ML20127G886
| ML20127G886 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 12/12/1977 |
| From: | Desiree Davis Office of Nuclear Reactor Regulation |
| To: | Mayer L NORTHERN STATES POWER CO. |
| References | |
| NUDOCS 9211170453 | |
| Download: ML20127G886 (13) | |
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DEisenhut Dgp' e 1977 Attorney, OELD Locket No. 50-263 I&E (3)
RPSnaider MDiggs Horthern States Power Company JWctmore ATTN: Hr. L. O. Heyer, Manager TJCarter a,
lluclear Support Services JRDuchanan 414 Nicollet Mall - 8th Floor H-
. Minneapolis, Minneseta. 55401 e
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v RE: MONTICELLO ltVCLEAR GENERATING PLANT
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In August we sent letters to a ntnter of licensees who operate Colling Water P.cactor (WR) type reactor facilities regarding survelliance requirenents for the relief and safety-relief velves that are installed in the reactor coolant systen and/or the automatic depressurization syster. The letters requested licensees to propose chanacs to their Technical Specifications to incorporate a variabic frequency test schedule for unerability testing of relief and safety-relief valves, and to insti-
? increased inspection of the relief and saf ety-relief valve line restraints in the torus. flodel Technical Specifications were included for guidance in preparing plant specific renuirenents.
Inis letter was sent to you on August 3,19U.
Fron the responses we received fron various licensees, it was evident that the liodel Technical Specifications were subject to risinterpretation in several areas. Consequently, as a result of coreents received f rom these licensees, ano further consideration by ourselves, we have revised the Model Technical Specifications to proviac sore clarification of the requirements. The major revisions, which are included in the enciesed Model Technical Specifications, are:
1.
Clarification that only the number of safety and safety-relief valves that are neded to comply with AS!!E Code requirements and the plant safety anaiysis are to be included in Model LCO 3.4.2.
The number of valves should not include any installed spares. Therefore, if a licensee's existing Technical Specifications include LCO's for spare safety or safety-relief valves, these LCO's should be inodified to exclude these valves. For some plants this represents a relaxation of previously overly restrictive requirements in this area.
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Modification of the action statenent of Model LCO 3.4.2 to address only the safety valve function of the safety relief valves. This modification makes the action statement more consistent with the safety objective of the specification which is to esture the preservation of the reactor coolant systes pressure boundaryf a
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3.' The addi' ion of a footnote to explicitly indicate that, with regard to the action statement of LC0 3.4.2, a safety-relief valve that fails to open as a result of a manual actuation signal needlnot be considered a failure of the safety valve function of the' valve.
t, This~ addition recognizes
- that in such cases the ' safety valve V '
function may still be operable and therefore a plant shutdown;
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is not required.
4.
Deletion of the required power level of less than 5t, and revision of the required reactor steam dome pressure to be inaintained during manual testinD of the safety-relief valves. The required pressure has been changed from nominal onerating pressure to any pressure greater than 100 psig.
These cnanges will provide more operating flexibility and yill significantly reduce the impact of the testing program on plant capacity factors. Moreover, the reduction in the required pressure will result in a lower incidence of the higher stresses on the relief valve line restraints that can result from testing valves at operating pressure.
It will also reduce the likelihood of aggravating pilot seat leakage.
5.
A change to the time period for observing inoperable valves for the purpose of determining the initial Hext Required Test Interval of Table 4.4-10.
This has been changed from the 16 month period beginning September 1,1977 to the 12 month period beginning furch 1,1978. This will allow additional lead time for Itcensees to develop and inplenent improved relief and safety-relief valve maintenance procedures, or other actions to improve nha retia-bility, prior to implementing the new testing program.
6.
The addition of a footnote to Table 4.4-10 indicating that valve testing is required following valve repair. maintenance or replacement. This footnote also further clarifies the treatment of valve failures or successes occuring during such testing.
Specifically, it eliminates the potential for unwarranted penalization should valves fail during testing following-completion of maintenance activities, and it explicitly gives appropriate credit for test successes obtained from such testing.
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i Northern States Power Company DEC 121977 In response to our August letter on this subject, a number of licensees described plans for improving the reliability of one type of safety-reitef valve. These plans were developed in consultation with General Some of these approaches were previously described J
Electric Company.
by GE in a meeting with the staff last spring, and are being implemented to varying degrees at several plants. On the whole, we commend these i
efforts'and are appreciative of the fact that the; industry recognizes j
the need to' improve the reliability of those valves.. However, we cannot j
agree with the contention advanced by some licensees that the variable frequency testing program would significantly ~eegrade valve reliability o
and thus should be deleted from the Model Technical Specifications.
l From a reactor. safety standpoint, we have concluded that in-situ. testing
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of these valves is the preferable method for adequately demonstrating their reliability.
While we recognize that some small amount of degradation of the valve may occur as a result of testing, we do not believe that the velve operating tine associated with the testing program l
is a significant fraction of the total average yeerly operating titue that has been historically associated with these valves. Therefore, we do not believe that the testina program will significantly contribute to valve unre11at:111ty. On the contrary, we have concluded that the variable frequency test sched01e will improve overall pressure relief i
systen reliability because it verifies system operability on a frequency based on demonstrated reliability.
Under this scheme, plants with j
well naintained and reliabic valves will not be required to perfom any additional testina beyond what is currently requirec; conversely, 4
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plants with valve failbres will be requireu to demonstrate reliability throuch additional testing. We consider the variable frecuency test scheoule to be essential to maintaining a nore unifonn level of l
reliability for this equipnent and have retained this reouirement in the flodt.1 Technical Specifications.
We request that within 30 days you propose changes to your lechnical l
Specifications that incorporate the recuirements of the enclosed revised hooel Technical Specifications.
If you have any questions, l
please contact us.
Sincerely, x
Don K. Davis, Acting Chief-Operating Reactors Branch #2 Division of Operating Reactors
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liojg}j)fe'echnical Specifications
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Northern States Power Company LEC ] 21977 cc w/ enclosure:
The Environmental Conservation Library Gerald Charnoff Esquire Shaw Pittman, Potts and Minneapolis Public Library 300 Nicollet Hall Trowbridge 1800 M Street N. W.
Minneapolis, Minnesota 55401 Washington. D. C.
20036 Arthur Rencuist. Esquire Vice Presitent - Law Northern States Power Company 414 Nicollet Mall Minneapolis Minnesota 55401 Mr. L. R. Eliason Plant Manager Monticello Nuclear Generating Plant Northern States Power company Monticello, Minnesota 55362 Russell J. Hatling. Chairman Minnesota Environmental Control Citizens Association (MECCA)
Energy Task Force 144 Holbourne Avenue. S. E.
Minneapolis, Minnesota 55414 Mr. Kenneth Dzugan Environmental Planning Consultant Office of City Planner Grace Building 421 Wabasha Street St. Paul, Minnesota 55102 Sandra S. Gardebring Executive Director Minnesota Pollution Control Agency 1935 W. County Road B2 Roseville. Minnesot) 55113 Mr. Steve Gadler 2120 Carter Avenue St. Paul, Minnesota-55108 Anthony 2. Roisman, Esquire Sheldon, Hannon & Roisman 1025 15th Street N. W.
5th Floor Washington, D. C.
20005
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SAFETY VALVES AND SAFETY-RELIEF VALVES 3/4.4.2 LIMITING CONDITION FOR OPEPATION At least the following reactor coolant system safety valves and safety-relief valves shall be operable with lift settings within + 1%
3.4.2
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of the indicated pressures.
Safetyvalves0(1240)psig
)/ Safety-relief valves 0 1100 psig Safety-relief valves 0 1090 psig I
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Safety-relief valves 0 1080 psig APPLICABILITY: CONDITIONS 1, 2 and 3.
ACTION _:
k!ith one or rore of the above required reactor coolant system safety valves or the safety valve function of one or more of the above required safety-relief valves inoperable, either restore the inoperabic valve (s) to ooerable status or be in at least HOT SHUTDOWN within 12 COLD' SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
SURVEILLANCE REQUIREMENTS in addition to the applicable ASME Boiler and Pressure tessel Code,Section XI requirements, each safety relief valve shall
- 4. 4. 2.1 be demonstrated operable:
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, by verifying bellows integrity a.
through instrument indication.
Until March 1,1979, at least once per 18 months by:
b.
Manually opening each r.motely operated safety-reliefe e
1.
valve with reactor steam dome pressure > 100 psig, and verifying each valve opens by observing that either:
' Number to be consistent with ASME Code requirements and plant safety analyses.
Do not include installed spares.
With regard to the action statement of 3.4.2 above, a failure to open on manual actuation need.not be considered a failure of the safety valve function of the valve.
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _. - ~ _ _ - - _
REACTOR COOL ANT S STEM 4
SURVEILt.ANCE REOUIREMENTS (Continued) 4 The turbine bypass or control valve (s) indicate a a.
j compensating valve movenent, or b.
The reactor coolant system pressure decreases by an amount equivalent to the valve pressure relieving capacity for the test conditions.
Conducting a visual inspection of the safety-relief valve i
2.
line restraints in the torus to verify strJctural integrity for continued operation.
Af ter March 1,1919, by perfomance of the following test program:
c.
1.
Manually oper a + F w M1y operated safety-relief valve in acco i n a 4 -
c.he St schedule of Table 4.4-10 with reent 3 u @et, pressure > 100 psig, and verifying cia.
- c.
- cpes by observTng that either:
1 a.
The turbine bypass or control valve (s) indicate a compensating valve movement, or b.
The reactor coolant system pressure decreases by an amount equivalent to the valve pressure relieving capacity for the test conditions.
2.
The initial Next Required Test Interval of Table 4.4-10 shall be detemined by the number of remotely operated i
relief and safety-relief valves found inoperable from March 1, 1976 to March 1, 1979.
3.
The initial valve tests of Table 4.4-10 shall be completed by, the earlier of:
The completion of the next refueling outage occurring a.
after March 1, 1979, or b.
The time period defined by March 1,1979 plus the initial test interval, determined above.
4 At least once per 18 months, by conducting a visual inspection of the safety-relief valve line restraints in the torus to verify structural integrity for continued operation.
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Q REACTOR COOLANT SYSTEM URVEILLANCE RE00lREMENTS (Continued)
S Each safety valve and the safety valve function of each safety-i ts of the elief valve shall be demonstrated operable per the requ remen) Edition and A 4.2.2 4
. ASME Boiler and Pressure Vessel Code (
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TABLE 4.4-10_
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REMOTELY OPERATED RELIEF AND SAFETY-RELIEF VALVE TEST i
NEXT REQUIRED t
l NUMBER OF REMOTELY OPERATED RELIEF AND SAFETY-RELIEF VALVES fEST INTERVAL
- FOUND INOPERABLE DURING OPERATION, TESTING OR TEST INTERVAL **
I 18 months + 25%
T 0
184 days T 25%
1 92 days
[25%
2 31 days
+ 25%
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Early tests may The required test interval shall not be lengthened more than on d
f the nominal h i the test interval.
time less the negative 25% tolerance band). tests of the same interval, however, tl a valve failure for the purposes of this test schedule.
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- Setpoint drift is not considered to ba OPERABLE Each affected remotely operated relief and safety-relief valve shall be demonstrated 2
1 s applicable, pursuant to Specifica tions' 4.4.2.1.b.1, 4.4.2.1.c.1, 4.5.2.b.1 and 4.5.
.c.
, a d
l mant work is within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> after exceeding 100 psig, whenever maintenance, repair or rep aceSuccessful tests p j
l" provided such tests
. performed on a valve or its associated actuator.
r.uy be used 'to satisfy the test requirenents for a " required test interva l
band.
l are perfort,ed within the current " required test interval" and its associated to erance idered inoperable Valve failures detected during testing under this provision shall not be cons i
valves for the purpose of this table.
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i 3/4.4 REACTOR _C00 TANT SYSTEK BASES l
SAFETY VALVES AND SAFETY RELIEF VALVES, l
3/4.4.2 The reactor coolant system safety valves operate.to prevent the reactor coolant system fron being pressurized above the Safety Li
_ psig in accordance with the ASME Code.1bs per hour at the valve of j
safety valve is designed to relieveThe capacity of the re11eT73afety-r i
to meet the SAR stated requirement that these valves shall function set point.
j The spring to prevent opening of the spring loaded safety valves.
loaded safety valves are not expected to be required to function under i
the most limiting transient, assuming proper reli j
order to comply with ASME Code requirements.-
The testing frequency applicable to the relief valve function of a
the safety-relief valves is provided to ensure operability and demor. st reliability of the valves. This variable frequency test schedule effective on March 1,1979. tion of_the_ Mark I Safety-Relief Valve Tne required testing interval varies with observed The number of inoperabic valves found during both generic concern.
l operation and testing of these valves determines the time interval valve failures.
Early. tests may be for the next required test of these valves.
performed prior to entering the next required tes Early tests may be used as a new reference point for tests of the same time interval; however, they are not acceptable fo
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band as required by Table 4.4-10.
Demonstration of the safety valves and safety-relief valve lif t settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Vessel Code.
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EMERGENCY CORE COOLING SYSTEMS AUTOMATIC DEPRES$UR12ATION SYSTEM LIMITING CO'tDITION FOR OPERATION The Automatic Depressurization System (ADS) shall be OPERABLE 3.5.2 with at least (6)* OPERABLE ADS valves.
APPLICABILITY: CONDITIONS 1, 2 and 3.
ACT!0ti:
With one of the above required ADS valves inoperable, operation may continue provided the actuation logic of the a.
remaining ADS valves is operable and the CSS and LPCI tystems are operable, and the' HPCI system is demonstrated operable within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; restore the inoperable ADS valve to operabic status within 14 days or be in at least HOT SHUTDOWN within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUT 00WN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
With two or more of the above required ADS valves inoperable, b.
be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
SURVEILLANCE REQUIREMENTS In addition to the applicable ASME Boiler and Pressure Vessel 4.5.2 Code,Section XI requirements, the ADS shall be demonstrated operable:
At least once per 18 months by performance of a system functional test which includes simulated automatic actuation a.
through the automatic depressurization sequence, but excluding valve actuation, Until March 1,1979, at least once per 18 months by:
b.
Manually opening each ADS valve with a reactor steam 1.
dome pressure > 100 psig, and verifying each valve opens by observing that either:
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i EMERGENCY CORE COOLING SYSTEMS _
AUTOMATIC DEpRESSURIZATION SYSTEM _
I SURVEILLANCE REQUIREMENTS (Continued)
The turbine bypass or control valyc(s) indicate a 1
a.
compensating valve movement, or The reactor coolant system pressur,e decreases by an b.
amount equivalent to the valve pressure relieving
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capacity for the test conditions.
a Conducting a visual inspection of the safety-relief and relief valve line restraints in the torus to verify 2.
structural-integrity for continued operation.
Af ter March 1,1979, by perfomance of the following test c.
program:
Manually openine each ADS valve in accordance with the 1.
test schedule of Table 4.4-10 with reactor steam dome ressure > 100 psig, and verifying each valve opens
)y observing either:
The turbine bypass or control valve (s) indicate a a.
compensating valve movement, or The reactor coolant system pressure decreases by an b.
amount equivalent to the valve pressure relieving capacity for the test conditions.
The initial Next Required Test Interval of Table 4.4-10 shall be determined by the number of remotely operated 2.
relief and safety-relief valves found inoperable from i
March 1,1978 to March 1,1979.
The initial valve tests of Table 4.4-10 shall be completed 3.
by, the earlier of:
The completion of the next refueling outage occurring a.
after March 1, 1979, or 3
The time period defined by March 1, 1979 plus the b.
initial test interval, detemined above.
At least once per 18 months by conducting a visual inspection of the :,afety-relief and relief valve line restraints in the 4
torus to verify structural integrity _for continued operation.
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f 3/4.5 EMERGENCY CORE COOLING SYSTEtt 1
BASES l
AUTOMATIC DEPRESSUR12ATION SYSTEM (ADS) i l
3/4.5.2 Upon failure of the HPCIS to function properly after a small 1
break loss-of-coolant accident, the ADS automatically causes the j
safety-relief valves to open, depressurizing the reac i
4 ADS time to limit fuel cladding temperature to less than 22000F.
is conservatively (required to be operable whenever
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pressure exceeds provide adequate core cooling up to (350) psig.
i ADS automatically controls (7) safety-relief valves althouchTherefor thesafetyanalysisonlytakescreditfor(6).
ADS '/alves are required to be OPERABLE.
l The testing frequency applicable to ADS valves is provided to The l
ensure operability and demonstrate reliability of the valves.
The required testing interval varies with observed valve failures.
l number of inoperable valves found during both operation and testing of these valves determines the time interval for the next required test Early tests may be performed prior to entering the l
of these valves.
next required test interval (i.e., in advance of the nominal time less Early tests may be used as a new the negative 25% tolerance band). reference point for tests of t l
i are not acceptable for lengthening the test interval since they were i
not performed within the +25% tolerance band as required by Table 4.4-10.
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TABLE 1.X OPEPATIONAL C0f4DIT10NS l
MODE SWITCH AVERAGE REACTOR CONDITION POSIT!0ft COOLANT TEMPERATURE 1.
POWER OPEPATION Run Any temperature 2.
STARTUP Startup/ Hot Standby Any temperature j
3.
HOT SHUTDOW!1 Shutdown
> 2120f j
4.
COLD SHUTDOWN Shutdown 1 2120f 5.
REFUELING
- P,efuel or Startup i 2120F e
a SReactor vessel head unbolted or removed and fuel in the vessel.
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