ML20213C923
| ML20213C923 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 11/03/1986 |
| From: | GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| Shared Package | |
| ML20213C907 | List: |
| References | |
| NUDOCS 8611100356 | |
| Download: ML20213C923 (6) | |
Text
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.I.
Technical Specification Change Request (TSCR) No.162 GPUN requests that the following pages be inserted into the existing Technical Specifications:
Revised pages 3-26, 3-26a, 3-26b, and 3-26c These pages are attached to this Change Request.
II.
Reason for Change Current Technical Specifications state that all eighteen (18) Main Steam Safety Valves (MSSVs) must be OPERABLE at Reactor Coolant System (RCS) temperatures greater than 250*F. However, to test the MSSVs, the Main Steam pressure must be approximately 900 psig.
Current Technical Specification may not provide for in-place testing after Valve Maintenance during cold shutdown or refueling and may not allow RCS temperature increases above 250*F.
Technical Specifications have had an additional section added to allow for RCS temperature increase to Hot Shutdown to permit testing of safety valves prior to power operation.
III. Safety Evaluation Justifying Change Plant specific analysis shc,ws that one MSSV is sufficient to relieve Reactor Coolant Pump heat and stored energy when the reactor is subcritical' by 1% A K/K for at least one hour.
One MSSV is required by the 1968 ASME Boiler and Pressure Vessel Code to be OPERABLE to relieve steam pressure.
This is confirmed by calculations which made the following assumptions:
1.
Steam conditions are dry saturated.
2.
Atmospheric dump valves (MS-V-4A/B) remain closed.
3.
The plant has been subcritical for at least one hour.
Analysis shows that one Main Steam Safety Valve.is sufficient to relieve reactor coolant pump heat and stored energy when the reactor is subcritical by 1% A K/K for at' least one hour. The 1968 ASME Boiler and Pressure Yessel Code,Section III, Article 9, Paragraph N-910.4 states that one safety relief valve is required on Nuclear Pressure Vessels to relieve steam pressure. However, this Tech. Spec. Change will require two MSSVs on each Steam Generator to provide conservative redundancy.
' IV.
No Significant Hazards Considerations GPUN has determined that this Technical Specification Change Request poses no significant hazards as defined by the NRC in 10 CFR 50.92.
8611100356 861103 PDR ADOCK 05000289
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A.
Operation of the facility in accordance with the proposed amendment would not involve a significant. increase in the probability of occurrence or consequences of an accident previously evaluated. The Technical Specification changes are to allow for testing of the MSSVs prior to Power Operation. The results of this change will not impact the events analyzed in Chapter 14 of the TMI-l FSAR and the TMI-l Reload Reports will remain bounding.
Therefore, the Technical. Specification change for MSSV testing does not involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated.
B.
Operation of the facility in accordance with the proposed amendment would.not create the possibility of a new or different kind of accident from any accident previously evaluated. Analyses have been performed to ensure that technical specification limits are not exceeded.
Results show that MSSV operability requirements are conservatively bounded by the existins analysis in all cases.
Therefore, it is concluded that the Technical Specification change for MSSV operability does not create the possibility of a new or different kind of accident from any accident previously evaluated.
C.
Operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety.
All safety criteria as described in the Technical Specification bases are preserved by the additional MSSV operability.information.
Therefore, it is concluded that the Technical Specification change for MSSV operability does not involve a significant reduction in a margin of safety.
V.
Implementation -
It is requested that the amendment authorizing this change become effective immediately upon issuance.
VI.
Amendment Fee (10 CFR 170.21) ~
Pursuant to the provisions of 10 CFR 170.21, attached is a check for
$150.00.
3.4.1.2.1 With the Reactor from 250*F to HOT SHUTDOWN and subcritical for at least one (1) hour, two (2) Main Steam Safety Valves per Steam Generator shall be OPERABLE. ' With less than two (2) Main Steam Safety Valves per Steam Generator OPERABLE, restore at least two (2) MSS Valves to '0PERABLE status for each Steam Generator within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
3.4.1.2.2 When the Reactor is above HOT SHUTDOWN, all eighteen (18) Main Steam Safety Valves shall be OPERABLE or, if any are not OPERABLE, the maximum overpower trip setpoint (see Table 2.3-1) shall be reset as follows:
Maximum Number of Maximum Overpower Safety Valves Disabled on Trip Setpoint Any Steam Generator
(% of Rated Power) 1.
92.4 2
79.4 3
66.3 With more than three (3) Main Steam Safety Valves INOPERABLE, restore at least fifteen (15) Main Steam Safety Valves to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
3.4.2 Reactor Coolant System temperature 250*F or less.
3.4.2.1 With Reactor Coolant temperature 250*F or less, at least two of the following means for maintaining decay heat removal capability shall be OPERABLE and at least one shall be in operation except as allowed by Specifications 3.4.2.2, 3.4.2.3 and 3.4.2.4.
a.
Decay Heat Removal String "A".
b.
Decay Heat Removal String "B".
c.
Reactor Coolant Loop "A", its associated 0TSG, and its associated emergency feedwater flowpath.
d.
Reactor Coolant Loop "B", its associated 0TSG, and its associated emergency feedwater flowpath.
l 3.4.2.2 Operation of the means for decay heat removal may be suspended l
provided the core outlet temperature is maintained below j
saturation temperature.
3.4.2.3 The number of means for decay heat removal required to be operable per 3.4.2.1 may be reduced to one provided that one of the following conditions is satisfied:
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a.
The Reactor is in a Refueling Shutdown condition with the Fuel Transfer Canal water level greater than 23 feet above i
the reactor vessel flange.
j 3-26 l
Amendment No. 4, 78, 119, l
b.
Reactor coolant temperature is less than 140*F with BWST level greater than 44 feet and an associated flow path through the RCS OPERABLE such that core outlet temperature can be maintained subcooled for at least 7 days.
c.
Equipment Maintenance on one.of the means for decay heat removal specified by 3.4.2.1 is required and the equipment outage does not exceed 7 days.
3.4.2.4 Specification 3.4.2.1 does not apply when either of the following conditions exist:
a.
Decay heat generation is less than 188 KW with the RCS full.
b.
Decay heat generation is less than 100 KW with the RCS drained down for maintenance.
3.4.2.5 With less than the above required loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible.
3-26a Amendment No.119
i Bases A reactor shutdown following power operation requires removal. of core decay heat. Normal decay heat removal is by the steam generators with the steam dump to the-condenser when RCS temperature is above 250*F and by the decay heat removal system below 250*F. Core decay heat can be continuously dissipated 'up to 15 percent of full power via the steam bypass to the condenser as feedwater in the steam generator is converted to steam by heat
. absorption.
Normally, the capability to return feedwater flow to the steam l
generators is provided by the main feedwater system.
i The main steam safety valves will be able to relieve to atmosphere the total j
. steam flow if necessary. During hot shutdown or below, only a minimum number of Main Steam Safety Valves need to be operable as stated in Technical Specification 3.4.1.2.1.
This is to provide Steam Generator overpressure protection during hot shutdown when hot functional testing is being performed. The minimum number of valves required to be operable allows margin for testing without jeopardizing p~1 ant safety.
Plant specific analysis shows l
that one Main Steam Safety Valve is' sufficient to relieve reactor coolant pump heat and stored energy when the reactor is subcritical by 1% AK/K for at least one hour.
One MSS valve is required by the 1968 ASME Boiler and Pressure Vessel Code Section III, Article 9, Paragraph N-910.4 to relieve steam pressure.. However, two MSSVs per Steam Generator will provide conservative redundancy.
During power operations, if Main Steam Safety Valves are inoperable, the power. level must be reduced, as stated in Technical Specification 3.4.1.2.2 such that the remaining safety valves can prevent overpressure on a turbine trip.
In _the unlikely event of complete loss of off-site electrical' power to the station, decay heat removal is by -either the steam-driven emergency feedwater pump, or two half-sized motor-driven pumps.
Steam discharge is to-the atmosphere via the Main Steam Safety Valves and controlled atmospheric relief valves, and in the case of the turbine driven pump, from the turbine exhaust.(1)
Both motor-driven pumps are required initially to remove decay heat with one eventually' sufficing.
The minimum amount of water in the condensate storage-tanks, contained in Technical Specification 3.4.1.1., will allow cooldown to 250*F with steam being discharged to the atmosphere.
After cooling to 250*F, the decay heat removal system is used to achieve further cooling.
l When the RCS is below 250*F, a single DHR string, or single OTSG and its associated emergency feedwater flowpath is sufficient to provide removal of ~
l decay heat at all times following the cooldown to 250*F.
The requirement to maintain two OPERABLE means of decay heat removal ensures that a single failure _ does not result in a complete loss of decay heat removal capability.
The. requirement to keep a system in operation as necessary to maintain the l
system subcooled at the core outlet provides the guidance to ensure that steam conditions.which could inhibit core cooling do not. occur.
Limited reduction in redundancy is allowed for preventive or corrective l
maintenance on the primary means for decay heat removal to ensure that maintenance necessary to assure the continued reliability of the systems may be accomplished.
l 3-26b i
Amendment No.119
. As decay heat ~ loads are reduced through. decay time or fuel off loading,
. alternate flow paths will provide adequate cooling for a time sufficient to take compensatory action if the normal means of heat removal is lost.
With the reactor vessel head removed and 23 feet of water above the reactor vessel flange, a large heat sink is available for core cooling. The BWST with level at 44 feet provides an equivalent reservoir available as a heat sink.
Operability of the BWST is to be determined using calculations based on actual plant data or through plant testing at the time the system is. to be declared
. operable. At such times that either of these means is determined to be operable, removal of the redundant or diverse cooling system is permitted.
Following extensive outages or major core off loading, the decay heat generation being removed from the Reactor Vessel is so low that ambient losses are sufficient to maintain core cooling and no other means of heat removal is required. The system is passive and requires no redundant or diverse backup system.
Decay heat generation is calculated in accordance with ANSI 5.1-1979 to determine when this situation exists.
l An unlimited emergency feedwater supply is available from the river via either of the two motor-driven reactor building emergency cooling water pumps for an indefinite period of time.
The requirements of Technical Specification 3.4.1.1 assure that before the l
reactor is heated to above 250*F, adequate auxiliary feedwater capability is
)
available. One turbine driven pump full capacity (920 gpm) and the two half-capacity motor-driven pumps (460 gpm each) are specified. However,- only one half-capacity motor-driven pump is necessary to supply aux.fliary feedwater i
flow to the steam generators in the onset of a small break loss-of-coolant accident.
l REFERENCES-(1) FSAR Section 10.2.1.3 i
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1 3-26c Amendment No. 36,119
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