ML20076K719
| ML20076K719 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 09/09/1983 |
| From: | GEORGIA POWER CO. |
| To: | |
| Shared Package | |
| ML20076K711 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.K.3.22, TASK-TM TAC-51680, TAC-51681, TAC-52640, NUDOCS 8309150243 | |
| Download: ML20076K719 (12) | |
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NRC DOCKET 50-366 OPERATI?G LICENSC NPF-5 EDWIN I. HA'IGI NIELEAR PLANI' UNIT 2 REQUEST 'IO AMEND TECHNICAL SPECIFICATIONS The proposed change to the Technical Specifications (Appendix A to Operating License NPF-5) would be incorporated as follows:
Remove Page Insert Page 3/4 3-34 3/4 3-34 3/4 3-35 3/4 3-35 3/4 3-36 3/4 3-36 3/4 7-9 3/4 7-9 3/4 7-10 3/4 7-10 B 3/4 7-1 B 3/4 7-1 B 3/4 7-2 B 3/4 7-2 8309150243 830909 PDR ADOCK 05000366 P
h TABLE 4.3.4-1 s
REACIOR CORE ISOIATION COOLING SYSITM ACIUATION INSIMMNPATION SURVEILIANCE REQUIRDENIS c
CHAMEL H
CHAMEL MJPCTICNAL CHANNEL N
FtICTIONAL INITS CHECK TIEF CALIBRATION a.
Reactor Vessel Water Level-D M
Q Low 14w b.
Condensate Storage Tank NA M
Q Level-Iow c.
Suppression Pool Water NA M
Q Level-High w
5 TABLE 3.3.4-1
=
REACIOR CORE ISOIATICN 000LItG SYSTEM ACIUATION INSTRLMNTATION E:;
anmut NmsER &
OPERABLE CHANNEIS w
FUNCTICNAL UNITS PER TRIP SYSID1 a.
Reactor Vessel Water Level - Iow Low 2
(2B21-NO31 A, B, C, D) b.
Condensate Storage Tank 2(a)
Water Level - Iow (2E51-N060, 2E51-N061) y c.
Suppression Pool Water 2(a) y Level-High (2E51-N062A, B)
(a)
Provides Signal to RCIC Punp Suction Valves Only
l TABLE 3.3.4-2
-5eg RF1CIOR CORE ISOIATION COOLING SYSTEM AC'IUATION INSTRUMENTATION i
e ALIOKM3LE EUPCTICNAL UNI'IS TRIP SETPOINP VALUE e
N a.
Reactor Vessel Water Level - Low Low 2 -38 inches
- 2 -38 inches
- b.
Condensate Storage Tank Ievel - Iow 2 0 inches **
20 inches **
c.
Suppression Pool Water Level-High
$ 151 inches
$_151 inches
- See Bases Figure B 3/4 3-1 g
3 This corresponds to a level of 131'-0" above mean sea level.
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PLANT SYSTEM 3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM LIMITING CONDITION FOR OPERATION 3.7.3 The Reactor Core Isolation Cooling (RCIC) System shall be OPERABLE with an OPERABLE flow path capable of (AUTOMATICALLY) taking suction from the suppression pool and transferring the water to the reactor pressure vessel.
APPLICABILITY: CONDITIONS 1, 2, and 3 with reactor steam dome pressure
> 150 psig.
ACTION:
a.
With the RCIC system inoperable, operation may continue and the provisions of Specification 3.0.4 are not applicable provided the HPCI system is OPERABLE; restore the RCIC system to OPERABLE status within 14 days or be in at least HOT 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 reduce reactor steam dome pressure to < 150 psig within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
b.
With the surveillance requirements of Specification 4.7.3 not performed at the required intervals due to low reactor steam pressure, the provisions of Specification 4.0.4 are not applicable provided the appropriate surveillance is performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate to perform the tests.
SURVEILLANCE REQUIREMENTS 4.7.3 The RCIC system shul be demonstrated OPERABLE:
a.
At least once per 31 days by:
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1.
Verifying that the system piping from the pump discharge valve to the system isolation valve is filled with water,.and 2.
Verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed or otherwise secured
[
in position, is in its correct po'sition.
b.
At least once per 92 days by verifying that the RCIC pump develops a flow of 400 gpm on recirculation flow when ste'am is being supplied to the turbine at normal reactor vessel operating ' ressure, 1000 + 20, p
- 80 psig.
HATCH - UNIT 2 3/4 7-9 l
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f PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) c.
At least once per 18 months by:
1.
Performing a system functional test which includes simulated automatic actuation and verifying _that each automatic valve in the flow path actuates to its correct position, but may exclude actual injection of coolant into the reactor vessel.
2.
Verifying that the system will develop a flow of at least 400 gpm on recirculation flow when steam is supplied to the turbine at a pressure of 150 + 15, - O psig.
3.
Verifying that suction for the RCIC system is automatically transferred from the condensate storage tank to the suppression pool on a low condensate storage tank level or a high suppression pool level signal.
HATCH - UNIT 2 3/4 7-10
3/4.7 PLANT SYSTEMS BASES 3/4.7.1 SERVICE WATER SYSTEMS The OPERABILITY of the service water systems ensures that sufficient cooling capacity is available for continued operation of safety-related equipment during normal and accident conditions. The redundant cooling capacity of these systems, assuming a single failure, is consistent with the assumptions used in the accident conditions within acceptable limits.
3/4.7.2 MAIN CONTROL ROOM ENVIRONMENTAL CONTROL SYSTEM The OPERABILITY of the main control room environmental control system ensures that (1) the ambient air temperature does not exceed the allowable temperature for continuous duty rating for the equipment and instrumentation cooled by this system, and (2) the control room will remain habitable for operations personnel during the following all credible accident conditions.
The OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5 rem or less whole body, or its equivalent. This limitation is consistent with the requirements of General Design Criterion 10 of Appendix "A",10 CFR Part 50.
3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM The reactor core isolation cooling (RCIC) system is provided to assure adequate core cooling in the event of reactor isolation from its primary heat sink and the loss of feedwater flow to the reactor vessel without requiring actuation of any of the emergency core cooling equipment. The RCIC system is conservatively required to be OPERABLE whenever reactor pressure exceeds 150 psig even though the residual heat removal (RHR) system provides adequate core cooling up to 350 psig.
ihe RCIC system specifications are applicable during CONDITIONS 1, 2 and l
3 when reactor vessel pressure exceeds 150 psig because RCIC is the primary non-ECCS source of emergency core cooling when the reactor is pressurized.
Two sources of water are available to the RCIC system. Suction is initially taken from the condensate storage tank and is automatically transferred to the suppression pool upon low CST level or high suppression pool level.
With RCIC inoperable, adequate core cooling is assured by the demonstrated OPERABILITY of the HPCI system and justifies the specified 14 day out-of-service period.
HATCH - UNIT 2 8 3/4 7-1
3/4.7 PLANT SYSTEMS BASES (Continued)
The surveillance requirements provide adequate assurance that RCIC will be OPERABLE when required, Although all active components are testable and full flow can be demonstrated by recirculation during reactor operation, a complete functional test requires reactor shutdown. The pump discharge piping is maintained full to prevent water hammer damage and to start cooling at the earliest moment.
3/4.7.4 HYDRAULIC SNUBBERS The hydraulic snubbers are required OPERABLE to ensure that the structural integrity of the reactor coolant system and all other safety related systems is maintained during and following a seismic or other event initiating dynamic loads. The only snubbers excluded from this inspection program are those installed on nonsafety related systems and then only if their failure or failure of the system on which they are installed, would have no adverse effect on any safety related system.
The inspection frequency applicable to snubbers containing seals fabricated from materials which have been demonstrated compatible with their operating environment, only ethylene propylene compounds to date, is based upon main-taining a constant level of snubber protection. Therefore, the required inspection interval varies inversely with the observed snubber failures. The number of inoperable snubbers found during an inspection of these snubbers determines the time interval for the next required inspection of these snubbers.
Inspections performed before that interval has elapsed may be used as a new reference point to determine the next inspection. However, the results of such early inspections performed before the original required time interval has elapsed, nominal time less 25%, may not be used to lengthen the required inspection interval. Any inspection whose results require a shorter inspection interva? sill override the previous schedule.
To provi10 further assurance of snubber reliability, a representative sample of the installed snubbers will be functionally tested during plant shutdowns at 18 month intervals. These tests will include stroking of the snubbers to verify proper piston movement, lock-up and bleed. Observed failures of these sample snubbers will require functional testing of additional units. To minimize personnel exposures, snubbers installed in high radiation zones or in especially difficult to remove locations, as identified in Table 3.7.4-1, may be exempted from these functional testing requirements provided the OPERABILITY of these snubbers was demonstrated during functional testing at either the completion of their fabrication or at a subsequent date.
HATCH - UNIT 2 B 3/4 7-2
EtCIOSURE 2 NRC DOCKET 50-366 OPERATING LICENSE NPF-5 EIMIN I. HATCH NUCLEAR PLANP UNIT 2 REQUEST TO AMEND TECHNICAL SPECIFICATIONS Pursuant to 10 CFR 50.92, Georgia Power Company has evaluated the attached proposed amendment and has determined that its adcrption would not involve a significant hazard.
The basis for this determination is as follows:
a.
PROPOSED CHANGE Change the RCIC system Limiting Condition for Operation to require operability of the automatic suction transfer to the Suppression Pool.
BASIS This change constitutes an additional restriction not presently included in the Technical Specifications.
This change does not affect the probability or consequences of an accident or malfunction analyzed in the FSAR.
This change does not create the possibility of an accident or malfunction of a different type than any analyzed in the FSAR. The margin of safety as defined in the basis for any Technical Specification is not affected.
The effect of this change is therefore within the acceptance criteria and the change is consistent with Item (ii) of the " Examples of Amendments that are Considered Not Likely to Involve Significant Hazards Considerations" listed on page 14,870 of the April 6, 1983, issue of the Federal Register.
b.
PROPOSED CHANGE Change the RCIC system surveillance requirements such that the automatic suction transfer must be demonstrated operable at least once per 18 months.
BASIS This change constitutes an additional restriction not presently included in the Technical Specifications.
This change does not affect the probability or consequences of an accident or malfunction analyzed in the FSAR.
This change does not create the possibility of an accident or malfunction of a different type than any analyzed in the FSAR.
The margin of safety as defined in the basis for any Technical Specification is not affected. The effect of this change is therefore within the acceptance criteria and the change is consistent with Item (ii) of the " Examples of Amendments that are Considered Not Likely to Involve Significant Hazards Considerations" listed on page 14,870 of the April 6, 1983, issue of the Federal Register.
c.
PROPOSED CHANGE Add the Condensate Storage Tank low level switches and the Suppression Pool high level switches to the list of instrumentation which controls the RCIC system.
BASIS This change constitutes an additional restriction not presently included in the Technical Specifications.
This change does not affect the probability or consequences of an accident or
- malfunction analyzed in the FSAR.
This change does not create the possibility of an accident or malfunction of a different type than any analyzed in the FSAR.
The margin of safety as defined in the basis for any Technical Specification is not affected.
The effect of this change is therefore within the acceptance criteria and the change is consistent with Item (ii) of the " Examples of Amendments that are Considered Not Likely to Involve Significant Hazards Considerations" listed on page 14,870 of the April 6, 1983, issue of the Federal Register, d.
PROPOSFD CHANGE Specify allowable trip setpoints of 0 inches minimum for low Condensate Storage Tank level and 151 inches maximum for high Suppression Pool level, t
BASIS The allowable setpoint for the Condensate Storage Tank level switches ensures that the PCIC system will have an uninterrupted supply of water during suction transfer.
The allowable setpoint for the Suppression Pool level switches protects against the accumulation of an excessive vol;ume of water.
This change constitutes an additional restriction not presently included in the 4
Technical Specifications.
This change does not affect the probability or consequences of an accident or malfunction analyzed in the ESAR.
This change does not create the possibility of an accident or malfunction of a different type than any analyzed in the FSAR.
The margin of safety as defined in the basis for any Technical Specification is not affected. The effect of this change is therefore within the acceptance criteria and the change is consistent with Item (ii) of the " Examples of Amendments that are Considered Not Likely to Involve Significant Hazards Considerations" listed on page 14,870 of the April 6,1983, issue of the Federal Register.
e.
PROPOSED CHANGE Change the RCIC system instrumentation surveillance requirements to include monthly functional test and quarterly calibration of the Condensate Storage Tank level switches and Suppression Pool level switches.
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BASIS This change constitutes an additional restriction not presently included in the Technical Specifications.
This change does not affect the probability or consequences of an accident or malfunction analyzed in the FSAR.
This change does not create the possibility of an accident or malfunction of a different type than any analyzed in the FSAR.
The margin of safety as defined in the basis for any Technical Specification is not affected.
The effect of this change is therefore within the acceptance criteria and the change is consistent with Item (ii) of the " Examples of Amendments that are Considered Not Likely to Involve Significant Hazards Considerations" listed on page 14,870 of the April 6, 1983, issue of the Federal Register, f.
PROPOSED CHANGE Change bases to reflect the above changes.
BASIS This is a
purely administative change to the Technical Specifications.
This change does not affect the probability or consequences of an accident or malfunction analyzed in the FSAR.
'Ihis change does not create the possibility of an accident or malfunction of a different type than any analyzed in the FSAR. The margin of safety as defined in the basis for any Technical Specification is not affected.
The effect of this change is therefore within the acceptance criteria and the change is consistent with Item (i) of the " Examples of Amendments that are Considered Not Likely to Involve Significant Hazards Considerations" listed on page 14,870 of the April 6, 1983, issue of the Federal Register.
FJCIDSURE 3 NBC DOCKET 50-366 OPERATING LICENSE NPF-5 EIMIN I. HATCH NUCLEAR PLANT UNIT 2 REQUEST TO AMEND TECHNICAL SPECIFICATIONS Pursuant to 10CFR 170.22, Georgia Power Company has evaluated the attached proposed amendment to Operating License NPF-5 and has determined that:
a.
The proposed amendment does not requires the evaluation of a new safety analysis report or a rewrite of the facility license; b.
The proposed amendment does not require evaluation of several complex issues, involve ACRS review, or require an environmental inpact statement; c.
The proposed amendment involves one safety issue, namely the Technical Specification change associated with implementation of NUREG-0737 Item II.K.3.22.
d.
The proposed amendment is therefore a Class III amendment, 1
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