ML20126C346

Safety Evaluation Supporting Amend 51 to License DPR-30

ML20126C346
Person / Time
Site:	Quad Cities
Issue date:	03/20/1980
From:	Office of Nuclear Reactor Regulation
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ML20126C340	List: ML20126C337 ML20126C342 ML20126C346 ML20126C350
References
NUDOCS 8003310192
Download: ML20126C346 (5)

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SAFETY EVALUATION BY THE OFFICE . NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 51 TO FACILITY OPERATING LICENSE NO, DPR-30 COMMONWEALTH EDISON COMPANY 1 AND -

IOWA-ILLINOIS GAS AND ELECTRIC COMPANY QUAD CITIES STATION, UNIT NO. 2 DOCKET NO. 50-265 Introduction By letter dated August 30,1979 (Reference 1), and supplemented by Reference 2 and Reference 9, Commonwealth Edison (CE), the licensee, proposed amendments to Quad Cities Station Unit 2 (QC-2) License and Appendix A, Technical Specifications. CE has proposed these amendments to support its review of future reloads for QC-2 under the provisions of 10 CFR 50.59.

Our approval is only for the proposed amendment and does not constitute approval of CE's future reloads under the provisions of 10 CFR 50.59.

Evaluation Safety Limit Critical Power Ratio (SLMCPR[

This change provides SLMCPRs in the Technical Specifications for all currently approved core loadings. With retrofit 8x8 fuel in the core the SLMCPR limit is specified as 1.07. Without retrofit 8x8 fuel the SLMCPR limit is 1.06.

These limits have previously been found acceptable for this use in Reference 4 and on this basis the proposed change is acceptable.

Rod Drop Accident (RDA) Design Limit The RDA design limit has been modified from 1.3%a maximum rod worth to 280 cal /gm peak fuel enthalpy rise. The 280 cal /gm design limit is acceptable per Standard Review Plan NUREG-75/087. Also, the power level below which the rod worth minimizer is required was increased from 10% to 20% of rated.

This is conservative by comparison to the previous specification and is

) consistent with reactor safety analyses. 1 l

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Maximum Average Planar Linear Heat Generation Rate (MAPLHGR)

New MAPLHGR curves from Reference 6 have been proposed. This employed methodology accepted for OC-2 in Reference 7. The only change to Reference 6 from that previously accepted is to incorporate prepressurized fuel in the analyses.

This addition was performed in compliance with our guidelines in our acceptance of prepressurized fuel application (Reference 8). Therefore, the use of Reference 6 continues to be acceptable for QC-2 and the addition of Reference 6 MAPLHGR curves to the Technical Specification is also acceptable.

Linear Heat Generation Rate (LHGR) Power Spiking Penalty The LHGR power spiking penalty for 8x8 fuel has been incorporated in the safety analyses by the reduction of the LHGR limit an equivalent amount. This change has been generically accepted by our Reference 5 letter.

Power Peaking The next proposed change is to go from a total peaking factor fonnulation to a ratio of the fraction of limiting power density to fraction of rated power formulation for the local power peaking adjustment of neutron flux reactor protective system logic. This change in formulation has previously been approved for other BWRs, e.g., Reference 3. These two formulations are identical in their resul.ts but the proposed formulation eliminates the need for different limits for different fuel t.ypes. Also the pro vides for increasing the Average Power Range Monitor (APRM) posedofchange pro-gains instead a reduction in setpoints. This results in the same protective function.

Safety / Relief Valve (SRV) Setpoint A reduction in the SRV safety function setpoint has been proposed. The SRV relief function setpoint was reduced to preclude multiple relief valves discharges. The proposed change would reduce the SRV safety function setpoint to the same value for consistency. The licensee plans to use this setpoint in future analyses. This reduction is conservative with respect to reactor vessel pressure relief function. It is also a small reduction =15 psi, so that there is no significant increase in valve demand, and, thus, probability of failure.

Water Level Setpoints The current Technical Specifications have water level setpoints referenced to the top of the active fuel. Different active fuel lengths, as is the case for the 8x8 and 8x8R fuels, may confuse the specification and surveillance requirements. Therefore, the licensee has proposed to define the top of the active fuel as 360 5/16" above reactor vessel zero and the reactor low water level scram and ECCS initiation setpoints at 143 7/8" and 83 7/8" above the top of the active fuel, respectively. These definitions and setpoints are conservative values compared to those used in the reactor safety analyses.

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-S-These findings are based on current acceotable fuel assemcly designs and any application of other designs would require specific justification of water level setooints.

The reactor low water level scram setting had been established as 144 inches above the top of the core. With the licensee's agreement, we have modified the trip setpoint to be greater than 144 inches to assure that the scram occurs at or before the value assumed in reactor safety analyses.

Overcressure Protection Margin to Safety Valve Setcoints The first procosed cnange is to delete the portien of the license restricticn that requires reactor power level restrictions to maintcin pressure margin to safety valve (SV) set points during the worst case cressurization transient.

This restriction was imposed by the licensee to avoid an extensive outage in the event of SV discharge to the drywell. Our criteria for overpressurization protection (Standard Review Plan 5.2.2, NUREG-79/087) has been that "for the design basis normal operational transients, relief valve capacity must be sufficient to limit the pressure so as to prevent SV discharge directly to the containment," and, "for the nost severe abnormal operational transient, with reactor scram, the SV capacity should be sufficient to limit the pressure to less than 110% of the reactor coolant pressure boundary design pressure."

These criteria are satisfied by the proposed change.

Further, we do not consider the SV discharge to the drywell a safety concern, since all safety systems are to be qualified for LOCA environment which is more severe than the possible SV discharge. We have also reviewed BWR pressure relief systems operating experience (NUREG-0462) and have found that operating experience with SVs has been essentially failure free.

Coastdown Feedwater Heater Restrictions In response to our concern on the potential for power operation outside the bounds of the transient analyses for coastdown operation, the licensee has proposed that the license restriction include a requirement to perform a safety evaluation if off-normal feedwater heater operation is needed (Reference 2). This requirement satisfies our concern.

Tycograonical Corrections and Clarification of Bases The remaining changes fall into the category of typographical corrections and clarification of bases and do not, as such, represent a significant safety Concern.

Environmental Considerations We have determined that the amendment does 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 the amendment involves an action which is insigni-ficant from the standpoint of environmental impact, and pursuant to 10 CFR Section 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 the amendment.

. 5 Conclusion We have concluded, based on the considerations discussed above, that: (1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a signi-ficant decrease in a safety margin, the amendment does 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 Commission's regulations and the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Dated: March 20, 1980 l

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References

1. Letter from Cordell Reed (CECO), to the Director of Nuclear Reactor Regulation (USNRC) dated August 30, 19/9.

2. Letter from D. Louis Peoples (CECO), to Director of Nuclear Reactor Regulation (USNRC) dated October 24, 1979.

3. Letter from T. A. Ippolito (USNRC) to G. T. Berry (Power Authority of the State of New York), dated November 22, 1978.

4. Letter from D. G. Eisenhut (USNRC) to R. Gridley (General Electric, GE) dated May 12, 1978.

5. Letter from D. G. Eisenhut (USNRC) to R. Gridley (GE) dated June 9, 1978.

6. " Loss-of-Coolant Accident Analysis Report for Dresden Units 2,3 and Quad Cities Units 1, 2 Nuclear Power Stations," NED0-24164 A, dated April 1979.

7. Memorandum from R. L. Baer to D. L. Ziemann " Evaluation of Quad Cities Unit 2 for Cycle 4 Operation," dated February 21, 1978.

8. Letter from T. A. Ippolito (USNRC) to R. Gridley (GE), dated April 16, 1979.

9. Letter from D. L. Peoples (CECO) to Director, NRR, dated March 7,1980.

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