ML20148H100

Safety Evaluation Supporting Util 871113 Proposed Tech Spec Changes Reflecting Unit 2 Cycle 2 Refueling & Addition of Boron Dilution Mitigation Sys for Unit 2

ML20148H100
Person / Time
Site:	Catawba
Issue date:	01/22/1988
From:	Office of Nuclear Reactor Regulation
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ML20148H096	List: ML20148H092 ML20148H100
References
NUDOCS 8801270128
Download: ML20148H100 (6)

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SAFETY EVALUATION BY THE 3FFICE OF NUCLEAR REACTOR REGULATION RELATING TO TECHNICAL SPECIFICATION CHANGES AND UNIT 2/ CYCLE 2 RELOAD DUKE POWER COMPANY CATAWBA UNITS 1 AND 2 DOCKET NOS. 50-413 AND 50-414 INTRODUCTION By letter oated November 13, 1987 (Ref. 1), Duke Power Company (the 'icensee) requested Changes to the Technical Specifications for Catawba Nuclear Station Units 1 and 2, to reflect the Unit 2 refueling and the addition of the Boron Dilution Mitigation System for Unit 2. In addition deletion of Surveillance 4.3.3.12.1(b) which requires an ANALOG CHANNEL OPERATIONAL CHECK prior to startup is requested for both units. A second letter dated December 11, 1987 (Ref 2) provided a discussion of the Justification and No Significant Hazards Considerations. Additional infonnation and justification were provided in letters dated January 15 and 20, 1988 (Refs. 10 and 11).

EVALUATION A. Unit 2 Cycle 2 Reload

1. General Design The Catawba Unit 2, Cycle 2. reactor core contains 193 Optimized Fuel Assemblies. During the Cycle 1/2 refueling 64 Region 1 fuel assemblies will be replaced with 64 Region 4 fuel assemblies. The Region 4 fuel is very similar to that used in Regions 1, 2, and 3.

Region 4 fuel assemblies have a smaller rod plenum spring than those used in Regior;s 1, 2, and 3. This new spring design is being generally incorporated by Westinghouse and the justification was submitted in Reference 3. The Region 4 fuel has been designed according to the fuel performance model in WCAP 8785 (Ref. 4). Tne fuel is designed and operated so that clad flattenin as provided by the Westinghouse model in WCAP Ref 5).8377 For(g allwill not occ fuel regions, the fuel rod internal pressure design basis, which is discussed and shown acceptable in WCAP-8964 (Ref. 6) is satisfied, 1 The licensee provided a 2eload Safety Evaluation (RSE) for Catawba 2, l Cycle 2, as an attachment to Reference 1. The RSE pre:ents a l Cycle-specific evaluation for Cycle 2 which demonstrates that the l core reload will not adversely affect the safety of the plant. This l evaluation was performed utilizing the approved reload design l t

methods of WCAP-9273-U-A (Ref. 7).

l 8801270128 880122 PDR ADOCK 05000413-P PDR

2. Nuclear Design The Cycle 2 Core loading is designed to meet an [Fn(Z)xP] ECCS limit of less than or equal to 2.32xK(Z). Adherence to Yhe Fn limit is obtained by using the F TS surveillance described in WCAP-10217-A (Ref. 8). F surveillabceispartoftheRelaxedAxialOffsetControl (RA0C) and rkplaces the previous F surveillance by~ comparing a measuredFnlimit.ThisprovidesaMoreconvenientformofassuring plant operNtion below then F limit while retaining the intent of using a measured parameter Yo verify operation below TS limits. The above discussion is consistent with Reference 8 which was approved.

Thus, the staff finds that the TS change to Fg surveillance is acceptable.

RAOC will be employed in Cycle 2 to enhance operational flexibility during non steady state operation. RA0C makes use of available margin by expanding the allowable di band, particularly at reduced power. ROAC is described in Reference 8 and was approved by the staff. Thus, it is acceptable for use in Catawba Unit 2.

During operation at or near steady state equilibrium conditions core peaking factors are significantly reduced due to the limited amount of xenon skewing possible under these operating conditions. The licensee proposes to use Base load TS to recognize this reduction in core peaking factors. The proposed Base load TS are identical to those that the staff has previously approved for McGuire Units 1 and 2, and Catawba Unit 1 and are therefore acceptable.

The RSE provides a table of Cycle 2 kinetics characteristics which are compared with the current limits based on previously approved accident analyses. The RSE also provides a table showing the results of the calculated Cycle 2 control rod worths and requirements ct the most limiting condition during the cycle (end-of-life). These results include a standard 10% allowance for calculational uncertainty, From this information, the staff concludes that sufficient control rod worth will be available to provide the required shutdown margin for Cycle 2 operation. Control rod insertion limits were increased for less than 100% power for Cycle 2. Since the required shutdown margin is maintained, the TS change proposed to reflect the increased insertion is acceptable.

3. Thermal and Hydraulic Design The thermal hydraulic methodology, DNBR correlation and core DNB limits used for Cycle 2 are consistent with the current licensing basis described in the FSAR and approved by the staff.

The power distributions produced by the cycle-specific RA0C analysis were analyzed for normal operation and Condition II events, limits on the 4 allowable operating flux difference as a function of power level from these considerations were found to be less restrictive than those l

resulting from LOCA Fn considerations. The Condition 11 analyses generate DNB core limYts and resultant Over-Temperature Delta-T setpoints. These generated a change to the F(41) function in the TS. The change is acceptable because it results from cycle-specific calculations using approved methods (Refs 7 and 8). Therefore, the staff concludes that the Cycle 2 thermal-hydraulic analysis is acceptable.

4. Accident Analysis The effects of the reload on the design basis and postulated accidents analyzed in the FSAR were examined. in all cases it was found that the effects were accommodated within the conservatism of the initial assumptions used in the previous applicable safety analysis, the safety evaluation performed in support of the RTD Bypass Elimination licensing submittal and the safety evaluation performed in support of the UHI Elimination licensing submittal.

(Refs.1, 2 and 111 A core reload can affect accident analysis input parameters through control rod worths, core peaking factors and core kinetic characteristics. The Cycle 2 parameters in each of these areas were examined and found to be within the bounds of the current limits.

5. Technical Specification Changes The Technical Specification changes for the Unit 2 Cycle 2 Reload I are:

1. RA0C and Axial Flux Difference Limits i

2. Fg Surveillance

3. Base load Technical Specifications  ;

4. Rod Insertion Limits

5. OT AT f3 (61)

Acceptability of items 1 - 4 was discussed in Section 2 Nuclear Design. Acceptability of item 5 was discussed in Section 3, thermal and hydraulic design. The proposed changes are for Unit 2 only but the actuel change pages involve both Units. The revisions to the bases are also acceptable.

B. Boron Dilution Mitigatien Syst_em 1.0 Introduction The Boron Oilution Mitigation System (BDMS) which is being installed in Unit 2 is the same as the BDMS which was installed on Unit 1. The BDMS was described in letters dated June 6, 1986 and September 9, 1986 (Refs.

12 and 13).

2. Technical Specification Changes The changes for Unit'2 which deal with the Boron Oilution Mitigation System (BDMS) are to Specifications 4.1.1.1.3, 4.1.1.1.4, 4.1.1.2.2, Table 3. 3-1; item 6.b. Table 3.3-1; Action 5, Table 4.3-1; Note (9), 3/4.3.3.12, and 3/4.9.2. Changes to Specifications 4.3.3.12.1(b), 3.9.2.1 and 4.9.1.3 apply to both Units. Each change is discussed below.

The changes that apply to Unit 2 only are identical to the changes made to the Unit 1 Technical Specifications when the Boron Dilution Mitigation System was installed on that Unit. They were reviewed and approved at that time (Ref 9). It was requested that these changes not apply to Unit 2 until after the BDMS system has been calibrated, tested and declared operable. The licensee had stated (Ref.10) that all the Technical Specifications applicable to the boron dilution accident which are to be deleted will be adminstratively maintained in this interim period. We find this acceptable.

Specification 4.3.3.12.1(8)

This specification will be deleted since it requires surveillance prior to Mode 2 but the Specification itself is not applicable in Modes 1 and 2. We find this change acceptable.

Specification 3.9.2.1 This change is an editorial change which deletes a phrase "and control room" which appeared twice in the sentence. Thus it is acceptable.

Specification 3.9.1.3 This specification verifies that potential boron dilution flow paths are isolated when the unit is in Mode 6. The deletion of TS 3.9.1.3 is acceptable since the BDMS provides for automatic isolation of potential boron dilution flow paths.

CONCLUSION We have concluded, based on the-considerations discussed above, that: (1) there is reasonable assurance that the health and safet of the public will not be endangered by operation ir, the proposed manr.er 2 and (y) such activities will be conducted in compliance with the Comission's regulations and the issuance  :

of these amendments will not be inimical to the common defense and security or the health and safety of the public.

References

1) letter from Hal B. Tucker (Duke Power Companyl to NRC, November 13, 1987

2) letter from Hal B. Tucker (Duke Power Company) to NRC, December 11, 1987

3) letter from E. P. Rahe Jr. (Westinghouse) to L. E. Phillps (NRC), April 12, 1984, NS-EPR-2893

Subject:

Fuel Handling load Curtain (6g vs 49)

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

5) George, R. A., (et al.), "Revised Clad Flattening Model, WCAP-8377, July, 1977

6) Risher, D. H., (et al.), "Safety Analyses for the Reused Fuel Rod Internal Pressure Design Basis," WCAP-8964, June, 1977

7) Bordelon, F. M., (et all, "Westinghouse Reload Safety Evaluation Methodology", WCAP-9273-A, July, 1985

8) Muller, R. W. , (et al.), "Relaxation of Constant Axial Offset Control-FnSurveillance Technical Specification, "WCAP-10217-A, June, 1983

9) Mercrandum to Kahtan Jabbour, NRC, from Charles E. Rossi, NRC, October 1, 1986, (TAC 61740)

10) Letter frw Hal B. Tucker (Duke Power Company) to NRC, January 20, 1988

11) letter from Hal B. Tucker (Duke Power Company) to NRC, January 15, 1988

12) Letter from Hal 8. Tucker (Duke Power Company) to Harold Denton (NRC),

June 6, 1986

13) Letter from Hal B. Tucker (Duke Power Company) to Harold Denton (NRC),

September 9, 1986 l

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ENCLOSURE 2 Functional Areas

1. Management Involvement in Assuring Quality.

Rating: Category 3

2. Approach to Resolution of Technical Issues from a Safety Standpoint.

Rating: Category 3

3. Responsiveness to NRC Initiatives Rating: Category 3

4. Enforcement History N/A

5. Operational and Construction Events N/A

6. Staffing (including Management)

N/A

7. Training and Qualification Effectiveness N/A i

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