ML17054C054
| ML17054C054 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 04/14/1987 |
| From: | Capra R Office of Nuclear Reactor Regulation |
| To: | Mangan C NIAGARA MOHAWK POWER CORP. |
| References | |
| NUDOCS 8704160308 | |
| Download: ML17054C054 (16) | |
Text
DocketfNo. 50-410 Mr. C.
V. Mangan, Senior Vice President Niagara Mohawk Power Corporation 301 Plainfield Road
- Syracuse, New York 13212
'ri4 14, I
DISTRIBUTION NRCPDR Local PDR PDI-1 Rdg.
R.
Capra M. Haughey OGC
Dear Hr. Hangan:
SUBJECT:
DRAFT SAFETY EVALUATION REPORT FOR NINE MILE POINT 2 HAIN STEAM ISOLATION VALVE LEAKAGE CONTROL SYSTEM I9SZ J. Kudrick J. Craig F. Witt ACRS(10)
D. Neighbors C.
Vogan S. Varga Enclosed is a draft Safety Evaluation Report concerning the evaluation of the deletion of the main steam isolation valve (MSIV) leakage control system (LCS) when using wye-pattern globe valves at Nine Mile Point Unit 2 (NHP-2).
The enclosed draft is being sent to you in response to your request and does not represent a final NRC decision.
A final NRC decision is expected to be included in our response to your request for an amendment to the NMP-2 license and technical specifications dated March 11, 1987.
Two Notices of Consideration of Issuance of Amendment have been issued concerning your March 11, 1987 request.
The associated amendments to the license and the Technical Specifications, as well as the final determination with respect to your need for an MSIV LCS, cannot be issued until the 30 day comment period for these notices has expired.
Furthermore, if the deletion of the MSIV LCS is determined to be acceptable, subsequent changes concerning the MSIVs or containment bypass
- leakage, such as changing the MSIV allowable leak rate in the Technical Specifications, or continued excessive
If you have 'any questions concerning the enclosed draft Safety Evaluation Report please contact the Licensing Project, Manager, Mary Haughey (301) 492-7136.
Sincerely,
Enclosure:
As stated cc:
See next page Robert A. Capra, Director Project Directorate I-1 Division of Reactor Projects, I/II PDI-1 MHaughey 4/87
~eC PDI-1 RCapra 4/A /87 870'41IBI0308 87041f PDR 'DOCK 05000410 P
- . 0
Mr. C. Y. Mangan Niagara Mohawk Power Corporation Nine Mile Point Nuclear Station Unit 2 CC:
Mr. Troy B. Conner, Jr.,
Esq.
Conner 5 Wetterhahn Suite 1050 1747 Pennsylvania
- Avenue, N.W.
Washington, D.C.
20006 Richard Goldsmith Syracuse University College of Law E. I. White Hall Campus
- Syracuse, New York 12223 Ezra I. Bialik Assistant Attorney General Environmental Protection Bureau New York State Department of Law 2 World Trade Center New York, New York 10047 Resident Inspector, Nine Mile Point Nuclear Power Station P. 0.
Box 99
- Lycoming, New York 13093 Mr. John W. Keib, Esq.
Niagara Mohawk Power Corporation 300 Erie Boulevard West
- Syracuse, New York 13202 Mr. James Linville U. S..Nuclear Regulatory Commission Region I 631 Park Avenue King of Prussia, Pennsylvania 19406 Mr. Peter E. Francisco, Licensing Niagara Mohawk Power Corporation 301 Plainfield Road
- Syracuse, New York 13212 Don Hill Niagara Mohawk Power Corporation Suite 550 4520 East West Highway
- Bethesda, Maryland 20814 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, Pennsylvania 19406 Mr. Paul D. Eddy New York State Public Serice Commission Nine Mile Point Nuclear Station-Unit II P.O.
Box 63
- Lycoming, New York 13093 Mr. Richard M. Kessel.
Chair and Executive Director State Consumer Protection Board 99 Washington Avenue
- Albany, New York 12210 Jay Dunkleberger Division of Policy Analysis and Planning New York State Energy Office Agency Building 2, Empire State Plaza
- Albany, New York 12223
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%~*g4 OO UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION REPORT BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO WYE PATTERN GLOBE VALVES (MSIVs WITHOUT A LEAKAGE CONTROL SYSTEM NIAGARA MIOHAWK POWER CORPORATION NINE MILE POINT, UNIT 2 DOCKET NO. 50-410
- 1. 0 INTRODUCTION The Main Steam Line Isolation Valves (MSIVs) of a Boilina Water Reactor (BWR) are designed to isolate the Reactor Pressure Vessel (RPV) in the event of a design basis steam line break downstream of the MSIVs, a design basis Loss of Coolant Accident (LOCA), or any other event that would warrant containment isolation.
This is required by 10 CFR Part 50 Appendix A, General Design Criteria (GDC) 54 and 55.
The closure of the MSIVs should terminate releases of radioactivity from the RPV for accidents within the design bases, and ensure that offsite and onsite dose guidelines of 10 CFR Part 100, and 10 CFR Part 50, Appendix A, GDC 19, respectively, are not exceeded.
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In April 1979, the licensee selected 24 inch positive seal ball-type MSIVs to replace the wye pattern globe valves and the leakage control system as described in the Nine Mile Point Unit 2 Preliminary Safety Analysis Report.
Because of the unique design of the positive valve seal, the leakage from the ball valves was expected to be very low.
Therefore, a leakage control system was not considered necessary for the'all-type MSIVs.
This conclusion was documented in a letter to G.
K. Rhode of Niagara Mohawk Power Corp.
from R. L. Tedesco, dated January 2, 1981.
Experience with the ball-type'SIVs during preoperational testing at Nine Mile Point Unit 2 and laboratory prototype testing have failed to demonstrate that these valves will function as anticipated.
Delamination of the tungsten carbide coating on the ball was observed, which is believed to have been caused by wearing of the stellite seat.
This resulted in excessive seat leakage.
- Also, during the initial valve test at system temperatures in an offsite prototype facility, packing leakage developed.
In view of these engineering problems and a schedular
- concern, the licensee informed NRC in a letter dated March 11, 1987 (NMP2L 1004), that the Nine Mile Point Unit 2 ball type MISIVs will be replaced with wye pattern globe valves, manufactured by Rockwell, that are similar to those being used in other BWRs.
Shop acceptance test results indicate that the wye pattern globe valves leak between 2 and 4 scfh, which meet the Technical Specification limit of 6 scfh (limit for ball type MSIVs).
- Further, the new valves will close in 3 to 5 seconds, also in accordance with the Technical Specifications.
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By letter dated March 18, 1987 (NMP2L 1007), the licensee requested that a
Leakage Control System (LCS) not be required.
Regulatory Guide I.96 "Design of Main Steam Isolation Valve Leakage Control Systems for Boiling Water Reactor Nuclear Power Plants", describes a basis for implementing General Design Criteria (GDC) 54 with regard to a leakage control system (LCS) for the MSIVs to ensure that the radiological consequences of design basis accidents do not exceed the dose guidelines of 10 CFR Part 100.
The licensee proposed an alternative to a LCS using NUREG-1169 as guidance, pending final resolution of Generic Issue C-8, MSIY Leakage and LCS Failures.
A realistic fission product transport model developed by the BWROG in NUREG-1169 was used by the'licensee to assess the offsite and onsite dose consequences of alternate means of managing post-accident MSIV leakage using both safety-grade and non-safety-grade systems that could be available for service after a Loss of Coolant Accident (LOCA).. The licensee s radiological analysis takes credit for the isolated condenser (main steam line condensate drains open to the condenser) as a MSIV post-accident leakage management method.
The analysis demonstrates that the 10 CFR Part 100 dose guidelines will not be exceeded at leakage rates substantially in excess of the Technical Specification limit of 6 scfh.
The analysis further indicates that a total MSIV leak rate of 150 SCFH for all main steam lines (38 scfh/steam line) would not result in control room personnel doses in excess of 10 CFR Part 50, Appendix A, GDC 19.
By letter dated March 31, 1987 (NMP2L 1014), the licensee has provided the following additional information requested by the staff:
l.
A comparison of the Nine Mile Point 2 Rockwell MSIVs to Rockwell valves used at other nuclear power plants for the intended service; 2.
A compilation of industry leak rate testing results and experience; 3.
An evaluation of items I and 2 above, and a comparison to the NUREG-1169 analysis performed for Unit 2, including an estimate of leakage performance over the first operational cycle; and 4.
A discussion'of the maintenance practices planned at Unit 2 to enhance low leakage characteristics.
2.
EVALUATION 2.1 Leaka e Control S stem Beginning about 1970, the staff's concern over the possible dose consequences of MSIV 'leakage at or above the Technical Specification leakage limit led to the requirement that a Leakage Control System be installed in new plants.
Until a couple of years ago a majority of the "as found" MSTV leakage values were often in excess of Technical Specification limits.
In some cases, MSIY leakage rates were greatly in excess of the Technical Specification value, such that a
LCS would have been ineffective because of flow limitations in its design.
As a result of these
- concerns, the staff prioritized the MSIV leakage and LCS failures as a high priority Generic Issue (C-8).
Independently, the BhlR Owners Group (BMROG) formed the HSIY Leakage Control Committee to determine the cause of the high leakage rates associated with many of the MSIVs and to develop recommendations to reduce the leakage rates.
The'licensee has concluded that a
MSJV Leakage Control System is not necessary since Nine Mile Point 2 has a means of collecting, treating, and dischar ging from the stack MSIV leakage using existing systems.
This can be accomplished bv:
1.
A passive steamline drain system which automatically opens on loss of air power and first stage turbine pressure to the main condenser; 2.
Electric boilers capable of providing steam to the steam jet air
- ejectors, offgas system, and turbine gland seal and exhaust system; and 3.
In the event of a LOCA and/or loss of offsite power, NMP2 has the
- capability to re-establish condenser
- vacuum, the operation of the steam jet air ejector, the operation of the gland seal and exhaust
- system, and the offgas steam once offsite power is restored.
2.2 MSIV Leaka e
Ex erience To assess the expected HSIV leakage characteristics for Nine Mile Point Unit 2, leakage tests at operating plants using the Rockwell wye pattern globe valves that are similar to the MSIVs being installed at Nine Mile Point Unit 2 w'ere reviewed.
A total of 39K of the "as found" leakage test results were 6 scfh or below, which is the Technical Specification limit for Nine Mile Point Unit 2.
Cumu'latively, 85% of all test results were less than 38 scfh.
In the future the leakage rate percentage below 6 scfh as well as below 38 scfh could conceivably be higher, mainly due to the adoption of the recent BMR Owners Group recommendations.
These recommendations include improvements in test
- methods, maintenance procedures, training and tooling.
In addition, all other BWPs have MSIV Technical Specification leak rates of 11.5 scfh or above.
The Technical specification limit by itself does not ensure that the refurbished valves will not leak above 6 scfh.
The higher Technical Specification leak rates greater than 6 scfh, however, do bias the percentage of leakage rate results below 6 scfh on the low side.
It can be concluded that a sound maintenance program should limit valve leakage degradation and increase leakage test results within Technical Specification values.
2.3 MSIV Desi n Chan es Based on experience with Rockwell designed MSIVs for BMR service, the licensee made MSIV design changes using information provided by other Bh'R operating
- plants, valve suppliers, General Electric Co., and an evaluation of Inspection and Enforcement Bulletins, Notices and Circulars applicable to Nine Mile Point Unit 2 MSIVs.
These design changes include:
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Disc-piston connection confiouration changed from a spherical backseat to resolve disc-to-piston separation ouestions; 2.
Numatic air valves replaced with Norgren air valves to resolve sticking air valve spools; 3.
Improved stem/stem-disc and main disc/piston connection (joints) to resolve stem/stem-disc and main disc to piston separation potential; 4.
Spring flange bronze bushing used to reduce the tendancy for galling/
friction between yoke guides and tubes; 5.
New spring divider material used to reduce the tendancy for galling/
scoring of the yoke guide tubes; and 6.
Modified packing chamber design with graphite rings were used to replace asbestos packing to enhance packing and stem leak. lightness capability.
2.4 Radiolo ical Assessment In the event that leakage values are in excess of the Technical Specification limit, 10 CFR Part 100 offsite and 10 CFR Part 50, Appendix A, GDC 19 control room operator dose guidelines would not necessarily be exceeded.
The 39 scfh leak rate, that 85K of the leakage tests met, is important from the standpoint that the calculated doses have been found to be within the controlling design basis accident dose guideline valves of 10 CFR Part 50, Appendix A, GDC 19.
Based on NUREG-1169 methodology, and using realistic assumptions of the holdup volume and surfaces of the main condenser and main steamlines and fission product atten-uation elsewhere,'ffsite and control room doses were evaluated.
The licensee's analysis indicated that IO CFR Part 100 offsite doses would be met.
The licensee indicated that the control room was limiting and that a combined MSIY leak rate of 150 scfh for all main steam lines (38 scfh per main steam line) would not result in control room personnel doses in excess of 10 CFR Part 50., Appendix A, GDC '9.
On the basis of our review, we conclude that the licensee's radiological evaluation, which takes credit for the isolated condenser, is reasonable.
This analysis is a
departure from the Standard Review Plan and Regulatory Guides in that some realistic assumptions were utilized for assessing control room habitability (GDC 19) if a design basis LOCA we> e to occur and the MSIVs leaked at rates in excess of their Technical Specification limit of 6 scfh.
For such an accident during which the MSIYs leaked at rates of 6 scfh, or less, the staff and licensee have both determined that the dose guidelines of GDC 19 would be met.
These analyses followed the guidance of the Standard Review Plan with two exceptions.
The first exception was the modeling of atmospheric dispersion.
The second was credit for post-accident fission product attenuation in the steamlines.
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.g0 Ji The staff also has reasonable assurance that the~88 scfh leak rate per main steam line represents an upper bound when one considpPs the expected improved leakage values for the Mine Nile Point Unit 2 ya'1ves
~ provided that effective and careful NSIV maintenance is followed.
The licensee~expects deterioration in the NSIV leakage to result
$ n leakzne rates less than 16 s h at the end of the first operating cycle.
The overall risks from accident sequen es in which hlSIV leakage is a significant factor are low without a LCS, and pos accident management schemes (including those stated above) were, shown to pr duce significant offsite dose reductions in lieu of a LCS.
NSIV leakage wa concluded to be a trivial safety concern, and NSIV leakage control was show to not be risk significant (most of the risk being from accidents resulting i core melt and containment failure).
- However, for accidents that do not resul in containment failure, NSIV leakage can still be important.
Several leakag treatment methods which make use of the holdup volume and surface of main s am lines and condensers, and fission product atten-uation elsewhere, were eval ted and indicated lower offsite dose consequences than with a LCS.
Mire Nile Poi Unit 2 design features are similar to the NUREG--269
- base plant and, therefore the conclusions of NUREG-2169 are considered applicable to Nine Nile Point Unit NUREG-2269 concluded that6the low public exposure (isolated condenser and 1.5 scfh leak rate - 5.9x10 4man rem/plant year whole body public exposure; LCS and 11.5 scfh leik rate - 2.0x10 man rem/plant'ear whole body public exposure) does not justify a LCS.
2.5 Technial S ecification Chan es QJQ W The replacement~
pattern globe NSIVs are air-operated (AOV) valves and the ball NSIVs cfire hydraulica'lly operated.
This necessitates valve nomenclature chanaes in lechnical Specification Table 3.6.1.2-1 and 3.6.3-1 from 2NSS"HYV6A, B, C, D and 2NSS+)IYV7A, B, C, D to ZNSS+AOV6A, B, C, D and 2NSS"AOV7A, B, C, D.
2.6 NSIV Naintenance and Procedures The staff is reasonably assured that the wye pattern globe valves beina installed in Nine Nile Point Unit 2 without a leakage control system, but with a post accident leakage treatment method, can perform their function without exceeding the dose guideline valves of 20 CFR Part 100 and GDC 19 of 10 CFR Part 50.
This assurance is dependant on proper maintenarce practices, and
'potential operator actions/emergency operating procedures to limit NSIV radioactivity releases.
By letter dated April 7, 1987, the licensee has committed to implement the following prior to criticality.
2.
vendor recommended maintenance boring, grinding, and lapping tools will be available for refurbishment as needed to restore NSIVs to less than 6 scfh leakage, 2.
maintenance procedures based upon NSIV instruction manuals, vendor and General Electric recommendations (including careful maintenance, supervision and inspections to indicate incipient failures);
3.
training proarams for MSIV maintenance personnel; 4.
operating procedures for the post-accident control and treatment of MSIV leakage to limit radioactivity releases as recommended by BWROG in NEDO-30324; and 5.
emergency operating procedures to limit radioactivity release through the MSIVs as reconmended by the BWROG in NED0-30324.
3.0 CONCLUSION
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On the basis of the evaluation above; the staff concludes that Nine Mile Point Unit 2 may resume plant operation without a leakage control system, but with post-accident leakage management.
This conclusion is based on a sound NSIV maintenance program committed to by the licensee which includes:
maintenance procedures, toolino and equipment, personnel training, operating and emergency procedures, management and inspection.
At the Technical Specification leak rate of 6 scfh, the dose guideline valves
- of 10 CFR Part 100 and GDC 19, calculated for a design basis LOCA, will not be exceeded.
In the event that the Technical Specification limit is exceeded, 10 CFR Part 100 "offsite and 10 CFR Part.50 GDC 19 control room operator dose guidelines calculated using the methodology from NUREG-1169 would not be exceeded.
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