ML20211E885
| ML20211E885 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 10/17/1986 |
| From: | MISSISSIPPI POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20211E882 | List: |
| References | |
| TAC-63175, NUDOCS 8610230233 | |
| Download: ML20211E885 (12) | |
Text
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- 1. NLS-86/09 Deletion of 2 minute requirement for stuck open SRV
SUBJECT:
Technical Specification 3.4.2.1, Action b; page 3/4 4-5 DISCUSSION: Theproposedchangewoulddeletethegminutegimelimitand change the temperature limit from 105 F to 110 F in Action Statement 3.4.2.1(b) of the Grand Gulf Technical Specifications.
JUSTIFICATION: Present Technical Specifications require that with one or more stuckopenSRV(s),thereactormodeswitchmustbeplacedinghe SHUTDOWN position, if the suppression pool temperature is 105 F or greater; or if the SRV(s) is not closed within 2 minutes. The present two minute limit does not allow enough time for operator action, as was shown in the event reported to the NRC in i LER 86-011-00 1986. Thesuppressionpgol i temperature limit datedMayf,shouldbechangedto110Finorder of 105 F '.
to be consistent with the reactor shutdown requirement in Technical Specification 3.6.3.1 Action b.2.
On April 7,1986 Grand Gulf Nuclear Station (GGNS) was required to shutdown when, following an inadvertent SRV lift, the SRV could not be verified to be closed within the two minute limit.
A technician, while resetting slave trip units, inadvertently completed the actuation logic for SRV F0518. Solenoid lights and the SRV valve open/ discharge line pressure high annunciator confirmed the valve to be open. The plant operator immediately took action to close the valve with its Division I handswitch.
The SRV could not be verified to be closed at that time, so the operator attempted to close the valve with the Division II handswitch. The Division II solenoid indicated deenergized (valve closed), however the SRV tail pipe pressure switch was still activated. After one more attempt to reclose the valve by cycling the Division I hand switch, the two minute time limit expired and the operators manually scrammed the reactor. The SRV was verified to be closed seconds later. Computer traces indicate that the valve actually closed 4 to 5 seconds before the scram.
The proposed deletion of the 2 minute time limit for stuck open relief valves would allow time for operator action while maintaining control of the ability of the Suppression Pool to perform its intended steam condensation / pressure suppression function. As discussed below, the suppression pool temperature limits and the present Action Statement in Technical
' Specification 3.6.3.1 would adequately address the situation of astuckopenSRVinregardstotherequirementtoshutdgwnthe reactor if the suppression pool temperature reached 110 F.
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1 As part of the depressurization system, the Suppression Pool
. water volume must be capable of absorbing the associated decay and structural sensible heat released during a reactor blowdown from 1060 psia. In order to ensure steam condensation / pressure suppression, Technical Specification 3/4.6.3 requires that the Suppression Pool level and temperature be maintained within defined limits and that these limits be verified at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. With either Suppression Pool water level or temperature outside the prescribed limits, appropriate action statements are entered to restore the Suppression Pool to an acceptable status.
Suppression Pool temperature is monitored by 24 separate sensors, four sensors in each of six sectors, with temperatures indicated in the control rogm. Alarms sound in the control room when temperature exceeds 95 F. Specific Technical Specification limitsonSuppressionPooftemperaturerequirethatthemaximum average temperature be 95 F except under certain conditions as prescribedbyTechnicglSpecifications3.6.3.1.b. Should the temperatureexceeg95,stepsaretakentorestorethe temperature to 95 within24hoursorbeinH0TSHUTgGWNwithin the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Should the temperature reach 110 F at any time, the reactor mode switch must be placed in the SH'JTDOWN position and at least one Residual Heat Pemoval loop placed in the Suppression Pool Cooling mode.
When an SRV sticks open, its functions of protecting the reactor pressure vessel from overpressurization and allowing for vessel depressurization are being satisfied. Therefore the primary concern is to maintain the capability to condense steam in the Suppression Pool, and by that, prevent overpressurization of containment from the bypassing of uncondensed steam. The instrumentation described above, in combination with Technical Specification limits and action statements for the suppression pool, assure that this capability is maintained.
SIGNIFICANT HAZARDS CONSIDERATIONS:
The proposed change does not involve a significant hazards consideration because operation of Grand Gulf Nuclear Station in accordance with this change would not:
(1) involve a significant increase in the probability or consequences of an accident previously evaluated. The
- temperature of the Suppression Pool is limited by the actions of part 3.6.3.1 of the GGNS Technical Specifications. These limiting conditions require the reactortobepfacedinHOTSHUTDOWNshogldtheSuppression Pool exceed 110 F or should it exceed 95 F for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Therefore, if the time limit for a stuck open relief valve j is eliminated, the more practical limitation on initial Suppression Pool temperature will help to ensure that DBA pool temperatures are not exceeded.
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FSAR Section 15.1.4 " Inadvertent Safety / Relief Valve Opening" takes no credit for a reactor shutdown within two
- minutes following the determination that an SRV is stuck j open. Table 15.1-5 " Sequence of Events For Stuck Open Relief Valve" indicates that a reactor shutdown is not taken into account in'the analysis until some time following the activation of RHR Suppression Pool Cooling mode of operation which takes place' twenty minutes into the
, event.
FSAR Subsection 15.1.4.5 states that the radiological consequences of a stuck open safety relief valve are less than those associated with MSIV closure. The event is therefore bounded by the analysis performed for MSIV closure as described in FSAR subsection 15.2.4.5. Any radiological release to the Suppression Pool would be subject to plant radiation controls and monitoring to
! ensure that personnel exposures are ALARA.
i (2) create the possibility of a new or different kind of accident from any accident previously evaluated since a
parameters which experience transient conditions due to a stuck open SRV, (i.e. Suppression Pool temperature, level' and radioactivity levels) are controlled by specified limits and associated technical specification action statements.
As identified above, FSAR Section 15.1.4 takes no credit for a reactor shutdown two minutes into the event.
(3) . result in a significant reduction in the margin of safety.
The proposed change could result in fewer challenges to the Reactor Protection System. A stuck open safety relief
- valve does not in itself pose a safety hazard except as-it relates to the above parameters which are covered by specified Technical Specification limits and action statements.
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3/4.4.2 SAFETY VALVES SAFETY / RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.2.1 For the following safety / relief valves:
- a. The safety valve function of at least 7 valves and the relief valve function of at least 6 valves other than those satisfying the safety valve function requirement shall be OPERABLE with the specified lift settings, and
- b. 'The safety / relief tail pipe pressure switches for each safety / relief valve shall be OPERABLE.
Number of Valves Function Setpoint* (psig) 8 Safety 1165 + 11.6 psi 6 Safety 1180 T 11.8 psi 6 Safety 1190 7 11.9 psi Ig Relief 1103{15 psi 10 Relief 1113 + 15 psi -
9 Relief 1123115 psi APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.
ACTION:
- a. With the safety and/or relief valve function of one or more of the above required safety / relief valves inoperable, 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 next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. t to
- b. With one or more safety / relief valves stuck,open, provided thatkup:ak pression
/ pool average water temperature is less than'1958F,klose the stuck open 130 relief valve (s); " =91c te clete the eper v@!e(s) "ithia 2 i-"tet er if suppression pool average water temperature is 1959F or greater, place the reactor mode switch in the Shutdown position. Il0
- c. With one or more safety / relief tail pipe pressure switches inoperable, restore the inoperable switch (es) to OPERABLE status within 7 days or be in at least H0T SHUTDOWN within the next 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 />.
- d. With either relief valve function pressure actuation trip system "A" or "B" inoperable, restore the inoperable trip system to OPERABLE status within 7 days; otherwise 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 />.
SERVEILLANCE REQUIREMENTS 4.4.2.1.1 The tail pipe pressure switch for each safety / relief valve shall be demonstrated OPERABLE with the setpoint verified to be 30 1 5 psig by performance of a:
( *The lift setting pressure shall correspond to ambient conditions of the-valves at nominal operating temperatures and pressures.
- I nitial opening of 1821-F051B is 1103 + 15 psig due to low-low set function.
GRAND GULF-UNIT 1 3/4 4-5 Amend. men E MO -
- 2. NPE-86/14 Addition of smoke detectors to CRD repair room i
SUBJECT:
Technical Specification Table 3.3.7.9-1 page 3/4 3-85.
DISCUSSION: The proposed change will add 3 ionization smoke detectors to the Control Rod Drive (CRD) repair room at elevation 166' in Zone 2-7. This change results from a design change to add pumps, filters and an ultrasonic cleaner to the CRD repair room.
- JUSTIFICATION: The CRD repair room (Fire Zone 1A430) does not contain any safe shutdown components. It is, however, connected to Fire.
Zone 1A424 by non-fire rated double doors. Fire Zone IA424 contains safety-related equipment and Division II safe shutdown components and is also used to store combustibles during refueling operations. Fire Zone 1A424 is protected by smoke detection and automatic sprinkler system, and provides accessibility to manual hose streams and portable fire extinguishers.
- ' The addition of the new pumps, filters, and ultrasonic cleaner in the CRD repair room creates the possibility of a fire in
, that area. The addition of 3 ionization smoke detectors, along with accessibility to existing manual hose-streams and portable
, fire extinguishers, provide appropriate fire protection capability for this area.
SIGNIFICANT HAZARDS CONSIDERATION:
The proposed change will provide increased fire protection for area 1A430. Following installation, this change will not
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- rease the potential for damage to electrical. cables of safe shutdown systems.
The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated because it adds improvements not currently listed in the Technical Specifications. The potential or probability of i fire damage to electrical cables of safe shutdown systems in the adjacent areas to IA430 will not be increased since this
- change mitigates consequences of a fire related accident i initiating from this area.
The proposed change does not create the possibility of a new or different kind of accident from any accident previously 4
evaluated. This change is being proposed to reduce the possible effects of fire related accidents and does not create i the possibility of any new or different kind of accident from those presently analyzed.
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i The proposed change does'not involve a significant reduction in
, -the margin of safety because with the addition of smoke detection, the margin of safety involving fire protection of cables associated with Division II is maintained at present levels'.
Therefore,:the proposed change involves no significant hazards
- considerations.
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TABLE 3.3.7.9-1 (Continued)
FIRE DETECTION INSTRUMENTATION MINIMUM INSTRUMENTS OPERABLE
- ROOM ELEV ROOM NAME HEAT FLAME (1) SM0KE(1)
. (X/Y) (X/Y) (X/Y)
- 3. Zone 2-4 (Continued) 1A129 108' RHR "B" Heat Ex. Rm.
1A202 119' RHR "A" Heat Ex. Rm.
1A203 119' Piping Penetration Rm.
1A204 119' Piping Penetration Rm.
1A205 119' Piping Penetration Rm.
1A206 119' RHR "B" Heat Ex. Rm.
1A207 '119' Electrical Swgr. Rm. 0/3(C02 )
1A208 119' Electrical Swgr. Rm. 0/3(C02 )
1A209 115' RWCU Recirc Pump "A" Rm.
1A210 115' RWCU Recirc Pump "B" Rm.
1A223 128' Passage
- 4. Zone 2-5 5/0 1A318 139' Electrical Penetration Room 0/2(C02 )
1A319 139' RPV Instr. Test Rm.
1A320 139' Electrical Penetration Room 0/2(C02 )
- 5. Zone 2-6 26/0 1A301 139' East Corridor 1A302 139' Southeast Corridor 1A303 139' RHR "A" Heat Ex. Rm.
1A304 139' Piping Penetration Rm.
1A306 139' Piping Penetration Rm.
1A307 '139' RHR "B" Heat Ex. Rm.
1A308 139' Electrical Penetration Room 0/3(C02 )
1A309 139' Electrical Penetration Room 0/3(C0.)
2 1A314 139' South Corridor (Partial) 1A316 139' North Corridor (Partial) 14
- 6. Zone 2-7 4+/0 l 1A417 166' North Corridor (Partial) 1A420 166' South Corridor (Partial) 1A424 166' Set Down Area (Partial) 1A428 166' West Corridor 1A432 166' FPC & CU Pump Rm.
1A434 166' South Passage 164.50 166' CRO Ropd - Rm. l GRAND GULF UNIT 1 3/4 3-85 A mendment MO. --
- 3. NPE-85/17 RCIC turbine bypass valve, thermal overload protection bypass
SUBJECT:
Technical Specification Table 3.8.4.2-1, page 3/4 8-47 DISCUSSION: It is proposed to . change' the subject Technical Specification table to reflect the addition of a continuous thermal overload protection device bypass tc the RCIC turbine bypass valve, Motor Operated Valve (M0V) Q1E51-F095. This proposed change.
will provide continuous bypass of the thermal overload protection to valve Q1E51-F095 except during periodic or maintenance testing activities.
JUSTIFICATION: The function of Q1E51-F095 is to initially supply a smaller amount of steam to the RCIC turbine prior to the opening of the RCIC main steam supply valve, MOV Q1E51-F045. The initial supply of steam from Q1E51-F095 allows the RCIC turbine to operate at~ lower speeds during the starting transient and thereby establish the necessary hydraulic oil pressure needed for the operation of the turbine governor valve. The hydraulic oil pump which supplies the pressure is driven by the RCIC turbine shaft. By allowing a smaller amount of steam to enter the RCIC turbine via Q1E51-F095, potential overspeed trip conditions are avoided. Further information can be found in FSAR Section 5.4.6.
The purpose of continuously bypassing the thermal overload protection device to Q1E51-F095, except during testing activities, is to ensure that the valve performs its safety function of supplying steam to the RCIC turbine during initial system startup. By continuously bypassing the thermal overload protection, the requirements of Regulatory Guide 1.106 " Thermal Overload Protection For Electric Motor-0perated Valves" are met. MP&L committed to meet Regulatory Guide 1.106 requirements in FSAR Section 7.1.2.6.22. The thermal overload protection will be placed in service during testing activities only since, at these times, the valve is not called upon to perform its normal safety function.
This proposed change will ensure that the thermal overload I
protection device for MOV Q1E51-F095 will be continuously
! bypassed except during testing activities, thereby complying l with the requirements of Regulatory Guide 1.106.
i SIGNIFICANT HAZARDS CONSIDERATIONS:
o The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated because by continuously bypassing the thermal overload protection, the availability of QlE51-F095 to perform its safety function is enhanced. Therefore, overall system reliability is not degraded.
J16Al(C86101601 - 1
The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated because this change does not effect the original safety function of Q1E51-F095. This change complies with the requirements of Regulatory Guide 1.106.
The proposed change does not involve a significant reduction in the margin of safety because it does not adversely affect the operation of Q1E51-F095 or of the RCIC system.
Therefore, the proposed change involves no significant hazards considerations.
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TABLE 3.8.4.2-1 MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION BYPASS DEVICE (CON-l TINUOUS) (ACCIDENT SYSTEM (S) ;
VALVE NUMBER CONDITIONS) (NO) AFFECTED Q1E51F010 Continuous RCIC System Q1E51F013 Continuous RCIC System Q1E51F019 Continuous RCIC System Q1E51F022 Continuous RCIC Sysfem
, Q1E51F031 Continuous RCIC System
- Q1E51F045 Continuous RCIC System Q1E51F046 Continuous RCIC System Q1E51F059 Continuous RCIC System Q1E51F068 Continuous RCIC System RCIC Trip and Throttle Continuous RCIC System ValveggTurbineQ1E51C002 g k1F065A No Reactor Coolant System Q1821F065B No Reactor Coolant System Q1821F098A No Reactor Coolant System Q1B21F0988 No Reactor Coolant System Q1B21F0980 No Reactor Coolant System Q1821F0980 No Reactor Coolant System Q1821F019 Continuous Reactor' Coolant System Q1821F067A Continuous Reactor Coolant System Q1B21F0678 Continuous Reactor Coolant System Q1821F067C Continuous Reactor Coolant System Q1B21F067D Continuous Reactor Coolant System Q1B21F016 Continuous Reactor Coolant System MSL Drain Post LOCA Leak-Q1821F147A Continuous age Control Q1B21F1478 Continuous MSL Drain Post LOCA Leak-age Control Q1B33F019 Continuous Recirculation System Q1833F020 Continuous Recirculation System Q1833F125 Continuous Recirculation System Q1833F126 Continuous Recirculation System Q1B33F127 Continuous Recirculation System Q1B33F128 Continuous Recirculation System Q1D23F591
- Drywell Monitoring System
- Drywell Monitoring System
- Q1D23F592 Q1D23F593
- Drywell Monitoring System Q1023F594
- Drywell Monitoring System Q1E12F040 Continuous RHR System Q1E12F023 Continuous RHR System Q1E12F006A Continuous RHR System I
Q1E12F052A Continuous RHR System i
Q1E12F008 Continuous RHR System l Q1E12F394 Continuous RHR System l
GRAND GULF-UNIT 1 3/4 8-47 Amendment No. l
- 4. NPE-86/12 RHR Flush Line to LRW Table Description Clarification
SUBJECT:
Technical Specification Table 3.6.6.2-1, page 3/4 6-54 DISCUSSION: It is proposed to change the subject technical specification table item, "RHR 'A' Loop Discharge to Liquid Radwaste Valve (E12-F203)-(A)" to "RHR Discharge to Liquid Radwaste Valve (E12-F203)-(A&B)." This proposed change is strictly administrative in order to clarify the valve description in the table to reflect the actual design of GGNS.
JUSTIFICATION: There is ne design change involved. The following discussion is provided to clarify the wording change. The primary function of RHR "A&B" loops discharging to the Liquid Radwaste Surge Tank is to allow flushing of the RHR system piping. Several RHR Loop A and B flushing lines tie intu a common header which discharges to the Liquid Radwaste Surge Tank through Air Operated Valve (A0V) E12-F203 which requires air to open. A0V F203 receives air from solenoid valves SV-F537A and SV-F5378. The two solenoid valves are in turn powered by separate Er.gineered Safety Feature (ESF) power divisions "A" and "B", respectively. The two ESF power divisions provide redundancy to ensure closure of E12-F203 during a secondary containment isolation condition.
SIGNIFICANT HAZARDS CONSIDERATIONS:
The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated because this change is purely administrative in nature. This change will only reflect the actual design in the technical specification table.
i The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. This change will enhance the technical i specifications by reflecting the actual design of the RHR systen and does not create the possibility of new or different kind of accident because the actual design of the system is not affected.
The proposed change does not involve a significant! reduction in the margin of safety because the change reflects the actual design of the RHR system and does not make any new design changes.
Therefore, the proposed change to Technical Specification Table 3.6.6.2-1 involves no significant hazards considerations.
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TABLE 3.6.6.2-1 (Continued)
SECONDARY CONTAINMENT VENTILATION SYSTEM AUTOMATIC ISOLATION DAMPERS / VALVES MAXIMUM ISOLATION TIME VALVE FUNCTION (Seconds)
Valves (Continued)
PSW Aux. Bldg. Isol. Valve (P44-F122)-(A) 100 PSW Aux. Bldg. Isol. Valve (P44-F117)-(A) 100 PSW Aux. Bldg. Isol. Valve (P44-F118)-(A) 100 PSW Aux. Bldg. Isol. Valve (P44-F120)-(B) 100 PSW Aux. Bldg. Isol. Valve (P44-F123)-(B) 100 PSW Aux. Bldg. Isol. Valve (P44-F116)-(B) 100 PSW Aux. Bldg. Isol. Valve (P44-F119)-(B) 100 RHR EA" l-eep-Discharge To Liquid Radwaste Valve (E12-F203)-fat- -: 30 I
(AIB)
GRAND GULF-UNIT 1 3/4 6-54 MENbME'Nf No.