ML15112A862

Safety Evaluation Supporting Amends 73,73 & 70 to Licenses DPR-38,DPR-47 & DPR-55,respectively

ML15112A862
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Site:	Oconee
Issue date:	06/22/1979
From:	Office of Nuclear Reactor Regulation
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ML15112A861	List: ML15112A860 ML15112A862 ML19225B108
References
SER-790622, NUDOCS 7907230517
Download: ML15112A862 (7)

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l RE 0o UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 73 TO FACILITY OPERATING LICENSE NO. DPR-38, AMENDMENT NO. 73 TO FACILITY OPERATING LICENSE NO. DPR-47, AND AMENDMENT NO. 70 TO FACILITY OPERATING LICENSE NO. DPR-55 DUKE POWER COMPANY OCONEE NUCLEAR STATION, UNITS NOS. 1, 2 AND 3 1.0 Introduction By letter dated March 30, 1979 (Reference 1) and supplement dated May 17, 1979(2),Duke Power Company (DPC) has proposed changes to the Oconee Nuclear Station Technical Specifications. The proposed changes include modifications for Oconee Nuclear-Station Unit 3 (ONS-3) operation after reload for Cycle 5 and page number modifi cations for administrative purposes.

Most of the proposed Technical Specification modifications and review effort have been associated with the refueling of ONS-3 for Cycle 5 operation.

The information submitted by DPC in connection with this refueling is presented in Reference 1 which describes the fuel system design, nuclear design, thermal-hydraulic design, accident analyses, and startup test program. The referenced supplement provides confir mation that the presented analysis and specification of Reference 1 are applicable for actual previous cycle exposure.

The refueling of ONS-3 for Cycle 5 will result in a core loading of 56 fresh Mark B4 assemblies, 108 previously burned Mark B4 assemblies, and nine previously burned Mark B2 fuel assemblies. For ONS-3, this evaluation has taken into consideration the proposed refueling of the core as described in Reference 3 and subsequent operation for the targeted 292 effective full power days (EFPDs) during Cycle 5.

The changes in the core loading and mode of operation are the only physical modifications associated with the refueling. The evaluation of DPC's proposed modifications to the Technical Specifications of ONS-1, 2 and 3 is presented in the following sections.

-2 2.0 Evaluation of Modifications to ONS-3 Core Design 2.1 Fuel System Design We have evaluated the proposed fuel loading and operation. Tables 4-1 and 4-2 of Reference 3 summarize the design characteristics of the reload fuel types. The fresh Mark B4 assemblies are identical to the previously burned Mark B4 fuel with regard to assembly mechanical design, fuel rod design and thermal design.

The fuel designs of Mark B4 fuel types have been evaluated for ONS-3 in association with earlier refuelings and found acceptable (References 4, 5 and 6).

2.1.1 Cladding Creep Collapse Fuel rod cladding creep collapse analyses have been performed for the most limiting (i.e., most highly exposed) fuel assemblies to be included for Cycle 5. The analyses were performed according to the conservative methods and assumptions described in Reference 7 which has been accepted by the NRC staff. (This reference is a proprietary version, but nonproprietary versions are available also.)

These anal yses show that the time to rod cladding collapse will be in excess of 30,000 effective full power hours (EFPHs). Because no assembly will reach a total exposure as high as 30,000 EFPH during Cycle 5 (Table 4-1 of Reference 3), we conclude that cladding creep collapse will not occur during the cycle.

2.1.2 Cladding Stress and Strain For this cycle, the cladding stress due to differential pressure, temperature gradient or axial loads and restraints will not exceed the yield stress or ultimate strength of the material. The antici pated cladding strain was shown to be less than the 1% plastic cladding strain limit for up to 55,000 MWd/MTU, well below the exposure to be accumulated by the end of cycle. We previously accepted these criteria for cladding stress (Reference 8) and strain (Reference 6) and we conclude that they are also valid for this cycle.

2.1.3 Fuel Thermal Design The thermal linear heat rate (LHR) limits have been established with the TAFY Code (Reference 9) and assumed fuel densification to 96.5% of theoretical density. These limits are stated in Table 4-2 of Refer ence 2. The thermal LHR limits which ensure that fuel center melting does not occur are less restrictive than the Loss of Coolant Accident (LOCA) LHR limits. Because the LOCA LHR limits will be met by operating within the limiting conditions for operations, the thermal LHR limits will also be met. We conclude that the indicated thermal LHR limits are acceptable for preventing center melting and that the limits will not be exceeded.

-3 2.2 Nuclear Design Reference 3 indicates the proposed core loading arrangement, the initial enrichments and burnup distributions. Most of the fresh Mark B4 assemblies will be loaded into locations on the edge of the core and will be below fuel thermal limits.

Reactivity control will be supplied by soluable boron in the reactor coolant which will be supplemented by 61 full length control rods.

Also, APSRs will provide axial power distribution control profile.

Nuclear parameters, e.g., crifical boron concentrations, control rod worths, Doppler coefficients, moderator coefficients, xenon worth and effective delayed neutron fractions have been calculated using the same techniques as accepted for the previous cycle in Reference 6. These parameters are presented and compared in Reference 3 to the previous cycle values.

Shutdown margins have been calculated for beginning of cycle (BOC) and end of cycle (EOC).

The calculated minimum shutdown margin is larger than the required value.

We conclude that the nuclear design does not differ in a significant way from earlier cycles, that the nuclear parameters have been calcu lated by acceptable methods and are within the range of values expected for a cycle approaching an equilibrium cycle, and that the nuclear design has resulted in an adequate shutdown margin. The nuclear design for ONS-3 Cycle 5 is, therefore, acceptable.

2.3 Thermal Hydraulic Design The new fuel is thermal-hydraulically identical to that currently in use.

The thermal-hydraulic design evaluation used methods and models pre viously accepted in References 5 and 6. The results of this evaluation are included in Table 6-1 of Reference 3.

The flux/flow trip setpoint has been revised for this cycle to 1.08 from 1.05 of the previous cycle. This modification is based on the transient analysis of the loss of two reactor coolant pumps. The methods used are identical to those for previous cycle evaluations as found acceptable in References 4 through 6. The only change to this thermal hydraulic analysis is the use of reactor coolant pump flow rate coastdown values which are specific to this plant. Pre viously generic values were used. This revision and revisions to the FAH Technical Specification (Reference 5) provide sufficient margin to adjust the flux flow trip setpoint and still maintain reactor safety limits, e.g., Departure from Nucleate Boiling Ratio (DNBR) equal to or less than a ratio of 1.3.

-4 The licensee accounted for the reduction in DNBR due to fuel rod bowing based on Reference 10, which is a Babcock & Wilcox (B&W) interim method. It is the present staff position that a future modification to the methods of Reference 10 are required. The licensee has included a 10.2% DNBR margin in the setpoint calcula tions for Oconee Unit No. 3. We consider this sufficient to account for the increased reduction in DNBR due to future modifi cations to the methods described in Reference 10. Therefore, we consider the licensee'fuel rod bowing calculations to be acceptable.

3.0 Evaluation of Accidents and Transients Each Final Safety Analysis Report (FSAR) accident analysis (Reference

11) has been examined with respect to cycle-dependent parameter changes to determine the effect of the reload and to ensure that reactor performance during hypothetical transients is not degraded.

Fuel thermal analysis parameters are given in Reference 3. That reference compares the Cycle 4 and 5 thermal-hydraulic maximum design conditions. Reference 2 compares the key kinetics parameters from the FSAR and Cycle 5.

From the examination of Cycle 5 core thermal properties and kinetics properties with respect to acceptable previous cycle values, it is concluded that this core reload will not adversely affect the safe operation of ONS-3 during Cycle 5. The only parameter which is potentially less conservative for Cycle 5 than for the FSAR value is the Doppler coefficient. We have estimated the effect of this non conservatism for the FSAR transients and have concluded that there is sufficient conservatism in the FSAR analyses to compensate for this potential nonconservatism.

4.0 Emergency Core Cooling System (ECCS)

On July 9, 1975, DPC submitted an acceptable ECCS evaluation (Refer ence 12) for ONS-3. On April 12, 1978, we were informed of a potentially more severe limiting small break location than previously analyzed. By a Commission Order for Modification of License dated April 26, 1978, certain modifications to the ECCS and a procedure for prompt operator action were required for ONS-3 to permit future operation. An Exemption was granted on July 6, 1978 to 10 CFR 50.46(a),

"Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors," which superseded the Order. The Exemption provided for its own termination upon completion of the modifications required by the Exemption and for prior NRC staff approval of the design.

By letter dated December 13, 1978, we found the design of the modifications to be acceptable. DPC has installed the modifi cations and prepared acceptable operating procedures; thus, we conclude that the as-modified ECCS required by the Exemption of July 6, 1978 is acceptable.

-5 Due to the accident at Three Mile Island, Unit No. 2 (TMI-2) on March 28, 1979, where a pilot-operated relief valve on the primary system failed to close after opening, thus inducing the equivalent of a very small break LOCA, in conjunction with a complete loss of feedwater, the NRC is reexamining the ECCS of all pressurized water reactors in terms of very small breaks.

Our Order of May 7, 1979, issued in the aftermath of TMI-2, required DPC to:

"Complete analyses for potential small breaks and implement operating instructions to define operator action."

B&W, by a report dated May 1979 and a supplementary letter dated May 12, 1979, re sponded on the Oconee dockets to the requirements of our Order. Our May 18, 1979 Safety Evaluation of these submittals stated, "A principal finding of our generic review is a reconfirmation that LOCA analyses of breaks at the lower end of the small break spectrum (smaller than 0.04 sq. ft.) demonstrate that a combination of heat removal by the steam generators, high pressure injection system and operator action ensure adequate core cooling."

We concluded that DPC complied with the analysis portion of the quoted paragraph of the Order. We further stated that to support longer term operation of the Oconee Station, requirements will be developed for additional and more detailed analyses of loss of feedwater and small break LOCA events.

We concluded in our letter of May 18, 1979, that DPC could restart ONS-3 in that it met the requirements of our May 7, 1979 Order.

Based on the above, we conclude that the emergency core cooling system for Oconee Unit No. 3 is acceptable.

5.0 Startup Tests Startup tests have been proposed by DPC to provide assurance that ONS-3 has been loaded as intended. This test program is very similar to that used for ONS-3 Cycle 4. We have reviewed the test program and consider it acceptable.

6.0 Evaluation of Technical Specification Changes The Technical Specifications have been revised for Cycle 5 operation in accordance with the methods of Technical Specification bases to account for minor changes in power peaking and control rod worths inherent with non-equilibrium cycles. In addition, the power level cutoff restriction applied in previous cycles to the control rod

-6 position limits has been revised, not only for ONS-3 but also for ONS-1 and 2 operation. The change has been accomplished by desig nating the power level cut-off at 100% full power. Any operating restrictions resulting from transient xenon-induced power peaks, including xenon-free startup, are inherently included in the control rod position and axial imbalance limits. The remaining changes have been discussed in the previous text. We have reviewed these changes and found them acceptable.

7.0 Environmental Consideration We have determined that these amendments do not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact.

Having made this determination, we have further concluded that these amendments involve an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR §51.5(d)(4), that an environmental impact statement, or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of these amendments.

8.0 Conclusion We have concluded, based on the considerations discussed above, that:

(1) because the amendments do not involve a significant increase in the probability or consequences of accidents previously considered and do not involve a sinificant decrease in a safety margin, the amendments do not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Conmission's regulations and the issuance of these amendmentswill not be inimical to the common defense and security or to the health and safety of the public.

Dated:

June 22, 1979

-7 References

1. Letter from William 0. Parker, Jr., DPC, to Harold R. Denton, NRC, March 30, 1979.

2.

Letter from William 0. Parker, Jr., DPC, to Harold R. Denton, NRC, May 17, 1979.

3. Oconee Unit 3 Cycle 5 Reload Report, BAW 1522, March 1979.

4. Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendments Nos. 34, 34 and 31 to Facility Operating Licenses Nos. DPR-38, DPR-47, and DPR-55, Duke Power Company, Oconee Nuclear Station 3, October 22, 1976.

5. Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendments Nos. 52, 52 and 49 to Facility Operating Licenses Nos. DPR-38, DPR-47, and DPR-55, Duke Power Company, Oconee Nuclear Station 3, November 21, 1977.

6. Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendments Nos. 63, 63 and 60 to Facility Operating Licenses Nos. DPR-38, DPR-47, and DPR-55, Duke Power Company, Oconee Nuclear Station 3, July 6, 1978.

7. Program to Determine In-Reactor Performance of B&W Fuels Cladding Creep Collapse, BAW-10084P-A, Rev. 2, Babcock & Wilcox, Lynchburg, Va., January 1979.

8. Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendments Nos. 45, 45 and 42 to Facility Operating Licenses Nos. DPR-38, DPR-47, and DPR-55, Duke Power Company, Oconee Nuclear Station 3, July 29, 1977,

9. C. D. Morgan and H. S. Kao, TAFY -

Fuel Pin Temperature and Gas Pressure Analysis, BAW-10044, Babcock & Wilcox, Lynchburg, Va.,

May 1972.

10. Letter from J. H. Taylor, B&W, to D. B. Vassallo, NRC, "Determination of the Fuel Rod Bow DNB Penalty," December 13, 1978.

11.

Oconee Nuclear Station, Units 1, 2, and 3 Final Safety Analysis Report, Docket Nos. 50-269, 50-270, and 50-287.

12. R. C. Jones, J. R. BiTler, and B. M. Dunn, ECCS Analysis of B&W's 177-FA Lowered Loop NSS, BAW-10103A, Rev. 3, Babcock

& Wilcox, Lynchburg, Va.