ML18040A146
| ML18040A146 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 01/25/1995 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17059A640 | List: |
| References | |
| NUDOCS 9501310157 | |
| Download: ML18040A146 (10) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON' C 20miO001 T
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0 I. 0 2IIHKUUIR By letter dated July 21, 1994, Niagara Hohawk Power Corporation (the licensee or NHPC) submitted a request for changes to the Nine Hile Point Nuclear Station Unit No. 1'(NHP-1), Technical Specifications (TSs).
The requested changes would eliminate seven reactor head safety valves from TS 2.2.2, "Reactor Coolant System,"
TS 3.2.8/4.2.8, "Pressure Relief Systems Safety-Valves," and the associated Bases for these sections.
The current plant configuration includes 16 safety valves and 6 pressure relief valves.
The original design for NHP-1 overpressure protection was based on American Society of Hechanical Engineers Boiler and Pressure Vessel Code (ASHE Code),
Section I, 1962 Edition (Ref. 2).
ASHE Code,Section I, deals specifically with power boilers and not nuclea'r reactors.
(In 1962, ASHE Code,Section III, which deals with nuclear reactors did not exist.)
This edition was interpreted to require that the safety valve capacity be such that all potentially generated steam by the boiler (reactor vessel) be discharged without credit for fuel stoppage (reactor scram).
The current versions of the ASHE Code (Ref. 3), Sections I and III, allow credit for operating and/or safety controls in the boiler or nuclear reactor.
Based on the current ASHE Code and NUREG-0800, "Standard Review Plan" (Ref. 4),
NHPC proposes to take credit for the high flux scram and plans to remove 7 out of 16 safety valves provided at NHP-1.
NHPC proposed that the reduction of the number of safety valves would result in considerable reduction in man-rem exposure and savings due to reduced maintenance and surveillance testing.
The NHP-1 safety valves are grouped into 5 groups with each group of safety valves set to relieve at staggered groups of pressures as specified in TS 2.2.2.
The proposed'mendment would not change the specified relief set points for the 5 groups of valves; the proposed amendment would only decrease
'the number of valves in each group so that the total number of valves would be
.9 rather than 16.
9'501310157 9501250 PDR ADOCK 05000220 P
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TS 4.2.8 would be revised to require periodic testing of the safety valves in accordance with the NHP-1 Inservice Testing Program rather than having TS 4.2.8 specify that at least 8 of the 16 valves be tested at least once per operating cycle.
2.0 EIIALUATIGN The overpressure protection system prevents overpressurization of the reactor coolant pressure boundary under the most severe transients and limits the reactor pressure during normal operational transients.
Safety valves are required to be designed with sufficient capacity to limit the pressure to less than 110 percent of the reactor coolant pressure boundary design pressure of 1250 psig.
Currently, overpressure protection is provided by 16 safety and 6 relief valves located on the reactor head and on the two main steam lines between the reactor vessel and the first isolation valve inside the drywell.
The relief valves discharge to the suppression pool and the safety valves discharge to the drywall.
The combined relief and safety valve capacity is approximately 13.5x10 lb/hr and equivalent to 185 percent of the total steam flow.
The probability of lifting the safety valves is low due to the turbine'ypass capability of 40 percent and the 6 relief valves.
The proposed combined relief and safety galve capacity after removing the 7 safety valves would be approximately 9x10 lb/hr and equivalent to 124 percent of the total steam flow.
Since the only function of the safety valves at NHP-1 is to provide ASHE Code overpressure protection, NHPC performed an overpressure protection analysis to verify that 9 safety valves are sufficient to meet the acceptance criteria for overpressure protection.
The impact on Anticipated Transient Without Scram (ATWS) was also examined.
2.1 Ove ress re Protection H in Ste ine Isolation with Hi h Neutron Flux Scram The peak pressure with 9 safety valves was calculated using General Electric approved methodology documented in NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel (GESTAR)," February 1991.
The analysis was performed assuming that:
a) the reactor is in operation at the design rated power of 1850 HW, and design pressure of 1250 psig, b) main steam line isolation occurs without a scram, c) all pressure relieving devices fail, and d) the reactor scrambled on high neutron flux (120X of rated flux).
The high neutron flux signal is the second safety grade scram signal from the reactor protection system following main steam isolation valve (HSIV) closure.
Normally a reactor scram would occur on HSIV closure of 10 percent.
The analysis only took credit for the 9 safety valves.
The results of the analysis demonstrated that the calculated peak pressure of the vessel remained below the 1375 psig limit (110 percent of the vessel design pressure).
For the most severe transient, main steam line isolation with a high neutron flux scram at 120 percent, the calculated peak pressure is 1353 psig when only
9 safety valves are assumed to operate in the safety mode.
Since the calculated peak pressure, 1353 psig, is within the acceptance criterion of 1375 psig, the overpressur e protection analysis is acceptable.
2.2 t
ted t
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i Steam ine Isolation t
Hi et Scr m
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Tri Another plant upset which could be impacted by this change is the postulated ATMS event.
Mhile not part of the plant's original design basis, requirements were imposed by 10 CFR 50.62 to reduce the likelihood and consequences of an ATMS.
The Code of Federal Regulations at 10 CFR 50.62 required that all boiling water reactors have an alternate rod injection (ARI) system, a standby liquid control system (SLCS),
and automatic reactor recirculation pump trips.
An analysis was performed assuming that:
a) the reactor is in operation at the design rated power of 1850 HM, and design pressure of 1250 psig, b) main steam line isolation occurs withouk a scram, and c) credit taken for recirculation pump trip and six relief valve actuations.
Normally, the ARI logic at NHP-1 would cause the alternate rod injection valves to energize wheh either the reactor vessel high pressure trip setpoint or the low water level trip setpoint is reached.
Once energized, insertion of the control rods begins within 15 seconds.
For this transient, the peak calculated pressure was 1322 psig when only 9 safety valves are assumed instead of 16 safety valves and the ARI fails.
This result is consistent with the GE ATMS Event Evaluations (NEDE-24222).
Therefore, based on the analyses presented
- above, the staff concludes that operation with 9 safety valves instead 16 safety valves will not endanger the public health and safety.
2.3 R vi ion to S
et V lve S rveill ce e
irement The proposed change to TS 4.2.8 would require the safety valves to be tested in accordance with the NHP-1 Inservice Testing (IST) Program.
The NHP-1 IST Program is based on the ASHE Code,Section XI, 1983, including Summer Addenda which is in accordance with the requirements of 10 CFR 50.5a(f) and is, therefore, acceptable.
- 3. 0
~SUMMA NHPC performed an overpressure analysis consistent with Standard Review Plan 5.2.2 to support its proposal to remove 7 safety valves at NHP-1.
The setpoints, as specified in TS 2.2.2, for the 5 groups of safety valves were not changed in this analysis.
The results of the overpressure analysis were consistent with the NRC staff's acceptance criteria that pressure not exceed 110 percent of design.
The licensee also discussed the impact on ATMS
- response, indicating no significant impact.
Periodic surveillance testing of the safety valves will be performed 'in accordance with the NHP-1 IST Program.
Therefore, NHPC's proposal to remove 7 safety valves from the current
16 safety valves is acceptable.
The changes proposed in TS 2.2.2 and 3.2.8/4.2.8 reducing the number of safety valves and the bases are also acceptable.
4.0 STA CONSULT 0
In accordance with the Commission's regulations, the New York State official was notified of the proposed issuance of the amendment.
The State official had no comments.
5.0 V RO H
L C TIO The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes a surveillance requirement.
The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released
- offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure.
The Commission has previously issued a
proposed finding that the amendment involves no significant hazards consider ation, and there has been no public comment on such finding (59 FR 45027).
Accordingly, the amendment meets the eligibilitycriteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
6.0
~CII C
U I N
The Commission has concluded, based on the considerations discussed
- above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed
- manner, (2) such activities will be conducted in compliance with the Commission s regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
7.0 R
1.
Letter from B.
R. Sylvia (NHPC) to U.S.
NRC, "Nine Hile Point Unit 1,"
dated July 21, 1994.
2.
ASHE Boiler and Pressure Vessel
- Code,Section I, "Safety Valves," American Society of Hechanical Engineers,
- 1962, pgs.
75 80.
Cj
3.
ASHE Boiler and Pressure Vessel
- Code,Section III, Article NB-7000, "Protection Against Overpressure,"
American Society of Hechanical Engineers.
4.
NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants,"
June 1987.
Principal Contributors:
K. Kavanagh D. Brinkman Dated:
January 25, 1995
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