ML20082J003
| ML20082J003 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 04/12/1995 |
| From: | POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML20082H998 | List: |
| References | |
| NUDOCS 9504180036 | |
| Download: ML20082J003 (21) | |
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ATTACHMENT I to JPN-95-022 I
Proposed Changes to Technical Specification - l NSSS Surveillance Test intervals to Accommodate 24-Month Operatina Cycles (JPTS-95-001B) ;
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l New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59 9504180036 950412 PDR ADOCK 05000333 p PDR
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, JAFNPP TABLE 4.2-8 (cont'd)
MINIMUM TEST AND CALIBRATION FREQUENCY FOR .'
ACCIDENT MONITORNG INSTRUMENTATION instrument instrument instrument Functional Test Calibration Frequency Check
- 15. Core Spray Flow N/A Once/ Operating Cycle Once/ day
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- 16. Core Spray Discharge Pressure N/A Once/ Operating Cycle Once/ day
- 18. RHR Serwce Water Flow N/A Once/ Operating Cycle Once/ day
- 19. Safety / Relief Valve Position Indicator Once/24 months N/A Once/ month (Primary and Secondary)
- 20. Torus Water Level (narrow range) N/A Once/ Operating Cycle Once/ day
- 21. Drywell-Torus Differential Pressure N/A- Once/ Operating Cycle Once/ day i
Amendment No. 386,'j8f, 86a
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! JAFNPP I
3.5 (cont'd) 4.5 (cont'd)
Automatic Deora==urization System (ADS)
D. Automatic Deoressurization System (ADS) D.
- 1. The' ADS shall be operable with at least 5 of the 7 ADS 1. Surveillance of the Automatic Depressurization System valves operable: shall be performed at least once every 24 months as -l follows:
l a. whenever the reactor pressure is greater than 100 psig and irradiated fuel is in the reactor vessel, and . a. A simulated automatic actuation which opens all l pilot valves.
- b. prior to reactor startup from a cold condition.
- b. A simulated automatic actuation which is inhibited l by the override switches.
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.I Amendment No. M, p*,334, XM 119.
JAFNPP 3.6 (cont'd) 4.6 (cont'd)
E. Safetv/ Relief Valves E. Safetv/ Relief Valves
- 1. During reactor power operating conditions and prior to 1. At least 5 of the 11 safety / relief valves shall be bench startup from a cold condition, or whenever reactor coolant checked or replaced with bench checked valves every 24 pressure is greater than atmosphere and temperature months. All valves shall be tested every 48 months. The greater than 212 F, the safety mode of at least 9 of 11 testing shall demonstrate that each valve tested actuates safety / relief valves shall be operable. The Automatic at 1110 psig 13%. Following testing, lift settings shall be Depressurization System valves shall be operable as 1110 psig .11 %.
required by specification 3.5.D.
Amendment No. 36,26, M, .RT, },:HT, Mpr, 385, Jeg, y, m 142a
9 JAFNPP .
3.6 (cont'd) 4.6 (cont'd)
- 2. If Specification 3.6.E.1 is not met, the reactor shall be 2. At least one. safety / relief valve shall be disassembled and g placed in a cold condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. inspected every 24 months. l
- 3. Low power physics testing and reactor operator training shall 3. The integrity of the nitrogen system and components which -
be permitted with inoperable components as specified in provide manual and ADS actuation of the safety / relief valves Specification 3.6.E.1 above, provided that reactor coolant shall be demonstrated at least once every 3 months.
temperature is .s;. 212 F and the reactor vessel is vented or the reactor vessel head is removed.
- 4. The provisions of Specification 3.0.D are not applicable. 4. Manually open each safety / relief valve while bypassing . .
steam to the condenser and observe a _> 10% closure of the turbine bypass valves, to verify that the safety / relief valve has opened. This test shall be performed at least every 24 -l months while in the RUN mode and within the first 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after steam pressure and flow are adequate to perform the -
test.
- 5. The safety and safety / relief valves are not required to be operable during hydrostatic pressure and leakage testing with ,
reactor coolant temperatures between 212 F and 300 F and irradiated fuel in the reactor vessel provided all control rods are inserted.
Amendment No. A3,76, J,36, W.179, J.IPS,20(,pf,)41I 143
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JAFNPP -
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I 3.6 and 4.6 BASES (cont'd)
E. Safetv/ Relief Valves .
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i The safety / relief valves (SRVs) have two modes of operation; with the HPCI and RCIC turbine overspeed systems and the i the safety mode or the relief mode. In the safety mode (or Mark I torus loading analyses. Based on safety / relief valve' spring mode of operation) the spring loaded pilot valve opens testing experience and the analysis referenced above, the . i; when the steam pressure at the valve inlet overcomes the spring - safety / relief valves are bench tested to demonstrate that
! force holdog the pilot valve closed The safety mode of in-service opening pressures are within the nommal pressure operation is required during pressurization transients to ensure setpvints 13% and then the valves are retumed to service with . 'l' vessel pressures do not exceed the reactor coolant pressure opening pressures at the nominal setpoints i1%. In this' manner,
,l' safety limit of 1,375 psig. valve integrity is maintained from cycle to cycle.
I In the relief mode the spring loaded pilot valve opens when the The analyses with NEDC-31697P also provide the safety basis i spring force is overcome by nitrogen pressure wtuch is provided - for which 2 SRVs are permitted inoperable during continuous to the valve through a solenoid operated valve. 'Ibe solenoid power operation. With more than 2 SRVs inoperable, the margin ;
operated valve is actuated by the ADS logic system (for those to the reactor vessel pressure safety limit is siynicantly reduced, i SRVs which are included in the ADS) or manually by the therefore, the plant must enter a cold condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> operator from a control switch in the main control room or at the once more than 2 SRVs are determined to be inoperable. (See -
remote ADS panel Operation of the SRVs in the relief mode for reload evaluation for the current cycle). ,
the ADS is discussed in the Bases for Specification _3.5.D.
f A manual actuation of each SRV is performed to demonstrate Expenences in safety / relief valve testing have shown that failure that the valves are mechanically functional and that no blockage ,
or detedoration of safety / relief valves can be adequately detected exists in the valve discharge line. Valve opening is confirmed by1 if at least 5 of the 11 valves are bench tested once every 24 monitoring the iwspun=e of the turbine bypass valves and the
- months so that all valves are tested every 48 munilis. SRV acoustic monitors. Adequate reactor steam dome pressure Furthermore, safety / relief valve testing experience has - must be available to avoid damaging the valve. Adequate steam ,
demonstrated that safety / relief valves which actuate within 13% '__ . flow is required to ensure that reactor pressure can bei !
of the design pressure setpomt are considered operable (see maintamed during the test. Testing is performed in the RUN 4
' ANSI /ASME OM-1-1981). The safety bases for a single nominal' mode to reduce the risk of a reactor scram in response to small valve openmg pressure of 1110 psig are described in pressure flucuations which may occur while opening and ,
NEDC-31697P, " Updated SRV Performance Requirements for- reciosing the valves ~
~t he JAFNPP." The single nominal setpoint is set below the .
_ l reactor vessel design pressure (1250 psig) per the requirements Low power physics testmg and reactor operator training with of Article 9 of the ASME Code - Section lil, Nuclear Vessels moperable components will be conducted only when the
-_ The settmg of.1110 psig preserves the safety margins associated safety / relief valves are .
, - Amendment No. _43,444,44A G49-o 152 ,
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i ATTACHMENT ll to JPN-95-022 Safety Evaluation For Proposed Changes to Technical Specification l Surveillance Test intervals to Accommodate _
24-Month Operatina Cycles (JPTS-95-0018) !
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l New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59
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Attachment ll to JPN-95-022 Nuclear Steam Supply System SAFETY EVALUATION
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- 1. DESCRIPTION OF THE PROPOSED CHANGES 3 This application for amendment to the James A. FitzPatrick Technical Specifications-proposes to extend the frequency of the Nuclear Steam Supply System (NSSS).
sumeillance testing to accommodate a 24 month operating cycle. The specific surveillances that will be revised by this application are:
Page 86a, Item 19, Table 4.2-8, " Minimum Test and Calibration Frequency for Accident J Monitoring instrumentation," change the instrument functional test surveillance frequency of the safety / relief valve position indicator from "once/ operating cycle" to ;
"once/24 months".
Page 119, Specification 4.5.D.1, change the Automatic Depressurization System (ADS).
surveillances that are performed "during each operating cycle" to "at least once every - +
24 months". In addition, parts._ a and b 'of this specification are further clarified by changing " initiation" to " actuation" to better conform to the definition of " Simulated l Automatic Actuation" in Section 1.0. The revised specification reads: <
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" Surveillance of the Automatic Depressurization Systera shall be performed every 24 l months as follows:
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- a. A simulated automatic actuation which opens all pilot valves.
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- b. A simulated automatic actuation which is inhib.ted by the override switches."
Page 142a, Specification 4.6.E.1, change the safety / relief valve bench check from ,
"once per operating cycle" to "every 24 months." In the second sentence, change test u frequency for all valves from "every two operating cycles.'" to "every 48 months." The revised specification reads:
"At least 5 of the 11 safety / relief valves shall be bench checked or replaced with bench checked valves every 24 months. All valves shall be tested every 48 months."
Page 143, Specification 4.6.E.2, change the safety / relief valve disassembly and :
l inspection from "once/ operating cycle.*" to "every 24 months." The revised l specification reads: !
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l "At least one safety / relief valve shall be disassembled and inspected every 24 months."
Page 143, Specification 4.6.E.4, change test frequency to manually open each -
- safety / relief valve from "each operating cycle" to "every 24 months." The revised specification reads
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- 'l Attachment 11 to JPN-95-022 .l Nuclear Steam Supply System SAFETY EVALUATION
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"This test shall be performed at least once every 24_ months while in the RUN mode and within the first 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after steam pressure and flow are adequate to perform I this test."
In addition, the asterisk and footnote from pages 142a and 143 are deleted since the ~
conditions for surveillance test extension have expired.
i L Bases Page 152, Section 4.6, change bench testing requirements (third paragraph),
from "once per operating cycle so that all valves are tested every two operating cycles", Ij to "every 24 months so that all valves are tested every 48 months." The revised bases- i reads: j
" Experiences in safety / relief valve testing have shown that failure or deterioration'of safety relief valves can be adequately detected if at least 5 of the 11 valves are bench tested every 24 months so that all valves are tested every.48 months." -
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- 11. PURPOSE OF THE PROPOSED CHANGES l
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l: This application for amendment proposes to extend the NSSS surveillance test . l
- intervals to accommodate 24 month operating cycles. The proposed change in test
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!' frequency is every 24 months._ These changes are necessary to _ eliminate the need for mid-cycle outages to conduct these surveillance tests. :The proposed changes follow the guidance provided by Generic Letter 91-04, " Changes in Technical Specification l Surveillance Intervals to Accommodate 24-Month Fuel Cycle," (Reference 1)l Extension of the surveillance test intervals was evaluated and the results documented in Reference 2.
111. SAFETY IMPLICATION OETHE . PROPOSED CHANGES ~
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. Eleven safety / relief valves (SRVs) are located on the main steam lines within the.
drywell between the reactor vessel and the inboard main steam isolation valves.
Seven of these valves are Automatic Depressurization System'(ADS) valves. Each SRV is equipped with two acoustical detectors or accelerometers (primary indication),1 one of which is maintained in service. Each valve is also equipped with a backup -
thermocouple detector (secondary indication). The acoustic detectors monitor noise level and provide control room alarm upon indication of an open SRV. The j thermocouples provide continuous monitoring and recording of SRV discharge i
- temperature to detect valve leakage.
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Attachment 11 to JPN-95-022 ,
Nuclear Steam Supply System SAFETY EVALUATION Page 3 of 8 -
- 1. L Technical Specification 4.5.D.1 requires a once per cycle surveillance testing of the ;
(~ ' ADS. This surveillance test includes a simulated automatic actuation to open all pilot
' valves and a simulated automatic actuation which is inhibited by the override switches.
The existing on-line testing provides adequate assurance.of valve operability. .The l safety relief valve monitor instrument check, performed on a monthly basis,- !
demonstrates the ability of SRV_ tailpipe acoustic system to detect leaking or partially- ,
open SRVs by recording noise levels from each SRV accelerometer, r Another consideration for surveillance interval extension is the past performance of the I SRVs. Occurrence reports were reviewed which indicated that SRV setpoint drift is a >
! concem. However, the Authority has requested (References 3 and 4) and received :
' approval (Reference 5) of Technical Specification Amendment 217 which incorporated-a single nominal SRV setting of 1110 psig and a 3% setpoint tolerar.ce for the SRVs.'
The analyses performed to support this. Technical Specification change demonstrated -
the adequacy of plant piping and containment structures for the SRV setpoint and tolerance change. In addition, the Authority has re-confirmed the General Electric analysis supporting the SRV setpoint tolerance change. 'It was determined that SRV l drift with a 3% tolerance would be acceptable for (i.e., bounded by) a 24_to 30 month.-
interval (Reference 6). The Authority has also assessed potential SRV drift over a 60 i month period since proposed Specification 4.6.E.1 allows a maximum test frequency _ of -)
60 months. It was concluded that SRV drift, as measured by surveillance data, is not dependent on the time between surveillance intervals. Therefore, there is reasonable .
assurance that the SRVs will not drift excessively during a potential maximum 60 month period between surveillances. It should be noted that the NRC has approved similar SRV surveillance test extensions for Limerick Units 1 and 2 and Peach Bottom l
Units 2 and 3 as discussed in the NRC's safety evaluations for these plants (References 7 and 8, respectively).
The Authority is also participating in the BWR Owners' Group SRV setpoint drift program. Recommendations from this program are evaluated in reference to the FitzPatrick plant SRVs.
In summary, the ADS surveillance frequency can be safely extended because: i I
e Past performance indicates, aside from the setpoint drift' concerns, the SRVs have - i been mechanically reliable. !
= -On-line testing provides assurance of valve operability. Lecking or partially open SRVs -
are detected by the acoustic monitor and associated alarms.
- 2. Safety / Relief Valve Position Indicator Functional Test, item 19, Table 4.2-8.
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.. Attachment Il to JPN-95-022 Nuclear Steam Supply System SAFETY EVALUATION Page 4 of 8 Table 4.2-8 requires a once per operating cycle functional test of the safety / relief v position indicators (accelerometers and thermocouples). The performance of requires drywell access. Therefore, it can not be perfomled during power opera Test results from the previous four refueling outages were reviewed to determine system reliability. No failures of the functional test criteria were noted. Out of accelerometer readings, there were only three instances where the accelerometers failed to register. One of the failures resulted from a maintenance activity, not equipment degradation. In none of the cases were both the primary (in service backup (out of service) accelerometers simultaneously inoperable. Therefore, operators would always have the ability to detect SRV leakage or actuation.
In addition, on-line monthly instrument checks required by Technical Specification Table 4.2-8, demonstrate the operability of the safety relief valve tailpipe acoustic system by measuring noise levels from each SRV.
j l In summary, this functional test can be safe'y extended for the following reasons:
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The monthly surveillance of the SRV accelorometers and thermocouples ensures th operability.
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A review of past performance of the acoustic moaitors and thermocouples indicates they have been reliable.
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Bench testing, inspection / disassembly, and manual operation testing of SRVs, Specifications 4.6.E.1,4.6.E.2, and 4.6.E.4.
Specification 4.6.E.1 requires that 5 out of 11 SRVs be bench checked or replace with bench checked valves once per operating cycle. It also requires that all valves be tested every two operating cycles. The purpose of this requirement is to verify that SRV setpoints are within a specified tolerance.
Specification 4.6.E.2 requires the disassembly and inspection of at least one SRV each operating cycle. The purpose of this requirement is to identify early degradatio of the valves. This surveillance (and 4.6.E.1) must be performed when the reactor is in cold shutdown with the drywell de-inerted.
Specification 4.6.E.4 requires the manual operation of each SRV to observe a grea than or equal to ten percent closure of the turbine bypass valves. The manual operation of the valves also satisfies in-service testing exercise requirements, in the specifications above, the monthly testing requirements serve to provide a
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redundant and eariy means of detecting SRV degradation. As discussed previously, - J the Authority has received approval for SRV Technical Specifications changes (Reference 5) to increase the maximum permissible setpoint drift to 2 3%, consistent with ASME _OM-1-1981. The supporting evaluations for this amendment indicate that
!. ' SRV setpoint drift is not expected to be a concem over a 24 to 30 month interval since l
the historical drift data was typically within the new specified tolerances of 2 3%
it should be noted that the most recent occurrence of setpoint drift (LER 95-001,- ,
Reference 9) was reported under the provision of 10 CFR 50.73(a)(2)(i)(B) as_an !
operation of the plant in a condition prohibited _by the Technical Specifications that j
, were in effect during the past cycle (i.e., previous to amendment 217). The remote j l actuation of the ADS functions would not have been effected by this' event. The new l specifications allow continuous operation with 2 SRVs inoperable and a setpoint drift'of l
3% An analysis has shown that continuous operation of the plant would be ~
acceptable with 2 SRVs inoperable and nine SRVs actuating at 1195 psig. The analysis further confirmed that the setpoint drift of nine SRVs to the 1195 psig limit l would not adversely affect the following.
- Primary Containment Integrity system 4
- Fuel Thermal Umits !
- ECCS/LOCA performance ]
The Authority continues to subject all SRVs, rather than 5 out of 11 as specified in the ~ ,
"echnical Specifications, to testing, refurbishment, and recertification once each '
operating cycle.
The bases change to section 4.6, consistent with Generic Letter 91-04; clarifies the ~ l bench testing frequency of 24 months and does not impact any SRV safety limits.-
it should noted that the asterisk and footnote for these specifications on Technical Specification pages 142a and 143 are deleted since the conditions have expired. This :
one time extension of SRV bench checking requirements is no_ longer valid. No safety limits are impacted by this administrative change.
Based on the discussion above, these surveillance test intervals can be safely extended.
Attachment ll to JPN-95-022 Nuclear Steam Supply System SAFETY EVALUATION Page 6 of 8
-IV. EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION Operation of the FitzPatrick plant in accordance with the proposed Amendment would not involve a significant hazards consideration as defined in 10 CFR 50.92, since it would not:
- 1. involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed changes extend the surveillance test intervals for nuclear steam supply l
system components. These changes are consistent with the guidance provided in Generic Letter 91-04. The proposed changes do not involve any modification to the plant, nor do they alter equipment functions. On-line testing will provide a redundant and early means of demonstrating system operability. Based on past results, SRV mechanical performance has been good. No SRV setpoint changes are involved in this application. The proposed change to bases section 4.6 clarifies that the nuclear steam supply system surveillance testing interval is consistent with the length of the operating cycle. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed changes extend the surveillance test intervals for nuclear steam supply system components. These changes are consistent with the guidance provided in Generic Letter 91-04. The proposed changes do not affect the way in which the nuclear steam supply system operates nor alter the type of surveillance testing performed. SRV drift analyses indicate that SRV drift with a 3% tolerance would be acceptable for (i.e., bounded by) a 24 to 30 month interval. Leaking or partially open l
SRVs are detected by the acoustic monitoring system. Since the proposed changes do not modify the design or equipment of the plant, no new failure modes are -
introduced. The proposed change to bases section 4.6 clarifies that the nuclear steam supply system surveillance testing interval is consistent with the length of the operating cycle. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. involve a significant reduction in a margin of safety.
The proposed changes extend the surveillance tust intervals for nuclear steam supply system components. These changes are consistent with the guidance provided in Generic Letter 91-04. The proposed changes do not alter the configuration of the l
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Attachment ll to JPN-95-022 '
Nuclear Steam Supply System SAFETY EVALUATION Page 7 of 8 nuclear steam supply system nor change the maruer in which the system functions.
Operation of the facility remains unchanged by the proposed changes. An evaluation l of past equipment performance indicates that SRV mechanical performance has been j good. In addition, SRV drift has been analyzed to be within the allowable tolerance for the extended surveillance interval. The proposed change to bases section 4.6 clarifies i that the nuclear steam supply system surveillance testing interval is consistent with the l length of the operating cycle. Therefore, the proposed changes do not involve a l significant reduction in the margin of safety, j i V. IMPLEMENTATION OF THE PROPOSED CHANGE l
Implementation of the proposed changes will not adversely affect the ALARA or Fire Protection Programs at the FitzPatrick plant, nor will the changes affect the environment.
VI. CONCLUSION I The changes, as proposed, do not constitute an unreviewed safety question as defined in 10 CFR 50.59. That is, they: ,
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- 1. will not increase the probability nor the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the Safety Analysis Report; i
- 2. will not create the possibility of an accident or malfunction of a type different from any previously evaluated in the W N Analysis Report;
- 3. will not reduce the margin of safety as defined in the basis for any technical l specification; and l
- 4. involve no significant hazards consideration, as defined in 10 CFR 50.92.
Vll. REFERENCES l
- 1. Generic Letter 91-04, " Changes in Technical Specification Surveillance Intervals to Accommodate 24-Month Fuel Cycle," dated April 2,1991.
- 2. NYPA Report JAF RPT-RWR-00493, " Nuclear Steam Supply System Surveillance Test
, improvements," dated June,1992.
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Attachment ll to JPN-95-022 Nuclear Steam Supply System SAFETY EVALUATION Page 8 of 8
- 3. NYPA letter to NRC, dated December 20,1989, " Proposed Change to the Technical -
Specifications Regarding Updated SRV Performance Requirements and Miscellaneous - '
Changes."
- 4. NYPA letter to NRC, dated March 2,1994, Proposed Change to the Technical Specifications Regarding Updated SRV Performance Requirements and Miscellaneous Changes."
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- 3. NRC letter to NYPA, J. E. Menning to W. J. Cahill, issuance of Amendme.it 217' for James A. Fitzpatrick Power Plant, dated September 28,1994
- 6. NYPA memo NED-AP-94-379, JAF Safety Relief Valve Setpoint Change, dated August ;
16,1994
- 7. NRC letter to Philadelphia Electric Company, dated August 20,1992, "24-Month Fuel Cycle, Limerick Generating Station, Units 1 and 2."
- 8. NRC letter tu Philadelphia Electric Company, dated August 19,1992, "24-Month Fuel Cycle, Peach Bottom Atomic Power Station, Units 2 and 3."
- 9. NYPA letter to NRC, dated February 4,1995, "LER 95-001, Reactor Safety Relief Valve Setpoint Drift."
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' ATTACHMENT lli to JPN-95-022 l
Markup of the current Technical Specification pages 4 For NSSS Surveillance Test Intervals to Accommodate l 24-Month Operatina Cycles (JPTS-95-0018) 1 I
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JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333
- DPR-59 l:
, TABLE 4.2-8(cont d) ,
MINIMUM TEST AND CALIBRATION FREQUENCY FOR ACCIDENT MONITORING INSTRUMENTATION ,'
Instrument instrument instrument FunctionalTest Calibration Frequency Check
- 15. Core Spray Flow N/A Once/ Operating Cycle Once/ day
- 16. Core Spray Discharge Pressure N/A Once/ Operating Cycle Once/ day
- 17. IJPC1(RHR) Flow N/A Once/ Operating Cycle Once/ day
- 18. RHR ServiceWater Flow N/A Once/ Operating Cpcle Once/ day
- 19. Safety / Relief Valve Position inc5cator Once/OperstmgOyck N/A Once/ month (Primary and Secondary) 24 mon 4hs
- 20. TorusWaterlevel(norrow range) N/A Once/Operaung Cycle Once/ day .
- 21. Drywell-Torus Dillerential Pressure N/A Once/ Operating Cycle, Once/ day .
i Amendment No. Jaff,181
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JAFNPP -
3.5 (contd) 4.5 (cont'd) .'
D. Automatic Depresswizalian System (ADS) D. Adamatic Depressurization System (ADS)
Y Surveillance of the Automatic Depressurization System valves operable: sher be performedituring each operating as follows:
afleasf onceeve 2.4 men +hs
- a. whenever the reactor pressure is greater than 100 a. A simulated opens aR pilot poig and irradated fuel is in the reactor vessel, and valves.
actuation
- b. . prior to reactor story from a cold condtion. b. A simulated out Tnitiation which is inhibited by l the override switches. I achahon O
l Amendment No. M,M. 384,317, 119
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JAFNPP 3.c, (cont'd) 4.6 (cont'd) .
E. Safety /Rahaf Valves E. Safety /Reket Valves ever y 24 monA i
- 1. Dunng reactor power operating conditions and prior to startup 1. At least 5 of the 11 safety / relief valves shall ,
from a cold condition, or whenever reactor coolant pressure checked or replaced with bench checked v i is greater than atmosphere and temperature greater than (each operating cycle) All valves shall be tested every -
212 "F, the safety mode of at least 9 of 11 safety / relief valves Gwu up==>iiv cycies. The testing shall demonstrate ,
shall be operable. The Automatic Depressurization System tfiat each valve tes es at 1110 psig 13%. !
valves shall be operable as required by specification 3.5.D. Following testing, lift ettings shall be 1110 psig 1.1%
4Brnones.
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- The current surveillance interval for bench checkmg safety / relief valves is extended until the end of R11/C12 refueling outage ,
scheduled for January,1995. This is a one-time extension, effective only for this survedlance interval. ; The next surveillance interval will begin after the completion of the bench check testing and after the safety / relief valves are declared operable. '
Amendment No. 48,98,56,70,400,404,495,B+7, 219 142a
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3.6 (cont'd). 4.6 (cont'd) .
- 2. If Specification 3.6.E.1 is not met, the reactor shall be placed in 2. At least one safety / relief valve s_ hall be disassembled and a cold condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. inspected Once/operatino cycle.Ttf ever y 24 monfhs. ._ ,
- 3. Low power physics testmg and reactor operator training shall 3. The integnty of the nitrogen system and cuii pri. vin wluch : 1 be pornutted wish inoperable components as specified in provide manual and ADS actuation of the safety / relief valves.
Specification 3.6.E.1 above, provided that reactor coolant shall be demonstrated at least once every 3 months.
temperature is s212 T and the reactor vessel is vented or the ,
reactor vessel head is removed. .
- 4. The provisions of Specification 3.0.D are not applicable. 4. Manually open each safety / relief valve wtule bypassing steam to the condenser and observe a 210% closure of the turbine i' bypass valves, to verify that the safety #elief valve has opened.
This test while in theshall be performed RUN mode and wittun the at leas @first 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> aftere steam pressure and flow are e to perform the test.
every 29 rnonths
- 5. The safety and safety / relief valves are not required to be operable dunng hydostatic pressure and leaka0e testing with reactor coolant temperatures between 212 Y and 300 P and irradated fuel in the reactor vessel provided all control rods are inserted.
f* The current surveillance interval for disassembling and inspecung 3 at least one safety / relief valve is extended until the end of R11/C12 l refueling outage scheduled for January,1995. This is a one-time extension, effective only for this surveillance interval. The next
( surveillance interval will begin upon completion of this surveillance.]
Amendment No. 48, 79, 409, 404, +79, 495, 947, 219 l 143
( l
( .
JAFNPP -
3.6 and 4.6 BASES (cont'd)
E. Safetv/Reisef Valves l The safety #elief valves (SRVs) have two modes of operation; with the HPCI and RCIC turbine 'overspeed systems and the the safety mode or the relief mode. In the safety mode'(or Mark i torus loading analyses. Based on safety #elief valve ,
spnng mode of operation) the spring loaded pilot valve opens testing exponence and the analysis referenced above, the ,
when the steam pressure at the valve inlet overcomes the spnng safety #elief valves are bench tested to demonstrate that .
force holdng the pilot valve closed. The safety mode of in-service opening pressures are within the nominal pressure ,
operation is required dunng pressurization transients to ensure setpoints 13% and then the valves are retumed to sennce with !
vessel pressures do not exceed the reactor coolant pressure opening pressures at the nominal setpomts 11%. In this manner, safety limit of 1,375 psig. valve integnty is maintained from cycle to cycl.a.
in the relief mode the spring loaded pilot valve opens when the The analyses with NEDC-31697P also provide the safety basis spring force is overcome by nitrogen pressure which is provided for which 2 SRVs are permitted inoperable dunng continuous i to the valve through a solenoid operated valve. The scianoid power operation. With more than 2 SRVs inoperable, the margin , ,
operated valve is actuated by the ADS logic system (for those to the reactor vessel vessure safety limit is significantly reduced, SRVs which are included in the ADS) or manually by the therefore, the plant mutt enter a cold condton wilfun 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> operator from a control switch in the main control room or at the once more than 2 SRVs are determmed to be inoperable. (See remote ADS panel. Operation of the SRVs in the relief mode for reload evaluation for the current cycle).
the ADS is discussed in the Bases for Specification 3.5.D.
every 29 vard A manual actuation of each SRV is performed to demonstrate l
, Exponences in safety #elief valve testmg have show that failure that the valves are mechanically functional andr,at no blockage or deterioration of safety #elief valves can be ariary ely detected exists in the valve discharge line. Valve opening is conhrmed by if at least 5 of the 11 valves are bench tested oncher coerannorf monitoring the response of the turbme bypass valves and the Mso that all valves are tested everytwo operanng cycles >---- SRV acoustic monitors. Adequate reactor steam dome pressure Furthermore, safety #elief valve testing exponence has- 6 must be available to avoid damaging the valve. Adequate steam demonstrated that safety #elief valves which actuate withm 13% flow is required to ensure that reactor pressure can be of the design pressure setpomt are considered operable (see maintamed dunng the test. Testmg is performed in the RUN ANSI /ASME OM-1-1981). The safety bases for a single nominal mode to reduce the risk of a reactor scram in response to small valve openmg pressure of 1110 psig are desenbod in pressure flucuations which may occur while opening and NEDC-31897P, " Updated SRV Performance Requirements for reciosing the valves.
the JAFNPP." The single nominal setpoint is set below the .
. reactor vessel' design pressure (1250 psig) per the requirements . Low power physics testing and reactor operator training with of Article 9 of the ASME Code - Section Ill, Nuclear Vessels. inoperable components will be conducted only when the The settmg of 1110 psig preserves the safety margins cssociated safety #elief valves are l
. t8 rnonths. J Amendment No. 48,404, M7, 219 j 152
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