ML20245A358
| ML20245A358 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 06/14/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20245A348 | List: |
| References | |
| NUDOCS 8906210259 | |
| Download: ML20245A358 (4) | |
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION rn,
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SAFETY EVALUATION BY THE OFFICE OF MUCLEAR REACTOR REGULATION REGARDING INCREASED CORE FLOW ANALYSIS PARTIAL FEEDWATER HEATING ANALYSIS AND EXTENDED LOAD LINE LIMIT ANALYSIS PHILADELPHIA ELECTRIC COMPANY LIMERICK GENERATING STATION, UNIT 2 DOCKET NO. 50-353 By letter dated May 31, 1989 (Ref. 1), Philadelphia Electric Company (PECo)
(the licensee) provided a markup of the current Final Safety Analysis Report (FSAR) for the Limerick Generating Station, Units 1 and 2.
The markup of the FSAR pages was made to incorporate the extended load line limit analysis (ELLLA),
increased core flow (ICF) and partial feedwater heating (PFH) into the FSAR so these modes of operation could be included in the draft Technical
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Specifications (TSs) for Unit 2.
To support these operational conditions and the draft revision to the FSAR, PECo also provided two reports prepared by General Electric (GE) (Ref. 2 and 3). These changes were discussed with NRC staff on May 11, 1989 and are consistent with the Final Draft version of the Unit 2 Technical Specifications transmitted by NRC letter dated May 19, 1989.
The analyses supporting these operational conditions for Unit 2 are identical to the Unit I analyses which the NRC has previously accepted by safety evaluations dated February 17, 1987 (ICF and PFH), and August 14, 1987 (ELLLA).
(Ref. 4 and 5)
The PECo submittal proposes extensions to standard operating regions in the GESTAR II standard category of " Operating Flexibility or Margin Improvement Options". The selected options are ELLLA, ICF, FFWTR and FH005. These have become commonly selected and approved options for a number of reactors in recent years. These options are described end discussed in the GE topical reports for Limerick Unit 2 which provides generic analyses of transients and accidents.
(Ref. 2 and 3)
The proposed ELLLA changes the Average Power Range Monitor (APRM) rod block and scram lines on the power-flow map, and permits operation up to the new APRM rod block line (0.58W + 50%) up to the intersection with the 100 percent power line occurring at a flow of 87 percent. These are standard changes for ELLLA. For ICF the approved flow increase is to 105 percent of rated core flow at 100 percent power. The increased flow is allowed throughout the cycle and af ter normal end-of-cycle (with or without FFWTR) with reactivity coast down. FFWTR involves feedwater temperature reduction up to 60* F (to 360 F at full power) and is proposed only for operation after normal end-of-cycle.
Limiting events have been analyzed for cycle extension to the exposure attainable using FH005, ICF and FFWTR at full power.
8906210259 890614 PDR ADOCK 05000352 l
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t 4 l For the ELLLA extension the topical reports (Ref. 2 and 3) discuss a full range of transient and accident events relevant to the region extension, and presents results of calculations or previously approved conclusions. The transient analyses demonstrate that the licensing basis results (e.g.,100 percent flow, 100 percent power for pressurization transients) bound the ELLLA region results (e.g., 87 percent flow, 100 percent power). These conclusions apply to all relevant MCPR events such as pressurization, rod withdrawal and i
flow runout events. Changes to MCPR TS are not required because of ELLLA i
adoption. Other relevant areas such as over pressure protection, LOCA and i
containment analysis have also been examined, and the analyses indicate that j
results are within allowable design limits. Thermal-hydraulic stability will 1
be provided for by appropriate surveillance. The analyses have been done with l
approved methodologies and the results are similar to previously approved ELLLA extensions. Thus operation within the ELLLA region is acceptable for cycle 1 i
operation of Limerick Unit 2.
l Nuclear transient data LOCA analyses and thermal hydraulic stability analyses consistent with the analyses previously performed for Unit I were developed to include the combination of ELLLA with PFH and ICF. Lower initial operating piessure and steam flow rate (due to lower feedwater temperature) provide more overpressure margin for the limiting MSIV closure flux scram event. Hence, it I
is concluded that the pressure barrier integrity is maintained under PFH l
conditions. The licensee has analyzed the overpressurization limiting l
transient (MSIV closure) for increased core flow (ICF) without PFH. The 1
i analysis of this bounding transient predicted a peak vessel pressure of 1273 psig which is below the ASME code limit of 1375 psig and the analysis results are therefore acceptable.
The fuel loading error accident, rod drop accident, and rod withdrawal error have been evaluated by the licensee for ICF and/or PFH operation. The rod withdrawal error transient is limited by a rod block system. The addition of a "high flow clamped" trip setpoint limit of 106 percent and allowable value of 109 percent of rated flow for the rod block monitor upscale alarm in TS Table 3.3.6-2 ensures that the rod blocks currently in the TS cannot be exceeded.
The reactor coolant system recirculation flow upscale trip setpoints and allowable values are increased and the values for the recirculation pump MG set scoop tube mechanical and electrical stops are increased. These changes are necessary to accommodate the increased core flow operation and are acceptable. The licensee has stated that the fuel loading error and rod drop accident are not adversely affected by the proposed changes.
For the fuel loading error, the licensee has reported in Reference 6 a maximum increase in CPR of 0.04 from the value of 0.11 stated in the FSAR for this event at rated conditions. Thus the fuel loading error remains a non-limiting event. With regard to the rod drop accident, the LGS utilizes a banked position withdrawal sequence (3PWS) for control rod moYement. Based on prior staff review of BPWS as present3d by General Electric (Ref. 7, Section S.2.5.1.3), the staff agrees that this event is not adversely affected by the proposed changes.
. A loss of coolant accident (LOCA) with ICF and pFH was addressed by the licensee in Reference 3.
The LOCA analyses with ICF alone bound operation with ICF and pFH.
Since the peak clad temperature for ICF increases by less than 10' F for the limiting break compared to the rated core flow condition, the l
calculated peak clad temperature (PCT) of approximately 2100" F remains below the 10 CFR 50.46 cladding temperature limit. No changes to the current maximum average planar linear heat generation rates (MAPLHGR) are required.
In Reference 3, GE stated that PCT changes throughout the remainder of the large break spectrum will be of a similar magnitude (less than 10* F).
Consideration was given to the break spectrum range of 60 to 100 percent DBA for the separate effects of ICF for several classes of BWR plants with the resulting conclusion that increased core flow results in a peak clad temperature increase of less than 10 degrees F throughout the large break spectrum. The separate effect of reduced feedwater temperature is to reduce the calculated peak clad temperature.
A discussion was presented for both reduced feedwater temperature and increased core flow conditions which bounds the conditions described in the proposed amendment. Based on the staff's review of the information provided by the licensee, the staff agrees with the conclusion in NEDC-31578 that the effect of ICF will not alter the limiting break size. The calculated peak clad temperature remains below the 10 CFR l
50.46 c idding temperature limit and is acceptable.
The impact of the proposed operating mode on containment LOCA response was considered by the licensee. A conservative analysis resulted in a peak drywell deck downward differential pressure 2.6 psi higher than the value of 26.0 psid in the LGS FSAR. However, this is still below the design limit of 30.0 psid reported in the FSAR.
It was also stated that the peak suppression pool temperatures, chugging loads, condensation oscillations and pool swell bounding loads were all found to be bounded by the rated power analysis in FSAR Chapter 6.
We find this acceptable.
Reference 3 included a discussion of thermal-hydraulic stability (THS) for the LGS.
The proposed LGS Unit 2 technical specifications implement a generic set of operating recommendations (Ref. 8) to assure acceptable plant performance in the least stable portion of the power / flow map and to provide operator instructions for the detect-and-suppress mode of operation. The THS compliance for all licensed GE BWR core fuel is demonstrated on a generic basis by Reference 9 and has been approved by the staff (NRC Safety Evaluation Report Approving Amendment B to NEDE-24011-P contained in Appendix US-C to Reference 7).
PECo has also committed (Ref. 10) to implement GE recommendations for thermal-hydraulic stability actions as outlined in NRC Bulletin No. 88-07 l
supplement 1:
" Power Oscillations in Boiling Water Reactors (BWR), dated December 30, 1988. The staff concludes that acceptable THS provisions have been made.
We have reviewed the information provided by the Philadelphia Electric Company relative to the proposed operation of the Limerick Generating Station J' nit 2 in the ELLLA Region, combined with partial feedwatrer heating and increased core flow. Based on the results of the evaluation, the staff concludes that the proposed operations are acceptable.
Date: June 14, 1989
REFERENCES 1.
Letter and enclosures, G. A. Hunger, Jr., PEco, to WRC,
Subject:
" Draft Limerick Generating Station, Units 1 and 2 FSAR Revisions to Incorporate the Extended Load Line Region, Increased Core Flow, and Partial Feedwater Heating", dated May 31, 1989.
2.
NEDC-31577P " Extended Load Line Limit Analysis for Limerick Generating Station Unit 2, Cycle 1", dated March 1989.
3.
NEDC-31578P " Increased Core Flow and Partial Feedwater Heating Analysis for Limerick Generatiig Station Unit 2, Cycle 1", dated March 1989 with Errata and Addenda No. 1, dated May 31, 1989.
4 Letter, R. E. Martin, NRC, to E. G. Bauer, PEco, transmitting Amendment No. 3 to Facility Operating License No. NPF-39, dated February 17, 1987 5.
Letter, R. J. Clark, NRC, to E. G. Bauer, PECo, transmitting Amendment No.
7 to Facility Operating License No. NPF-39, dated August 14, 1987.
6.
Letter, M. J. Cooney, PECo, to W. R. Butler, NRC, dated January 2,1987.
7.
General Electric Standard Application for Reactor Fuel (Supplement for US), May 1986 (NEDE-24011-P-A-8-US, as amended).
8.
General Electric Service Information Letter No. 380, Revision 1 February 10, 1984 9.
G. A. Watford, " Compliance of the General Electric Boiling Water Reactor Fuel Designs to Stability Licensing Criteria," October 1984 (NEDE-22277-P-1).
- 10. Letter, J. W. Gallagher, PECo, to NRC dated March 7, 1989 - LGS 1/2, Response to NRC Bulletin No. 88-07, supplemental 1:
" Power Oscillations in Boiling Water Reactors (BWRs)" dated December 30, 1988.
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