ML17325A428

Reload Safety Evaluation DC Cook Nuclear Plant Unit 1 Cycle 10.

ML17325A428
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Site:	Cook
Issue date:	08/31/1987
From:	Dzenis E, Skaritka J INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
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Text

Attachment 3 to AEP:NRC:0940H Unit 1 Cycle 10 Reload Safety Evaluation Report 87i 1D30544

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RELOAD SAFETY EVALUATION D.C. COOK NUCLEAR PLANT UNIT 1, CYCLE 10 August, 1987 Edited by J, Skaritka Approved:

E. A. Dzenis, Manager Core Operations

TABLE OF CONTENTS Ti tie Page Table of Contents List of Tables List of Figures

1.0 INTRODUCTION

AND

SUMMARY

1. 1 Introduction 1.2 General Description 1.3 Conclusion 2.0 REACTOR DESIGN 2.1 Mechanical Design 2.2 Nuclear Design 2.3 Thermal and Hydraulic Design 3.0 POWER CAPABILITY AND ACCIDENT EVALUATION 3.1 Power Capabi 1 i ty 3.2 Accident Evaluation

4.0 REFERENCES

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LIST OF TA8LES Table Title Page Fuel Assembly Oesign Parameters Kinetic Characteristics Shutdown Requirements and Margins 13 3S18f;5-C70d1 9

LIST OF FIGURES Figure Title Page Core Loading Pattern 14 Heat Flux Hot Channel Factor Normalized 15 Operating Envelope for Westinghouse Fuel 39ldFld d70b19

1.0 INTRODUCTION

AND

SUMMARY

1. 1 INTRODUCTION This report presents an evaluation for D. C. Cook Unit 1, Cycle 10 design, which demonstrates that the core reload wi 11 not adversely affect the safety of the plant. The Cycle 10 evaluation was accomplished utilizing the methodology described in WCAP-9273, "Westinghouse Reload Safety Evaluation Methodology".'(1)

The RWST and Accumulator boron concentrations have been increased to 2400 to 2600 ppm with the approval of Technical Specification changes (2)

All of the accidents comprising the licensing bases (3) which could potentially be affected by the fuel reload have been reviewed for the Cycle 10 design described herein. The justification for the applicability of the Cook Unit 1 accident analyses is presented in Sections 3. 1 and 3.2.

1. 2 GENERAL DESCRIPTION The Cook Unit 1 Cycle 10 reactor core is comprised of 193 fuel assemblies arranged in the core loading pattern configuration shown in Figure 1. During the Cycle 9/10 refueling, 34 ENC fuel assemblies and 46 Westinghouse (W) optimized fuel assemblies (OFAs) will be replaced with W Regions 12A and 12B fresh OFAs. A summary of the Cycle 10 fuel inventory is given in Table l.

-'.NC -, Exxon Nuclea. Ccrporation

Consistent with the use of the W Improved Thermal Design Procedure ( ITDP) for (4b analyses', the core design parameters utilized for Cycle 10 are as follows:

Core Power (MWt) 3250 (3411)**

Core Pressure (psia) 2280+

Vessel Average Temperature ('F) 567.8 (577.1)**

Minimum Measured Flow (gpm) 366,400 Average Linear Power Density (kw/ft) 6.70 (7.03)**

(based on average active fuel stack length of 144 inches)

1.3 CONCLUSION

From the evaluation presented in this report, it was concluded that the Cycle 10 design does not result in the safety limits for any incident being exceeded. This conclusion is based on the following:

1. Cycle 9 actual burnup of 16,096 MWD/MTU

2. Cycle 10 burnup is limited to the end-of-full power capability (EOFPC)~

plus 1500 MWD/MTU for power coastdown.

3. There is adherence to plant operating limitations as given in the Technical Specifications; no changes are needed for Cycle 10 to operate safely.

4. A steam generator tube plugging level of 3.2% is acceptable provided the Technical Specification requirement on RCS flow (Tech. Spec. Table 3.2-1) can be satisfied.

~Definition - with control rods fully withdrawn and approximately 0 to 10 ppm of residual boron at the Cycle 10 3250 MWt rated power conditions.

"" Values used in Thermal Hydraulic Analysis

+ Corresponds to Pressurizer Pressure of 2250 psia 8518F.6 870819

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2.D REACTOR DESIGN 2.1 MECHANICAL DESIGN The new Region 12A and 128 fuel assemblies are M 15xl5 OFAs. The mechanical design of the Region 12 fuel assemblies is the same as the Region ll assemblies,'xcept for the use of 4g plenum springs and radiused (bullet nose) fuel rod bottom end plugs.

The 4g spring provides more gas plenum volume for fission gas release and reduces the potential for pellet chipping. The smaller pellet holddown spring in the fuel rod gas plenum satisfies a change in the nonoperational 6g loading design criterion to "4g axial and 6g lateral loading." Mestinghouse has incorporated this criterion change, and the justification of no unreviewed safety question was transmitted to the NRC via Reference 5.

The fuel rod bottom end plug was changed only from a chamfered end to a radiused end to improve rod loading and reduce the potential of grid damage during rod loading. This minor design change satisfies all applicable FSAR design criteria. Also, the impact of this change has been evaluated with respect to the licensing basis accident events (See Section 3.2).

Table 1 presents a comparison of pertinent design parameters of the various fuel regions. The Region 12A and 128 fuel have been designed utilizing the M fuel performance model' and the M clad flattening model (7) . The M fuel is designed and operated so that clad flattening will not occur for its planned residence time in the reactor, The fuel rod internal pressure design basis is satisfied for all fuel regions.

Mestinghouse's experience with 2ircaloy clad fuel and OFAs is described in Reference 9. This report is updated annually.

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L 2.2 NUCLEAR DESIGN For the Cycle 10 nuclear design the Mestinghouse Advanced Nodal Code (ANC),Reference 10, was introduced to perform core neutronics analyses. This supplemented the standard reload methodology design codes given in Reference

1. ANC incorporates several improvements to the PALADON code used in previous core designs. A favorable Safety Evaluation Report (SER), Reference 11, was received from the NRC approving the use of the ANC for nuclear design analyses.

The nuclear design of the Cycle 10 core used M codes approved by the NRC and the standard calculational methods described in the M Reload Safety Evaluation Methodology. (1) The Cycle 10 core loading satisfies the approved technical specification {F< (Z) x P] LOCA envelope limit of < 2. 10 (Figure 2 and Section 3. 1).

Table 2 provides a comparison of the Cycle 10 kinetics characteristics with the evaluation limits based on the accident analyses (2,3) ' It can be seen from the Table 2 parameters that all of the Cycle 10 values, except the Doppler temperature coefficient, fall within the evaluation limits. (These parameters are evaluated in Section 3.0.) Table 3 provides the beginning-and end-of-life control rod worths and requirements at the most limiting condition during the cycle; the available shutdown margin exceeds the minimum required.

The control rod insertion Technical Specification limits assure that peaking factors are not exceeded during anticipated power control maneuvers.

Region 12A and 12B fuel assemblies contain a total of 544 new HABA rods (Figure 1). These rods are required for moderator temperature coefficient and power peaking control. Four previously irradiated secondary source rods will be located in two Region 11A assemblies and two Region 12B assemblies (see Figure 1).

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2.3 THERMAL AND HYDRAULIC DESIGN No variation of thermal margins resulted from the all M 15x15 OFA core for the Cycle 10 reload. The present core safety limits in the technical specifications are conservative for the Cycle 10 reload core. Sufficient ONB transient to meet the design criteria (3,12) 'or margin exists for all Condition I and II events and the steamline break the .Cycle 10 reload core.

The table below shows the relationships which exist between the correlation limit ONBR, design limit DNBR, and the safety analysis limit DNBR values used for this design, using the Improved Thermal Design Procedure (4)

'A

( ITDP)

Typical Thimble Correlation Limit 1.17 1.17 Design Limit 1.32 1.31 Safety Analysis Limit 1.69 1.69 For events where conditions fall outside the range of applicability of the HRB-1 correlation (and ITDP), the W-3 correlation is used with the following 1 imi ts:

Correlation Limit 1.30 for core pressures > 1000

1. 45 for pressures 500 to 1000 ps i a The transition core penalty (5% ONBR) applied on the OFA for the Cycle 9 H/ENC mixed fueled core is no longer applicable, since the Cycle 10 core contains only H OFAs.

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3.0 POWER CAPABILITY AND ACCIDENT EVALUATION

3. 1 POKER CAPABILITY The plant power capability for Cycle 10 is evaluated by considering the consequences of those FSAR incidents which appear as the licensing basis accident analysis 3 ' 13

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It is concluded that the core reload will not adversely affect the ability to safely operate at the current 3250 NWt rated power during Cycle 10. For overpower transients, the fuel centerline temperature limit of 4700'F can be accommodated with margin in the Cycle 10 core. The time dependent densification model (14) was used for fuel temperature evaluations. The LOCA limit at 3250 Wt is met by maintaining

[FQ(z)xP] at or below [2. 10' xK(Z)], according to the normalized FQ envelope shown in Figure 2. This limit is satisfied by the power control maneuvers allowed by the Technical Specifications, which assure that the Final Acceptance Criteria (FAC) limits are mei for a spectrum of small and large LOCA's.

3.2 ACCIDENT EVALUATION The effects of the reload on the postulated licensing bases incidents (3) were examined. Supporting safety analyses (13) , submitted with the Technical Specification changes, justify a number of plant changes (See Section 1.1).

Also, an evaluation was performed for the fuel rod bottom end plug change identified in Section 2.1. In all cases it was found that the effects can be accommodated within the conservatism of the initial assumptions used in the applicable safety analysis An evaluation has demonstrated that the Cycle 10 reactor satisfies the subcritical requirement following a postulated large break LOCA.

"The 2. 10 maximum FQ value allows a maximum of 5 percent steam generator tube plugging.

3518F 6.87C8l9

In addition, the impact of a 3.2% steam generator tube plugging level has been examined. The conclusions of the current safety analyses remain valid pro-vided the Tech. Spec. requirement on RCS flow (Tech. Spec. Table 3.2-1) are satisfied.

A core reload can affect accident analysis input parameters in the following areas: core kinetic characteristics, control'rod worths, and core peaking factors. Cycle 10 parameters in each of these areas were examined, as discussed below, to ascertain whether new accident analyses were required.

3.2.1 Kinetics Parameters A comparison of Cycle 10 core physics parameters with current evaluation limits is given in Table 2. All the kinetics values, except the Doppler temperature coefficient (see Table 2), remain within the bounds of the analysis limits.

The least negative Doppler temperature coefficient is -1. 1 pcm/-'F compared to a limit of -1.4 pcm/'F. This coefficient is used in conjunction with the Doppler power coefficient to provide a correction to the power coefficient for fuel temperature changes in transients where the core water temperature changes. This difference, however, has been determined to result in a negligible effect on the accident analyses.

3.2.2 Control Rod Morths Changes in control rod worths may affect differential rod worths, shutdown margin, ejected worths, and trip reactivity. Table 2 shows thai the maximum differential rod worth of two RCCA control banks moving together in their highest worth region for Cycle 10 is less than or equal to the analysis limit. Table 3 shows that the Cycle 10 shutdown margin requirements are satisfied. Cycle 10 ejected rod worths are w>thin the bounds of the analysis limits.

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3.2.3 Core Peaking Factors Evaluation of peaking factors for the rod out of position and dropped RCCA incidents shows that the ONBR is maintained above the appropriate safety analysis limit ONBR value listed in Section 2.3. The peaking factors for the dropped RCCA incidents were evaluated, based on the approved new dropped rod methodology'15 The hypothetical steamline break transients were evaluated for Cycle 10. This evaluation showed that the Cycle 10 peaking factors are within the bounds of the previous analysis, and ONBR limits (See Section 2.3) are satisfied.

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4.0 REFERENCES

1. Bordelon, F. M. et al., "Westinghouse Reload Safety Evaluation Methodology," WCAP-9273, March 1978.

2. Letter'from D. L. Wigginton (USNRC) to John Dolan ( IIEMECo), dated June 10, 1987,

Subject:

NRC Approval of Amendment"Nos. 111 and 94 to Cook Unit 1 and 2 Technical Specifications.

3. Updated Final Safety Analysis Report - D. C. Cook Unit Number 1, Docket Number 50-315, updated through 1986.

4. Chelemer, H. et al., Improved Thermal Design Procedure," WCAP-8567, July 1975.

5. Letter from E. P. Rahe, Jr. (Westinghouse) to L. E. Phillips (NRC) dated April 12, 1984, NS-EPR-2893,

Subject:

Fuel Handling Load Criteria (6g vs.

4g)

6. Miller, J. V. (Ed), "Improved Analytical Model Used in Westinghouse Fuel Rod Design Computations," WCAP-8785, October 1976.

7. George, R. A., et al., "Revised Clad Flattening Model," WCAP-8377 Proprietary) and WCAP-8381 (Non-Proprietary), July 1974.

8. Risher, D. H. et.al., "Safety Analysis for the Revised Fuel Rod Internal Pressure Design Basis," WCAP-8964-A, August, 1978.

9. Skaritka, J., (Ed.), "Operational Experience with Westinghouse Cores (through December 31, 1986)," WCAP-8183, Revision 15, June 1987.

10. Liu, Y. S., et. al., "ANC: A Westinghouse Advanced Nodal Computer Code,"

WCAP-10966-A, dated September 1986.

11. NRC Letter from C. Berlinger (NRC) to E. P. Rahe, (Westinghouse)

"Acceptance for Referencing of Licensing Topical Report WCAP-10965-P, dated June 23, 1986.

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12. Hestinghouse Letter dated March 25, 1986, NS-NRC-86-3116, "Westinghouse Response to Additional Request on WCAP-9226-P/HCAP-9227-N-P, "Reactor Core Response to Excessive Secondary Steam Release," (Non-Proprietary).

13. Letter from M. P. Alexich ( Indiana and Michigan Electric Co.) to H. R. Denton (USNRC), letter AEP:NRC:091GH dated March 26, 1987,

Subject:

Proposed Technical Specification for Unit 1 Cycle 10 Reload and Related Unit 2 Proposals, Docket Nos. 50-315 and 50-316.

14. Hellman, J. M. (Ed), "Fuel Densification Experimental Results and Model for Reactor Operation," MCAP-8319-A, March 1975.

.15. Morita, T., et. al., "Dropped Rod Methodology for Negative Flux Rate Trip Plants," WCAP-10298-A, June 1983.

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TABLE 1 FUEL ASSEMBLY DESIGN PARAMETERS D. C. COOK UNIT 1

- CYCLE 10

~Re ion 10A 10B 'lA 11B 12A 12B Enrichment (w/o of U 235)" 3.297 3.598 3.404 3.600 3.298 3.600 Density (percent theoretical)" 95.143 95.295 94.995 95.042 95.286 94.911 Number of Assemblies 12 22 47 32 48 32 Burnup at Beginning of 28798 27073 19935 17103 0 Cycle 10 (MHD/MTU)**

Fuel Stack Height (inches, cold) 144 144 144 144 144 144 All values are as-built.

• • Assumes a Cycle 9 actual core average burnup of 16,096 MMD/MTU 3916F:6-670810

TABLE 2 KINET I CS CHARACTERISTICS D.C. COOK UNIT 1 - CYCLE 10 Reference Analysis Values 3 13 ~Cele 10 Moderator Temperature +5.0 (<70% RTP) +5.0 (<70% RTP)

Coefficient, (PCM/'F)" +5.0 linear ramp to 0.0. 0 0 (>70% RTP) from 70 to 100% RTP Doppler Coefficient -2.9 to -1.4 -2.9 to -1.1 (PCM/'F)

Delayed Neutron Fraction 0. 44 to 0. 75 0.44 to 0.75 Sef f (Percent)

Maximum Prompt Neutron < 26 Lifetime (p sec)

Maximum Differential Rod North 75 < 75 of Two Banks Moving Together at HZP

. (PCM/sec)*

• 1 PCM = 1.0 x 10 ap 12 3916F:6-670819

TABLE 3 SHUTDOWN REQUIREMENTS AND MARGINS D. C. COOK UNIT 1 - CYCLE 10 C cle 10'OC BOC Control Rod Worth ercent hp All Rods Inserted Less Horst 6.381 6.959 Stuck Rod (A) Less 10% 5.743 6.263 Control Rod Re uirements ercent hp Reactivity Defects (Doppler, 1.774 2.848 T $ V o i d, R e d i s t r i b u t i o n )

av Rod Insertion Allowance 0.50 0.50 (8) Total Requirements 2.274 3.348 Shutdown Mar in [ A - B ] 3.469 2.915 ercent hp Re uired Shutdown Mar in 1.6 1.6 ercent hp 13 3916F:5-6 70810

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FIGURE 1 D. C. COOK UNIT 1 CYCLE 10 REFERENCE LOADING PATTERN R P N M L K J H G F E 0 C B A 180 108 12A 128 108 128 12A 108 ifA 118 128 118 128 11A 128 118 118 liA 12 10A 108 128 1 1A 12A ifB iiA 118 12A iiA 108 10A 12 SS 12 118 12A 1 iA 12A fiA 12A 11A 12A 118 128 118 12 12 10A 128 11A 12A iOA f2A fOB ifA 108 12A 10A 12A ffA f28 10A 12 12 12 SS 12A ifB 12A ifA f2A 108 12A 11A 12A 108 12A 1 lA 118 12A 12 12 12 12 12A 108 12A 118 iiA 118 108 12A ffB 128 12 8 8 12 90 11A 1 lA ifA 11A 11A 11A 11A 11A lfA 11A 11A 11A 11A 1 iA 11A 128 118 12A 108 12A 118 fiA 118 12A 108 12A 118 12 8 8 12 118 f2A 11A 12A 1 08 12A 11A 12A 108 lfA 12A 118 12 12 12 12 10A 128 lfA 12A 10A 12A 108 iiA 108 12A iOA 12A 11A iOA SS 12 12 12 118 128 118 12A 11A 12A. iiA ifA 118 128 118 12 8 12 10A 108 128 11A 12A 118 1 1A 118 1 iA 128 108 iOA 12 SS 12 flA 118 f28 118 128 ifA 118 118 flA 12 12 108 12A 128 108 128 12A 108 0

X .- Region Number Y/SS~ Number of Fresh Burnable Absorbers/

Location ot Seconder y Source Rods 14

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FIGURE 2 HEAT FLUX HOT CHANNEL FACTOR NORMALIZED OPERATING ENVELOPE FOUR LOOP OPERATION, FQ ECCS LIMIT

= 2.10 1.2 8.8 1.8 6.8 1.8 1.0 .183 8.935

.00 12.8 8.714 0

Q .00

.40

0. 00 0 10 12 CORE HEIGHT (Ft) 15

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