ML20198G507
| ML20198G507 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 11/22/1985 |
| From: | Bernero R, Bernero R Office of Nuclear Reactor Regulation |
| To: | Thompson H Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20151L176 | List: |
| References | |
| FOIA-86-266 NUDOCS 8512090522 | |
| Download: ML20198G507 (15) | |
Text
,
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ie 50V 2 21985 MEMORANDUM FOR:
Hugh L Thompson, Jr., Director, Division of Licensing FROM:
Robert M. Bernero, Director, Division of Systems Integration
SUBJECT:
SAFETY IMPLICATIONS OF PWR MAIN STEAM SAFETY VALVES FLOW DEFICIENCY Purpose and Background Per the Operating Reactors Events Meeting 85-20, held on October 21, 1985, the Division of Systems Integration was assigned to evaluate the safety implications of the Main Steam Safety Valves' (MSSVs') flow deficiency.
The purpose of this memorandum is to address this issue.
During the last part of 1984 through aid 1985, Wyle Laboratories conducted several full flow tests on two MSSVs manufactured by Crosby Valve and Gage Company.
These two valves were chosen as a representative sample for the MSSVs that are to be installed on the Seabrook plant main steam lines. The purpose of the Wyle tests was to determine the adequacy of different discharge piping configurations. Test measurements indicated that the MSSVs (with the manufac-turer's recommended guide ring settings), have a flow capacity of about 50% of their design values at the design pressure.
The guide ring settings determine the force being exerted on the valve disc, thereby affecting the degree of valve lift and subsequently the discharge flow capacity of the valve.
During the Wyle tests the guide ring settings of both valves were substantially adjusted in order to achieve the full flow capacities.
Subsequent to the completion of the tests, the Public Service Company of New Hampshire, the owner of the Seabrook plant, concluded that in order to ensure full flow capacity of the plant's MSSVs they all should be adjusted downward by 130 notches from the manufacturer's settings.
While the Seabrook experience shows a deficiency in the capacity adjustment of the Crosby spring loaded safety valves, it strongly suggests a similar defi-ciency in similar valves made by other manufacturers, since they all work on the same basic concept.
Safety Implications Full flow testing of MSSVs is not normally performed by either reactor owners or valve manufacturers, nor is such testing an ASME requirement for capacity certification.
Such certification is obtained through extrapolation from tests on much smaller valves at low pressures.
M Based on the Seabrook experience and with the lack of sufficient data, it may be assumed that a number of deficient safety valves are installed in some operating plants and/or planned to be installed in plants yet to operate.
CONTACT:
S. Diab, RS8, x29440 pp}&M 4 m.
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m NOV 2 21985 1
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H-Thompson, Jr..
1 l
It may also be assumed that the capacity of some of these deficient valves may be as low as 50% (as for the Seabrook plant) of the design flow, or even lower.
With this potential deficiency, the design basis of the affected plants cannot be met.
The design basis of every pressurized water reactor (PWR) requires that overpressure protection for the primary and secondary sides of the plant be provided so that the pressure does not rise above 110% of the design value during anticipated operational occurrences and postulated accidents.
PWR vendors perform safety valve sizing analyses such that these valves have 3
sufficient capacity to mitigate the most severe overpressurization event with adequate margin.
Generic transient analyses, as opposed to sizing analyses, performed by Westinghouse for the Westinghouse PWRs show that, for the worst overpressurization event (loss of load without condenser bypass) the MSSVs' j
peak relieving capacity required is about 80% of the nominal valve flow.
It should also be noted that any flow degradation through the MSSVs increases the l
potential for the actuation of the primary safety valve actuation and increases the relieving load on these valves as well.
This is because any relief through i
the MSSVs will remove heat from the reactor coolant, which would otherwise j
I accumulate, expand, and overpressurize the primary system.
Genera". Design I
Criterion 15 of 10 CFR 50, Appendix A, requires that the reactor coolant system and associated auxiliary systems be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during normal operation and anticipated operational occurrences.
There ia insufficient test data available to the staff to show whether the valve flow would increase at high upstream pressure conditions and/or if the valve disc travel will be any higher.
Therefore, it is not clear whether overpressurizations of this nature are self limiting.
- )
With less than the required relieving capacity, the secondary side is expected to be overpressurized during postulated events, thus increasing the potential
}
for steam side leaks or breaks.
While the overpressurization described above would occur following a loss of load event, which is an anticipated operational i
occurrence, the consequences of that event may lead to a design basis event with its associated severe consequences. While plants are designed to acceptably accommodate their design basis events, the probability of occurrence of those events is sufficiently low.
However, if a design basis event were to occur with a substantially higher likelihood or the plant were to be pressurized over its design pressure limits with higher frequency, then the plant cannot meet its design basis requirements.
i Similar problems of inadequate ring settings of primary safety valves were discovered during the EPRI test program conducted in response to NUREG-0737, item II.D.1.
The problem was identified and confined to valves manufactured by Dresser Industries, Inc.
Users.of primary side Dresser valves made sub-i
}
mittals to justify continued operation of their facilities until the valve ring settings on their plants are readjusted consistent with the EPRI test i
findings.
The staff evaluated and approved those submittals.
Verification of i
the adequacy of all PWR primary safety valve ring settings is being pursued by the staff under multi plant action MPA F-14.
Iu i-
H. Thompson, Jr.
-3 NOV 2 2 G85 Based on the limited information available about the adequacy of the secondary side safety valves for overpressurization mitigation, the staff has a reason to doubt that all the safety valves currently in use or planned for use would serve their intended safety function if called upon.
Therefore, we suggest that the Division of Licensing send a request for additional information (RAI),
per 10 CFR 50.54(f), to PWR plant owners.
This RAI would request plant owners, in light of the Seabrook experience, to study the Seabrook experience and justify to the staff that their respective plants contiinue to have sufficient overpressure protection and are within their safety analyses.
The owners' justification may rely on any combination of: (a) relevant experience from which valve performance can be verified; (b) safety analyses assuming inadequate overpressure protection; or (c) representative valve testing.
v.13i=131? ole 73 yg,:st1. 0n:~2.;
Robert W. Bernero, Director Division of Systems Integration
Enclosure:
Suggested 10 CFR 50.54(f) Letter cc:
Edward Jordan, IE R. Baer, IE Mark Caruso, DL DISTRIBUTION:
Docket File DSI:AD:RS rdg.
DSI:D RSB rdg.
SSV Safety LMarsh SDiab rdg.
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Diab Rush, js
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DATE'$11/11/85 11/Ji/85 11/2 //85
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OFFICIAL RECORD COPY
ENCLOSURE _
i PROPOSED 10 CFR 50.54(f) LETTER Safety Implications of Main Steam Safety Valve Flow Deficiency During the last part of 1984 through-mid 1985, Wyle Laboratories conducted i
i several full flow tests on two MSSVs manufactured by Crosby Valve and Gage j
Company. These two valves were chosen as a representative sample for the MSSVs that are to be installed on the Seabrook plant main steam lines.
The i
purpose of the Wyle tests was to determine the adequacy of different discharge piping configurations.
Test measurements indicated that the MSSVs (with the manufacturer's recommended guide ring settings) have a flow capacity of about 50% of their design values at the design pressure.
The guide ring settings determine the force being exerted on the valve disc, thereby affecting the degree of valve lift and subsequently the discharge flow capacity of the valve.
During the Wyle tests the guide ring settings of both valves were substantially adjusted in order to achieve the full flow capacities.
Subsequent to the completion of the tests, the Public Service Company of New Hampshire, the owner of the Seabrook plant, concluded that in order to ensure full flow capacity of the plant's MSSVs they all should be adjusted downward by 130 notches from the manufacturer's settings.
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While the Seabrook' experience shows a deficiency in the capacity adjustment of the Crosby spring loaded safety valves, it strongly suggests a similar deficiency in similar valves made by other manufacturers, since they all work on the same basic concept.
Based on the Seabrook experience and with the lack of sufficient data, it may be assumed that a number of deficient safety valves are installed in some operating plants and/or planned to be installed in plants yet to operate. It may also be assumed that the capacity of some of these deficient valves may be as low as 50% of the. design flow (as for the Seabrook plant), or even lower.
With this potential deficiency, the design basis of the affected plants cannot be met.
The design basis of every pressurized water reactor (PWR) requires that ove n ressure protection for the primary and secondary sides of the plant be provided so that the pressure does not rise above 110% of the design value during postulated events.
PWR vendors perform safety valve sizing analyses such that these valves have sufficient capacity to mitigate the most severe overpressurization event with sufficient margin.
Generic transient analyses, as opposed to sizing analyses, performed oy, Westinghouse for the Westinghouse PWRs show that, for the worst overpressurization event (loss of load without condenser bypass) the MSSVs' peak relieving capacity required is about 80% of the nominal valve flow.
It should also be noted that any flow degradation through the MSSVs increases the potential for the actuation of the primary.
safety valves and increases the relieving load on these valves as well.
This is because any relief through the MSSVs will remove heat from the reactor coolant, which would otherwise accumulate, expand, and overpressurize the www.
h l
'. primary system.
General Design Criterion 15 of 10 CFR 50, Appendix A, requires that the reactor coolant system and associated auxiliary systems be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during normal operation and anticipated operational occurrences.
There is insufficient test data available to the staff to show whether the valve flow would increase at high upstream pressure conditions and/or if the valve disc travel will be any higher. Therefore, it is not clear whether overpressurizations of this nature are self limiting.
With less than the required relieving capacity, the secondary side is expected to be overpressurized, thus increasing the potential for steam side leaks or breaks. While the overpressurization described above would occur'fo11owing a loss of load event, which is an anticipated operational occurrence,'the con-sequences of that event may lead.to a design basis event with its associated severe consequences.
While plants are designed to acceptably accommodate their design basis events, the probability of occurrence of those events is suf-ficiently low.
However, if a design basis event were to occur with a sub-stantially higher likelihood or the plant.were to be pressurized over its design pressure limits with higher frequency, then the plant cannot meet its design basis requirements.
'Similar problems of inadequate ring settings of primary safety valves were discovered during the EPRI test program conducted in response to NUREG-0737,
.~
4_
item II.D.1.
The problem was identified and confined to valves manufactured by Dresser Industries, Inc.
Users of primary side Dresser valves made sub-mittals to justify continued operation of their facilities until the valve ring settings on their plants are readjusted consistent with the EPRI test findings.
The staff evaluated and approved those submittals. Verification of the adequacy of all PWR primary safety valve ring settings is being pursued by the staff under multi plant action MPA F-14.
Based on the limited information available about the adequacy of the secondary side safety valves for overpressurization mitigation, the staff has a reason to doubt that all the safety valves currently in use or planned for use will serve their intended safety function if called upon.
Therefore, per 10 CFR 50.54(f),
the staff requests that you, as a PWR owner, and in light of the Seabrook experience, justify that your plant continues to have sufficient overpressure protection, your facility continues to be in compliance with GDC 15, and the plant is within its safety analyses. Your justification may rely on any combination of: (a) relevant experience frorii which valve performance can be verified; (b) safety analyses assuming inadequate overpressure protection; or (c) representative valve testing.
1 Hugh L. Thompson, Director Division of Licensing
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ITEM NUMBER
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This item was placed on the February 84 agenda for s
tion.III and was withdrawn fol wing adverse coc:me s.
Frank Cherny came up ich new language that may be ac aptable to Sec III ubcommittee. This item will be tran -icted to Se ion III for action. Action is limited to h -7000.
ection III approved this item without negative v Floyd Mos.chini phone the Chairman and raise o ections to the definitions.
Frank Cherny vi check th him..
NP-1-85 Review of.' -7000 Paragraphs r Clearity and Uniformi The GPR continued with the review of the NX-7000 p ragraphs to clearify intent and maint n uniformity to the extent possible, for all classes.
he view will be continued ar,the next meeting.
NP-1-86 Production Testing of Class 2 Main Steam Safetv x
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Frank Cherny will report on thic item, flq '..,/,t/ f.[ 3 d q
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The next mee g will be he on March 12 & 13, 1986 at the United Engine ing enter, 6th Floor Board Room at 10:00 a.m.
This v e a two day meeting in order to complete item 85.
re will be no Agenda and this will the only notice the meeting.
Frank W. Catudal-Chairman, SGPR O
\\
NOV 2 21985 1
MEMORANDUM FOR:
Hugh L. Thompson, Jr., Director, Division of Licensing FROM:
Robert M. Bernero, Director, Division of Systems Integration
SUBJECT:
SAFETY IMPLICATIONS OF PWR MAIN STEAM SAFETY VALVES FLOW DEFICIENCY Purpose and Background Per the Operating Reactors Events Meeting 85-20, held on October 21, 1985, the Division of Systems Integration was assigned to evaluate the safety implications of the Main Steam Safety Valves' (MSSVs') flow deficiency. The purpose of this memorandum is to address this issue.
During the last part of 1984 through mid 1985, Wyle Laboratories conducted several full flow tests on two MSSVs manufactured by Crosby Valve and Gage Company. These two valves were chosen as a representative sample for the MSSVs that are to be installed on the Seabrook plant main steam lines. The purpose of the Wyle tests was to determine the adequacy of different discharge piping configurations. Test measurements indicated that the MSSVs (with the manufac-turer's recommended guide ring settings), have a flow capacity of about 50% of their design values at the design pressure. 'The guide ring settings determine the force being exerted on the valve disc, thereby affecting the degree of valve lift and subsequently the discharge flow capacity of the valve.
During the Wyle tests the guide ring settings of both valves were substantially adjusted in order to achieve the full flow capacities.
Subsequent to the completion of the tests, the Public Service Company of New Hampshire, the owner of the Seabrook plant, concluded that in order to ensure full flow capacity of the plant's MSSVs they all should be adjusted downward by 130 notches from the manufacturer's settings.
While the Seabrook experience shows a deficiency in the capacity adjustment of the Crosby spring loaded safety valves, it strongly suggests a similar defi-ciency in similar valves made by other manufacturers, since they all work on the same basic concept.
Safety Implications Full flow testing of MSSVs is not normally performed by either reactor owners or valve manufacturers, nor is such testing an ASME requirement for capacity certification. Such certification is obtained through extrapolation from tests on much smaller valves at low pressures.
Based on the Seabrook experience and with the lack of sufficient data, it ray be assumed that a number of deficient safety valves are installej in some operating plants and/or planned to be installed i plants yet t9 operate.
CONTACT:
S. Diab, RSB, x29440
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1 NOV 2 21995 H. Thompson, Jr..
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It may also be assumed that the capacity of some of these deficient valves may be as low as 50% (as for the Seabrook plant) of the design flow, or even lower.
^
With this potential deficiency, the design basis of the affected plants cannot be met.
The design basis of every pressurized water reactor (PWR) requires that overpressure protection for the primary and secondary sides of the plant be provided so that the pressure does not rise above 110% of the design value during anticipated operational occurrences and postulated accidents.
PWR vendors perform safety valve sizing analyses such that these valves have sufficient capacity to mitigate the most severe overpressurization event with adequate margin.
Generic transient analyses, as opposed to sizing analyses, performed by Westinghouse for the Westinghouse PWRs show that, for the worst overpressurization event (loss of load without condenser bypass) the MSSVs'
)
peak relieving capacity required is about 80% of the nominal valve flow.
i It should also be noted that any flow degradation through the MSSVs increases the i
i potential for the actuation of the primary safety valve actuation and increases the relieving load on these valves as well. This is because any relief through the MSSVs will remove heat from the reac' tor coolant, which would otherwise accumulate, expand, and overpressurize the primary system.
General Design Criterion 15 of 10 CFR 50, Appendix A, requires that the reactor coolant system and associated auxiliary systems be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during normal operation and anticipated operational occurrences.
There is insufficient test data available to the staff to show whether the valve flow would increase at high upstream pressure conditions and/or if the valve disc travel will be any higher. Therefore, it is not clear whether overpressurizations of this nature are self limiting.
With less than the required relieving capacity, the secondary side is expected to be overpressurized during postulated events, thus increasing the potential for steam side leaks or breaks. While the overpressurization described above would occur following a loss of load event, which is an anticipated operational occurrence, the consequences of that event may lead to a design basis event with its associated severe consequences.
While plants are designed to acceptably accommodate their design basis events, the probability of occurrence of those events is sufficiently low. However, if a design basis event were to occur with a substantially higher likelihood or the plant were to be pressurized over its design' pressure limits with higher frequency, then the plant cannot meet its design basis requirements.
Similar problems of inadequate ring settings of primary safety valves were discovered during the EPR1 test program conducted in response to NUREG-0737, item II.D.1.
The problem was identified and confined to valves manufactured by Dresser Industries, Inc.
Users of primary side Dresser valves made sub-mittals to justify continued operation of tilair facilities until the valve ring settings on their plants are readjusted consistent with the EPRI test findings.
The staff evaluated and approved those submittals.
Verification of the adequacy of all PWR primary safety val e ring settings is being pursued by the staff under melti plant action MPA F-14.
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'1 H. Thompson, Jr. NOV 2 21985 l
Based on the limited information available about the adequacy of the secondary side safety valves for overpressurization mitigation, the staff has a reason to doubt that all the safety valves currently in use or planned for use would serve their intended safety function if called upon.
Therefore, we suggest that the Division of Licensing send a request for additional information (RAI),
per 10 CFR 50.54(f), to PWR plant owners.
This RAI would request plant owners, in light of the Seabrook experience, to study the Seabrook experience and justify to the staff that their respective plants continue to have sufficient i
overpressure protection and are within their safety analyses. The owners' justification may rely on any combination of: (a) relevant experience from which valve performance can be verified; (b) safety analyses assuming inadequate overpressure protection; or (c) representative valve testing.
.f.-i p 31 203E73 j
Robert L:. I:02C Robert W. Bernero, Director Division of Systems Integration
Enclosure:
Suggested 10 CFR 50.54(f) Letter cc:
Edward Jordan, IE R. Baer, IE Mark Caruso,- DL DISTRIBUTION:
Docket File DSI:AD:RS rdg.
~
DSI:D RSB rdg.
RSB s/f MSSV Safety BSheron LMarsh SDiab rdg.
SDiab:js Doc Name:
Diab Rush, js
~
A
- DSI:RS, ] :DSI:RSBlQ:
-kp:DSI:R DSI:ME f
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- LMarsh FCherny
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- RBernero 7___7____...
DATE :11/2I/85
- 11/jM/85
- 11/2 //85
- 11 4r/85
- 11/l7/85
- </ /2 L/M :
OFFICIAL RECORD COPY
. =
=
l ENCLOSURE i
PROPOSED 10 CFR 50.54(f) LETTER Safety Implications of Main Steam Safety Valve. Flow Deficiency During the last part of 1984 through aid 1985, Wyle Laboratories conducted several full flow tests on two MSSVs manufactured by Crosby Valve and Gage Company.
These two valves were chosen as a representative sample for the MSSVs that are to be installed on the Seabrook plant main steam lines. The purpose of the Wyle tests was to determine the adequacy of different discharge piping configurations. Test measurements indicated that the MSSVs (with the i
manufacturer's recommended guide ring settings) have a flow capacity of about 50% of their design values at the design pressure. The guide ring settings determine t;ie force being_ exerted on the valve disc, thereby affecting the degree of valve lift and subsequently the discharge flow capacity of the valve.
During the Wyle tests the guide ring settings of both valves were i
substantially adjusted in order to achieve the full flow capacities.
l Subsequent to the completion of the tests, the Public Service Company of New Hampshire, the owner of the Seabrook plant, concluded that in order to ensure full flow capacity of the plant's MSSVs they all should be adjusted downward by 130 notches from the manufacturer's settings.
l
2-e While the Seabrook experience shows a deficiency in the capacity adjustment of the Crosby spring loaded safety valves, it strongly suggests a similar heficiencyinsimilarvalvesmadebyothermanufacturers,sincetheyallwork on the same basic concept.
Based on the Seabrook experience and with the lack of sufficient data, it may j
1 be assumed that a number of deficient safety valves are installed in some
)
operating plants and/or planned to be installed in plants yet to operate. It may also be assumed that the capacity of some of these deficient valves may be as low as 50% of the design flow (as for the Seabrook plant), or even lower.
With this potential deficiency, the design basis of the affected plants cannot l
be met.
The design basis of every pressurized water rezetor (PWR) requires that overpressure protection for the primary and secondary sides of the plant be provided so that the pressure does not rise above 110% of the design value during postulated events.
PWR vendors perform safety valve sizing analyses such that these valves have sufficient capacity to mitigate the most severe overpressurization event with sufficient margin.
Generic transient analyses, as opposed to sizing analyses, performed by Westinghouse for the Westinghouse PWRs show that, for the worst overpressurization event (loss of load without condenser bypass).the MSSVs' peak relieving capacity required is about 80% of the nominal valve flow.
It should also be noted that any flow degradation
~
through the MSSVs increases the potential for the actuation of the primary i
safety valves and increases the relieving load on these valves as well. This is because any relief through the MSSVs will remove heat from the reactor coolant, which would otherwise accumulate, expand, and overpressurize the i
r-
+
--_.-,-m--
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,m.we.,.
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~3-primary system.
General Design Criterion 15 of 10 CFR 50, Appendix A, requires that the reactor coolant system and associated auxiliary systems be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during normal operation and anticipated operational occurrences.
There is insufficient test data available to the staff to show whether the valve flow would increase at high upstream pressure conditions ar.d/or if the valve disc travel will be any higher.
Therefore, it is not clear whether overpressurizations of this nature are self limiting.
With less than the required reli.eving capacity, the secondary side is expected to be overpressurized, thus increasing the potential for steam side leaks or breaks. While the overpressurization described above would occur following a loss of load event, which is an anticipated operational occurrence, the con-sequences of that event may lead to a design basis event with its associated severe consequences. While plants are designed to acceptably accommodate their design basis events, the probability of occurrence of those events is suf-ficiently low.
However, if a' design basis event were to occur with a sub-stantially higher likelihood or the plant were to be pressurized over its design pressure limits with higher frequency, then the plant cannot meet its design basis requirements.
Similar problems of inadequate ring settings of primary safety valves were discovered during the EPRI test program conducted in response to NUREG-0737,
_4 item II.D.1.
The problem was identified and confined to valves manufactured by Dresser Industries, Inc.
Users of primary side Dresser valves made sub-mittals to justify continued operation of their facilities until the valve ring settings on their plants are readjusted consistent with the EPRI test findings.
The staff evaluated and approved those submittals.
Verification of the adequacy of all PWR primary safety valve ring settings is being pursued by the staff under multi plant action MPA F-14.
]
Based on the limited information available about the adequacy of the secondary side safety valves for overpressurization mitigation, the staff has a reason to doubt that all the safety valves currently in use or planned for use will serve their intended safety function if called upon. Therefore, per 10 CFR 50.54(f),
i the staff requests that you, as a PWR owner, and in light of the Seabrook experience,-justify that your plant continues to have sufficient overpressure protection, your facility continues to be in compliance with GDC 15, and the plant is within its safety analyses.
Your justification may rely on any combination of: (a) relevant experience from which v'lve performance can be a
verified; (b) safety analyses assuming inadequate overpressure protection; or (c) representative valve testing.
Hugh L. Thompson, Director Division of Licensing I
1
- - -. A
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g8 "84 9
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- o, UNITED STATES NUCLEAR REGULATORY COMMISSION 3
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j WASHINGTON, D. C. 20666
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Docket Nos. 50-317 and 50-318 MEMORANDUM FOR:
Ashok C. Thadani, Director PWR Project Directorate #8 Division of PWR Licensing-B FROM:
D. H. Jaffe, Project Manager PWR Project Directorate #8 Division of PWR Licensing-B
SUBJECT:
SUMMARY
OF MEETING WITH BALTIMORE GAS & ELECTRIC COMPANY (BG&E) CONCERNING OPERABILITY OF STEAM LINE SAFETY VALVES On November 26, 1985, representatives of BG&E and the NRC staff met in Room 2242 of the Air Rights Building in Bethesda, Maryland. Enclosure 1 contains the list of attendees. The purpose of the meeting was to discuss the operability of the Calvert Cliffs Unit 2 Main Steam Safety Valves (MSSVs).
The NRC staff indicated concern regarding the as-found setpoints for the Unit 2 MSSVs. The setpoints violated existing Technical Specification (TS) requirements and would have violated the new, more liberal (TS) requirements to be issued as part of a Unit 2, Cycle 7, license amendment.
BG&E responded to staff concerns regarding the Unit 2 MSSVs in a presentation summarized in Enclosure 2.
Although several minor MSSV problems seemed to exist, no single or cumulative cause for the setpoint problem could be identified. Evidence presented by BG&E seemed to point to a problem associated with setpoint measurement techniques.
In light of this finding, BG&E connitted to the following corrective actions related to MSSV setpoint verification:
Procedural Enhancements Set valves at 530*F vice 500*F Provide QC coverage while verifying setpoints Independently reverify setpoints of 4 valves 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after initial setting.
Verify setpoints of 4 valves during first outage after 4 month operation.
BG&E also presented its conclusions regarding the Unit 1 MSSV setpoints.
This material is contained in Enclosure 3.
w ee a,
O e
- - Following the BG&E presentation and discussions among the NRC staff, it was concluded by the NRC staff that based upon information presented by BG&E:
(1) safety analyses perfonned by BG&E, assuming as-found MSSV setpoints, i
showed no violation of safety limits or the criteria of 10 CFR 50.46, (2) improvements proposed by BG&E would likely improve MSSV setpoint measurement, and (3) the Unit 1 MSSVs appeared to be showing as-expected MSSV setpoint behavior. Based upon the above, it was concluded that no safety problems associated with MSSVs could be identified which would prevent the l
return of Unit 2 to power operati g
i-O. H.
f,4 roject Manager PWR Project Directorate #8 Division of PWR Licensing-B
Enclosures:
As stated cc w/ enclosures:
See next page M
e W
ee e
- - ~
,*---w
..._r..g
___y
t Mr. A. E. Lundvall, Jr.
Baltimore Gas,a Electric Company Calvert Cliffs Nuclear Power Plant cc:
Mr. William T. Bowen, President Regional Administrator, Region I Calvert County Board of U.S. Nuclear Regulatory Comission Commissioners Office of Executive Director Prince Frederick, Maryland 20768 for Operations 631 Park Avenue General Counsel ~
King of Prussia Pennysivania 19406 D. A. Brune, Esq.
. Baltimore Gas and Electric Company Mr. Charles B. Brinkman P. O. Box 1475 Manager - Washington Nuclear Operations Baltimore, Maryland 21203 Combustion Engineering, Inc.
7910 Woodmont Avenue George F. Trowbridge, Esq.
Bethesda, Maryland 20814 Shaw, Pittman, Potts and Trowbridge 1800 M Street, NW Mr. J. A. Tiernan, Manager Washington, DC 20036 Nuclear Power Department Calvert Cliffs Nuclear Power Plant Mr. R. C. L. Olson, Principal Engineer Maryland Routes 2 and 4 Nuclear Licensing Analysis Unit Lusby, Maryland 20657 Baltimore Gas and Electric Company Room 720 - G&E Building Mr. R. E. Denton, General Supervisor P. O. Box 1475 Training and Technical Services Baltimore, Maryland 21203 Calvert Cliffs Nuclear Power Plant Maryland Routes 2 and 4 Resident Inspector c/o U.S. Nuclear Regulatory Comission Combustion Engineering, Inc.
P. O. Box 437 ATTN: Mr. R. R. Mills, Manager Lusby, Maryland 20657 Engineering Services P. O. Box 500 Mr. Leon B. Russell Windsor, Connecticut 06095 Plant Superintendent Calvert Cliffs Nuclear Power Plant
' Department of Natural Resources Maryland Routes 2 and 4 Energy Administration, Power Plant Lusby, Maryland 20657 Siting Program ATTN: Mr. T. Magette Bechtel Power Corporation Tawes State Office Building ATTN: Mr. D. E. Stewart Annapolis, Maryland 21204 Calvert Cliffs Project Engineer 15740 Shady Grove Road Gaithersburg, Maryland 20760 Mr. R. M. Douglass, Manager -
Quality Assurance Department.
Baltimore Gas and Electric Company Fort Smallwood Road Complex P. O. Box 1475 Baltimore, Maryland 21203
e List of Attendees E
D. Jaffe T. Foley F. C. Cherny A. Thadani M. Caruso G. Hanner M. S. Wegner R. Perfetti
-868E L. B. Russel J. F. Williams J. A. Mihalcik J. T. Carroll R. R. Allen
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l ENCW6URE 2 AGENDA U-2
. Describe Safety Valves
. Describe As-Found Results
. Results of Safety Analysis
. Outline Test Program
. Test Program Results
. Conclusions
+
. Future Actions e
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STEAM GENERATOR #22 N
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l MSSV UNIT 2 LIFT SETPOINT TEST (AS FOUND/AS LEFT PSIG) l As Found With Setpoint Hydroset Valve (t 1% PSIG)
Oct.1982 Apr.1984 Oct.1985 Correction 3992 985 1009/982 991/991 987 i
3993 985 963/975 1015/985 1011 3994 995 1022/996 1001/1001 997 i
3995 995
% 2/989 1035/1004 1031 3996 1015 1021/1021 1037/1006 1024/1024 1020 3997 1015 1015/1015 1044/1008 1020/1020 1016 3998 1035 929/1035+10 1065/1038 1044/1044 1040 3999 1035 1032/1032 1053/1038 1057/1038 1053 4000 985 974/981 1037/993 1033 4001 985 987/987 972/985 1040/992 1036 4002 995 998/998 1004/1004 1059/993 1055 4003 995 996/996 1014/990 1047/997 1043 4004 1015 1018/1018 1070/1018 1065 4005 1015 1012/1012 1054/1018 1050 4006 1035 991/1036 1104/1035 1100 l
4007 1035 980/1034 1106/1039 1102 Test Results:
1 Low 6 Low 0 Low 0 Low 0 High 7 High 11 High 11 High 6 Sat 3 Sat 5 Sat 3 Sat mssvu2 1
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ACCIDEN'f15 PREDICTING MSSV OPERATION
. Loss of Load Imms of Load to One Steam Generator
. CEA Withdrawal J
. Food Line Break Lams of Non-Emergency AC Power
. Lams of Feedwater
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. Sean Break LOCA I
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_ MAIN STEAM SAFETY VALVE SAFETY ANALYSIS
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Nominal Aa-FoGimd As-Fomid UIC8*
Setpoint Setpoint Analysis Analysis PSIG PSIG PSIG PSIG I
EV-3993/4000 985 991/1037 1037/1037 1035**
RV-3993/4001 985 1015/1040 1040/1040 1035**
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EV-3994/4003 995 1001/1059 1059/1059 1035 l
RY-3995/4003 995 1935/1047 1047/1047 3035
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RY-3996/4004 1015 1024/1070 1070/1070 1985 i
RY-3997/4005 1015 1020/1954 1054/l054 1045 l-EV-3998/4006 1035 1044/1104 1104/1104+
1065 i
't RV-3999/4007 1035 1057/1106 1104/1106+
1965
- This configwation has brian shown to be appliemble for USC8.
l
" Small Break LOCA assumes 995 PSIG.
i RV-4006 and RV-4007 were assened stuck closed for Lams of Lead analysis.
+
EV-3998 and RV-3999 opened at 991 and 1001 PSIC, respectively.
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l RESULTS SMALL BREAK LOCA l
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. AvaDable High Pressure Safety Injoetion Dow in higher than ammuned la analysis, i
i Higher flow conspensates for higher M35V setpoints.
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j NON-LOCA SAFETY ANALY3B IJMTTING EVENTS
. Lams of Lead
. Loss of Load to one steam generator i
LDEFING PARAMETER Peak seeandery systesa premmere Limit less than lite PSIA.
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O RESUL115 NON-LOCA SAFETY ANALYSB i
LOSS OF LOAD TO ONE STEAM GENERATOR
. As-found setpoints used
. - Otherwise identieel to UICS analysis 6
. Peak maanadary system pressure = 1980 psia L(MB OF LOAD
. As-found setpoints used ErfC = 0.0E4 detta riso/F vs. +0.7E4 delta rho /F (U1CS)
. otherwise identieel to U1cs annaysis
. Peak smeandery system pressure = 1993 psia l
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MSSV TESTING PROGRAM i
2-RV-3993 i.
0 0
At,70 F ambient,500 F steam a.
Thermal equilibrium b.
Test set pressure c.
Full Flow Test d.
Check leakage e.
Retest set pressure.
~
2.
At 70 F ambient,530 F steam, repeat la - e.
0 0
0 0
3.
Heat up transient at 120 F ambient,530 F steam prior to thermal equilibrium, repeat Ib - d.
0 0
4.
Heat up to thermal equilibrium at 120 F ambient,530 F steam, repeat Ib - e.
5.
Set with new hydroset, check with old hydroset at 985,995,1015, and 1035 psig.
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0 0
6.
Reset to 985 psig at 120 F ambient and 530 F steam, repeat Ib - e and check with 2 full flow tests.
2-RV-3992 m
1.
0 0
At 120 F, ambient,530 F steam a.
Thermal equilibrium b.
Test set pressure c.
Full Flow Test d.
Checkleakage e.
Retest set pres,ure NOTE: For all tests, record value of temperature vs. time for inlet nozzle, body, spring, and outlet flange.
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Wyle Test Valve BM-7771 11/12 - 14/85 1.
Valve Set at 985 + 10 0 Ambi.ent - Avg. 86.3 F 0
System - Avg. 495.5 F Hydroset - Avg. 975.6 psi set pressure Full Flow - Avg. 986.3 psi pop pressure 0
2.
Ambient - Avg. 74.6 F C
System
- Avg. 537.7 F Hydroset - Avg. 964.3 psi set pressure Full Flow - Avg. 979 psi pop pressure 0
3.
Ambient - Avg.119.8 F 1 hr. heat up 0
System
- Avg. 326.2 F Hydroset - Avg. 973.8 psi Full Flow - Avg. 9Z9 psi 0
4.
Ambient - Avg.118.3 F 4 hr. heat up System
- Avg. 527'F Hyoroset - Avg. 965.8 psi set pressure Full Flow - Avg. 976 psi pop pressure 5.
Valve set at 1035 1 10 0
Ambient - Avg.119.5 F 0
System
- Avg. 527.5 F Hydroset - Avg.1035.6 psi set pressure Full Flow - Avg.1030.5 pop pressure 6.
Valve Reset at 9851 10 0
Ambient - Avg.119 F 0
System
- Avg. 323.5 F Hydroset - Avg. 967.5 psi set pressure Full Flow - Avg. 981.3 psi pop pressure 7.
Wyle Test Valve BM 7787 - 11/14/85 0
Ambient - Avg.122.7 F System
- Avg. 523 F Hydros' t - Avg. 966.3 psi set pressure e
Full Flow - Avg. 984.5 psi pop pressure wyle
T.
o 1985 Ring Setting Stem Avg. Disk Maximum As Found Valve Top Bottom Surface Run Out to Guide Set Pressure Set Pressure 2-RV (Note 2)
Film /Cond.
(Note 1)
Clearance **
(PSIG)
(PSIG) 3992
-7t
-2 very light / good OK 19t 995 987 3993
+22t
+12 very light / good OK 17t 995 1011 3994 Il3t
-15 very light /
8t 18t 1005 997 very light / good 3995 6t
-2 some wear OK 15t 1005 1031 3996
-lit
-3 very light / good 15t 15t 1025 1020 3997 12t
-2 very light / good 8t 13t 1025 1016 3998 13t heavy / good 10t 14t 1045 1040 3999
-7t
-1 heavy / good 24t 13t 1045 1053 4000 4t
-9 very light / good 21t 14t 995 1033 4001 12t
-1 heavy / good 14t 14t 995 1036 4002 95t
-3 very light / good 10t 14t 1005 1055 4003 12t
-7 heavy / good 13t 13t 1005 1043 4004 25t
-3 very light /
29t 11t 1025 1065 very light / good 4005 32t
-3 good 20t 15t 1025 1050 4006 27t
-1 very light / good 13t 13t 1045 1100 4007 22t
-1 very light / good
_18t 12t 1045
!!02 s
t = 10-3 inches
- Min. old disk to guide clearance 10t
- Min, new disk to guide clearance 15t
- 1. No effect below 0.0625"
- 2. As-Found ring positions affect setpoint by less than 1% and yield 15% or less blowdown.
l U-2 I
CONCLUSIONS t
Apparent setpoint changes not explained by as-found condition of valves.
Apparent setpoint changes may possibly be the result of measurement technique.
Rebuilt valves will perform as designed.
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U-2 FUTURE ACTIONS Procedural Enhancements 0
Set valves at 530 F vice 500'F Provide QC coverage while verifying setpoints Independently reverify setpoints of 4 valves 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after initial setting.
~
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Verify setpoints of 4 valves during first outage after 4 month operation.
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r ENCIIEURE 3 AGENDA U-1 Past Test Results Estimated Current Condition of Valves Conclusions s Future Actions -
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MSSV UNIT 1 LIFT SETPOINT TEST (AS FOUND/AS LEFT PSIG)
As Left Setpoint Hydroset Test Hydroset Test Hydroset Test Valve
(!!% PSIG)
Oct.11,1983 April 1985 June.1985 3992 985 959/975 933 3
3993 985 975/988 3994 995 995/995 3995 995 987/987 990 39 %
1015 989/1024 1024/1024 3997 1015 1014/1014
)
3998 1035 1023/1028 3999 1035 1010/1034 4000 985 990/9'90 980/980 978 4001 985 981/981 964/983 983 4002 995 961/9 %
990 4003 995 987/987 999 4004 1015 1024/1024 1016 4005:
1015 986/1010 1012 4006 1035 1058/1040 1030 4007 1035 1043/1043
~
Test Results:
1 Low 6 Low 1 High 0 High
^
3 Sat 8 Sat mssvul
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. U-1
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CONCLUSIONS Valves will open, provide full capaciti, 4
and reset as designed.
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No safety implications e
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-..,_-,,___,.--_-,.----,.,.,------.,,,--,-....,._,_-----.--_...,..a-
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1 1
U-1
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1 FUTURE ACTIONS l
Verify setpoints of all 16 valves during next outage.
~
Reset any valves outsidef 1%
If necessary to reset any valve, verify setpoint of valves during first outage after 3 months operation.
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