ML19309G485
| ML19309G485 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 04/16/1980 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19309G480 | List: |
| References | |
| NUDOCS 8005070037 | |
| Download: ML19309G485 (5) | |
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j. f, % ik j NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555
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%.m.s SAFETY EVAll'ATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMEN 0 MENT NO. 42 TO FACILITY OPERATING LICENSE NO. OPR-25 COMMONWEALTH EDISON COMPANY ORESDEN NUCLEAR POWER STATION, UNIT NO. 3 DOCKET NO. 50-249 Introduction By letter dated December 10,1979 (Reference 1), and supplemented by References 2 and 8, Commonwealth Edison (CE), the licensee, proposed amendments to Dresden Unit 3 License and Appendix A Technical Specifica-tions. CE has proposed these amendments to support its review of future reloads for Dresden Unit 3 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' l.06.
These limits have previously been found acceptable for this use in Reference 4 and on this basis the proposed change is acceptable.
Rod Droo 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 /gn: design limit is accept-able 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 acceptable.
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Maximum Average Planar Linear Heat Generation Rate (MAPLHGR)
New MAPLHGR curves reflecting the improved flooding characteristics of retrofit 8x8 fuel have been proposed by the licenser. Curves for 8x8, 8x8 retrofit, and 7x7 fuel of the various enrichments anticipated for future Dresden Unit 3 reloads and extending to burnups of 40,000 mwd /t have been proposed (i:eferences 1, 6, and 8).
Based on our previous approval of MAPLHGR curves reflectir.g 8x8 retrofit fuel reflood characteristics (Reference 10) and extension of burnup to 40,000 mwd /t (Reference 9), the licensee's proposed changes are acceptable.
However, because the.new curves are based on an assumed fuel loading with 156 retrofit assemblies, any reload with fewer such assemblies will be outside the scope of this approval.
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 by an equivalent amount.
This change has been genericclly accepted by our Reference 5 letter.
Power Peaking The licensee has proposed to change from a total peaking factor formulation 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 results but the proposed formulation eliminates the need for different limits for different fuel types. The limitations will be applied above 25% rated thermal power which is consistent with the LHGR surveillance requirements and the Standardized Technical Specifi-cations. This is acceptable.
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 SRY safety function set-point 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.
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. Water Level Setpoints The curront Technical Specifications have water level setpoints referenced to the top of the active fuel. Different active fuel lengths, as is the case for the Sx8 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.
These findings are based on current acceptable fuel assembly designs and any
, application of other designs would require specific justification of water level setpoints.
The reactor icw water level scram setting and ECCS initiation setting have been established as 144 inches and 84 inches above the top of the core, respectively, to assure that the scram occurs at or before the value assumed in reactor safety analyses.
Overpressure Protection Margin to Safety Valve Setpoint CE has proposed to delete the portion of the license restriction that requires reactor power level restrictions to maintain pressure margin to safety valve (SV) setpoints during the worst case pressurization 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 overpres-surization protection (Standard Review Plan 5.2.2, NUREG-79/087) have been that "for the desigr. basis nomal operational transients, relief valve capacity must be sufficient to limit the pressure so as to prevent SV dis-charge directly to the containment," and "for the most severe abnormal operational transient, with reactor scram, the SV capacity should be suf-ficient to limit the pressure to less than 110% of the reacto= coolant pressure boundary jesign pressure." These criteria are satisvied 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 pcssible SV discharge. We have also reviewed BWR pres-sure relief systems operating experience (NUREG-0462) and have foun that operating experience with SVs has been essentially failure free.
Coastdown Feedwater Heater Restrictions The licensee has proposed that the license restriction include a requirement to perform a safety evaluation if off-normal feedwater heater operation is needed. We consider this restriction appropriate.
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. Typographical 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 Consideration We have detennined 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 detemination, we have further concl*uded that the amendment involves an action which is insignificant 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 this amendment.
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 significant decrease in a safety margin, the amendment I
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
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will be conducted in compliance with the Commission's regulations and
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the issuance of the amendment will not be inimical to the comon defense and security or to the health and safety of the public.
Dated: April 16, 1980
>... References 1.
Letter from D. L. Peoples (CE) to the Director of Nuclear Reactor Regulation (USNRC), dated December 10, 1979.
2.
Letter from Rc bert F. Janecek (CE) to Director of Nuclear Reactor Regulation (USNRC), dated February 6,1979.
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 (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 D. L. Peoples (CE) to the Director of Nuclear Reactor Regulation (USNRC), dated March 24, 1980 9.
Letter from T. A. Ippolito (NRC) to D. L. Peoples (CE), dated December 28, 1979 10 Letter from D. L. Ziemann (NRC) to Cordell Reid (CE), dated April 24,1979.
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