F and also meets the Technical Specifications.

0 I

6.2 ANALYTICALMETHODOLOGY The methods used in the Cycle 4 core analysis are described in References 5, 6, and 7.

In sugary, the reference neutronic design analysis of the reload core was performed using the XTG

reactor (s) simulator code.

The input isotopics data were based on quarter core depletion calculations performed for Cycle 3 using the XTG code.

The fuel shuffling between cycles was accounted for in the calculations.

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

I XN-NF-82-37 Table

6. 1

. D. C.

Cook Unit 2 Neutronics Characteristics of Cycle 4 Compared with Cycle 3 Data Cele 3

BOC EOC Cele 4

BOG EOC Critical Boron

HFP, ARO, Equi librium Xenon (ppm)

HZP, ARO, No Xenon (ppm) 1493 (b )

21(b )

1921(a )

1089 1514 10 Moderator Temperature Coefficient

HFP, (pcm/oF)

HZP, (pcm/oF)

+3.8(a)

,(0.0

-27.5

+1.9

-21.6 Isothermal Temperature Coefficient

HFP, (pcm/oF )

HZP,.(pcm/oF)

Doppler Coefficient (pcm/oF)

Boron Worth, (pcm/ppm)

HFP HZP

+1.59(a)

-1.4

-7.8(a)

-,2. 9 4.5

+0.2

-,l. 4

-7.7

-8.0

-29.1

-23.3

-1.6

-8.7

-9.2 Total Nuclear Peaking Factor F

HFP, Equi librium Xenon N

Delayed Neutron Fraction

.0075

.0044 1.61(a) 1.47(b) 1.64 1.56

.0057

.,0051 Control Rod Worth of All Rods In Minus Most Reactive

Rod, HZP, (pcm)

Excess Shutdown Margin, (pcm) 5720 580 5580 100 5284 5978 656(c) 630(c)

(a )

Measured Data (b)

ENC Calculated (c)

Shutdown Margin Evaluation Based on Hot Full Power

XN-NF-82-37 Table 6.2 D.

C.

Cook Unit 2 Control Rod Shutdown Margins and Requirements of Cycle 4

Compared to Cycle 3

Control Rod Worth (HZP) cm Cele 3

BOC EOC C cle 4(b)

BOC EOC All Rods Inserted (ARI)

ARI less most reactive (N-1)

N-1 less 10K allowance [(N-l)*.9)j Reactivit Insertion, cm Power Defect (Moderator

+ Doppler)

Flux Redistribution Void Sum of the above three Rod insertion allowance Total Requirements 5720 5150 1690 1280 2970 5580 5020 2820 500 3320 6284 5284 4756 1450 600 50 2100 400 2500 6978 5978 5380 2000 600 50 2650 500 3150 Shutdown Margin (N-1) *.9 - Total Requirements Required Shutdown Margin Excess Shutdown Margin 2180 1700 2256 2230 1600'(a)

'600'(a) 1600(a) 1600(a) 580 100 '56

, 630 (a )

Technical Speci ficat i on Limit (b)

Shutdown evaluation based on hot full power.

If the most adverse combination of power level and rod insertion is assumed, the excess shutdown margin at BOC is reduced by no more than 200 pcm and the excess shutdown margin at EOC remains unaffected.

1200 1000 800 O

I 600 OIO 400 O

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CD 200 0

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4 Cycle Exposure (GwoAT) 10 12 14 16 Figure 6.1 O.

C.

Cook Unit 2, Cycle 4, Boron Letdown Curve

-ZU-XN-NF-82-3/

.954 1.092 1.025 1.057

.972 1.104 1.006 1.092

1. 178 1;279 1.098 1.239 1.087 1.109

.707 1.024 1.278 1.166

1. 156

1. 120

1. 216

.981

.821 1.057 1.097 1.156

l. 142 1.224

.982 1.084

.505

.972 1.238 1.119

l. 221

.928 1.126

.695 1.086

1. 215

.979 1.123

.952

.337 1.005 1.108

.980 1.082

.694

.337 Assembly Relative Power

.883

.707

.820

.504 Peak Assembly

= 1.279 (F9)

Pin F

H

= 1.388 (F9) gH Peak F~

= 1.641 (F9)

Figure 6.2 D.

C.

Cook Unit 2, Cycle 4, Relative Power Distribution 100 t1WD/t1T, 1089 ppm, 3411 HWt, ARO

I XN-NF-82-37

<<D

.886

.995

.981 1.009

.977 1.073 1.012

.879

.995 1.072 1.214 1.043 1.231 1.083 1.152

.750

.981 1.213 1.071 1.067 1.096

'1.246 1.002

.834 1.009 1.042 1.068 1.088 1.247 1.022 1.073

.551

.977 1.231 1.096

1. 246

.987 1.162

.756 1.073 1.083 1.246 1.021 1.161 1.007

.412 1.012 1.152 1.002 1.073

.756

.412 Assembly Relative Power

879

.750

.834

.551 Peak Assembly

= 1.247 (Dll)

Pin F~H

= 1.362, (C10)

Peak F

= 1.561 (C10)

Q Figure 6.3 D.

C.

Cook Unit 2, Cycle,4,'elative Power Distribution 14,150 NWD/NT, 10 ppm, 3411 NWt, ARO

XN-NF-82-37 7.0 THERMAL-HYDRAULIC DES IGN ANALYSIS Thermal-hydraulic design analyses for ENC fuel that is being placed in D.

C.

Cook Unit 2 for this "cycle are given in Reference 1.

Thermal margins for ENC fuel calculated in these analyses are conservatively applicable to mixed core loadings of ENC and the resident Westinghouse fuel.

XN-NF-82-37 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 are reported in Reference 2.

Thermal margins from this analysis are conservatively applicable to mixed core loadings of ENC and the resident Westinghouse fuels 8.2 ECCS ANALYSIS The LOCA-ECCS analysis for ENC fuel at D.

C.

Cook Unit 2 is reported in Reference 3.

This analysis is applicable to ENC fuel for mixed core loadings of ENC and the resident Westinghouse fuel at D. C.

Cook Unit 2.

The analysis remains valid so long as any NSSS modifications and changes in system operational parameters continue to be bounded by the analysis.

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 con-sequence 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 pro-cedures 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

ll,

~

XN-NF-82-37 I

Doppler or moderator feedback in the calculation of ejected rod worths or resultant peaking factors.

The calculations made for Cycle 4 using a

full core XTGPWR model were two-dimensional with appropriate axial buckling correction.

The total peaking factor, F~, were determined as the product T

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 167.0 cal/gm at BOC and,168.0 cal/gm at EOC.

The HZP pellet energy deposition was calculated to be less than 31 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 signi-ficant parameters for the analyses, along with the results, are summarized in Tables 8.1 and 8.2.

I I'

Table 8.1 D. C.

Cook Unit 2 Cycle 4, Ejected Rod Analysis, HFP Value BOC Contribution(a) 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, B

Power Peaking Factor Power Peaking Factor Used(c) 67.1 40.8 26.3 200

-1.0(e)

.0057 3.4 6.0 26.3 131 1.04(b) 1.02(b )

167.0(d) 65.2 40.8 24.4 268

-1.40(e)

.0051 6.4 7.5 24.4 155 0.89(b) 1.05'(b )

168.0(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, re specti ve ly-

. Table 8.2 D. C.

Cook Unit 2 Cycle 4, Ejected Rod Analysis, HZP BOC EOC A.

Initial Fuel Enthalpy (cal/gm)

.B.

Generic Initial Fuel Enthalpy (cal/gm)

C.

Delta Initial Fuel Enthalpy (cal/gm)

D.

Maximum Control Rod Worth (pcm)

E.

Doppler Coefficient, (pcm/oF)

F.

Delayed Neutron Fraction, B

G.

Power Peaking Factor H.

Power Peaking Factor Used(<)

Value 16.7 16.7 0.0 353

-1.0(e )

.0057 5.9 13 Contribution(a ) to Energy Deposition, (cal/ m) 0.0 20 1.03(b) 1.05(b)

Value 16.7 16.7 0.0 540

-1.5(e)

.0051 11.8 13.0 Contribution(>) to Energy Deposition, (cal/ m) 0.0 35

.73(b) 1.20(b)

TOTAL 21.6(d )

30.7(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.

I XN-NF-82-37

9.0 REFERENCES

1.

XN-NF-82-13 (NP),

"D. C.

Cook Unit 2, Design

Report, 17x17 Fuel Assembly",

Exxon Nuclear Company, to be issued.

2.

XN-NF-82-32, "Plant Transient Analysis for "Donald C.

Cook Unit 2 Reactor at 3425 MWt", Exxon Nuclear Company, April 1982.

3.

XN-NF-82-35, "D. C.

Cook Unit 2, LOCA-ECCS Analysis Using EXEM/PWR",

Exxon Nuclear Company, April 1982.

4.

XN-NF-78-44, "A Generic Analysis of The Control Rod Ejection Transient for Pressurized Water Reactors",

Exxon Nuclear

Company, January 1979.

5.

XN-75-27, "Exxon Nuclear Neutronics Design Methods for Pressurized Water Reactors",

Exxon Nuclear company, June 1975.

6.

XN-75-27, Supplement 1, September 1976.

7.

XN-75-27, Supplement 2,

December 1977.

8.

XN-CC-28, Revision 5, "XTG - A Two Group Three-Dimensional Reactor Simulator Utilizing Coarse Mesh Spacing",

Exxon Nuclear

Company, July 1979.

9.

XN-76-40, "Exxon Nuclear Power Distribution Control for Pressurized Water Reactors",

Exxon Nuclear

Company, September 1976.

10.

XN-NF-77-57, "Exxon Nuclear Power Distribution Control for Pressurized Water Reactors

- Phase II", Exxon Nuclear

Company, January 1978.

11.

XN-NF-77-57, Supplement 1,

June 1979.

12.

XN-NF-77-57, Supplement 2, September 1981.

I I

I

XN-NF,-82-37 Issue Date: 04/30/82 D. C.

COOK UNIT 2, CYCLE 4 SAFETY ANALYSIS REPORT D ISTR IBUT ION G. J.

BUSSELNAN.

G.

C.

COOKE A.

EV INAY N.

R.

KILLGORE J.

N.

MORGAN G.

F.

OWSLEY R. A.

PUGH F.

B.

SKOGEN G. A.

SOFER R. B.

STOUT G.

N.

WARD P.

D.

WIMPY AEP (5) /

HG SHAW DOCUMENT CONTROL (5)

I 1

i ti ll l