ML20205L175
| ML20205L175 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 03/31/1986 |
| From: | Devincentis J PUBLIC SERVICE CO. OF NEW HAMPSHIRE |
| To: | Thompson H Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.K.3.05, TASK-TM GL-85-12, SBN-976, NUDOCS 8604030266 | |
| Download: ML20205L175 (8) | |
Text
_.
March 31, 1986 SBN-976 T.F. B5.3.99 Office of Nuclear Reactor Regulation United States Nuclear Regulatory Commission Washington, DC 20555 Attention:
Mr. Hugh L. Thompsor., Jr., Director Division of PWR Licensing - A
References:
(a) Construction Permit CPPR-135 and CPPR-136, Docket Nos.
50-443 and 50-444 (b) USNRC Generic Letter No. 85-12, Implementation of 'IMI Action Item II.K.3.5 " Automatic Trip of Reactor Coolant Pumps,"
dated June 28, 1985.
Subject:
Response to Generic Letter 85-12
Dear Sir:
Attached is our response to Generic Letter 85-12, " Implementation of TMI Action Item II.K.3.5, ' Automatic Trip of Reactor Coolant Pumps.'"
Should you have any questions, please do not hesitate to contact us.
Very truly y urs, 4
John DeVincentis, Director Engineering and Licensing Attachment b
cc: ASLB Service List l
' ' t P.O. Box 300. Seabrook, NH O3874. Telephone (603) 474-9521
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Attachment'to SBN-976-A.
Determination of RCP' Trip Criteria
' Req $est 1.
I'dentify 'the instrumentation to lbelused to determine the RCP trip setpoint, including the degree of redundancy ~ of each parameter signal needed for the criterion chosen.-
Response-Seabrook Station has elected to use-reactor-coolant subcooling margin as the. method 'of providing guidance to operations for initiating a manual. trip of the reactor coolant pump in the event' of a small break e
LOCA. The'RCS!subcooling ' margin _is calculated based upon the wide range RCS pressure - and compensated core exit thermocouple. readings.
The value of RCS pressure utilized in the calculation is the output of the data quality algorithm implemented in the.subcooling margin monitor. The value of core exit thermocouple temperature is based upon the auctioneered high thermocouple quadrant average temperatures. Using the auctioneered. high thermocouple quadrant -
average temperature in the' calculation of core subcooling margin i.s consistent with the utilization in the WOG Emergency Response
- Guidelines (ERG).'
The: Seabrook Emergency Operating Procedure Setpoint Stu'dy presents 'the calculations which support the actual subcooling value 'of 30*F. _ The Setpoint Study calculations.were performed using pressure temperature c
- instrumentation with the largest calculated errors.. By using this approach, the operator is not -restricted to a specific type Hof
~
instrument for determination of-RCP trip. As an example, determina-tion of subcooling could be made by:
direct reading from.Subcooling Margin Monitor, i
direct reading from Main Plant Computer System, and
- manual calculation using steam tables and any combination of
- 1) MCB pressure indication (pressurizer or RCS pressure), and
- 2) MCB temperature (incore thermocouples or RCS loop wide range
~
RTDs).
Request-
~
2.
Identify the instrumentation uncertainties. for both normal and adverse
+
containment conditions. Describe the basis for the selection of Lthe '
adverse containment parameters. - Address, as appropriate, local.
conditions such as fluid jets or pipe whip which might influence the
. ins trumentation -' reliability.
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e Attachment to SBN-976 (continued)
Response
All instrumentation uncertainties used in the Emergency Operating Procedures are calculated using the methodology established by the Seabrook Station Protection System Setpoint Study. The subcooling uncertainty was obtained using temperature and pressure uncertainties and converting to subcooling using steam tables. The table below summarizes the specific values used for Seabrook.
Normal Adverse Containment Containment Temperature Uncertainty (Note - 1) i 18.6*F i 19.6*F Pressure Uncertainty (Note 2)
- 84.6 psi i 84.6 psi Raunded value of subcooling uncertainty used for RCP trip parameter 30'F 30*F Note 1: The adverse containment value for temperature uncertainty is based on engineering judgement concerning the response of RTD instrumentation in an adverse environment.
Note 2: RCS wide range pressure transmitters are located outside containment, therefore, the pressure' uncertainty is not affected by an adverse containment environment.
Request 3.
In addressing the selection of the criterion, consideration to uncertainties associated with the WOG supplied analyses values must be provided. These uncertainties include both uncertainties in the computer program results and uncertainties resulting from plant specific features not representative of the generic data group.
Response
The determination and use of RCS subcooling as an RCP trip parameter has been performed in accordance with the generic guidance provided in the Revision I version of the Westinghouse Owners Group (WOG) Emergency Response Guidelines. The WOG generic guidance includes determination of the RCP trip parameter, calculation of the RCP trip parameter, application of plant specific instrumentation errors, and finally, determination of the adequacy of the RCP trip parameter chosen.
e.
h Attachment to.SBN-976 (continued)
Response ~(continued)
The - SeabrookiStation. Emergency Operating Procedure Setpoint Study documents; the. derivation of the.Seabrook RCP trip parameter in accordance with theLWOG generic > guidelines. The_ final subcooling value meets all 'the WOG generic requirements and is therefore an
-acceptable RCP _ trip - parameter ' for use at Seabrook.
'B.-
Potential Reactor Coolant Pump Problems ERequest'
.l.-
Assure that containment isolation, including inadvertent isolation, will not cause problems. if it occurs for non-LOCA transients and
. accidents.
na.
Demonstrate that, if water services needed for RCP operations are terminated, they can be restored fast enough once a non-LOCA-situation is confirmed to prevent seal damage or failure.
- b.. Confirm that containment isolation with continued pump operation will not. lead to seal or pump ' damage or failure.
Response
~
The Reactor. Coolant Pump seals at Seabrook Station are cooled concurrently by two independent cooling. systems. During normal operation,; seal inject-lon flow from the Chemical and Volume Control System (CVCS) is provided to cool the RCP seals, and the ~ Component Cooling Water System provides flow to the thermal barrier heat exchanger _.to. limit the heat transfer from the reactor coolant ~to the RCP internals.
In the event of a inadvertent containment isolation signal the two discharge valves in the common leakoff line for the CVCS supplied seal cooling system would close. Seal return flow would continue through a relief valve that discharges to the pressurizer relief tank. The component cooling water system will not be.affected by an inadvertent containment isolation.
a.
The only credible event which would interrupt both of these sources of cooling water is a loss of offsite power. In the event of ~ 1oss of offsite power, the RCP motor is de-energized and both.
of the seal cooling supplies are terminated. Upon loss of offsite
. power, the diesel generators are automatically started and either seal injection flow or component cooling water to the thermal barrier heat exchanger is automatically restored within 12 or 32 seconds, respectively. Either of these cooling supplies is adequate to provide seal cooling and prevent seal failure due to loss of seal cooling during a loss of offsite power for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
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Attachment to SBN-976
-(continued)
Response (continued) b.
As stated above, containment-isolation with continued pump operation will not lead to seal or pump damage or failure because of Seabrook Station's system designs.
Request 2.
Identify -the-components required to trip. the RCPs, including relays, power supplies and breakers. Assure that RCP trip, when determined to
.be necessary, will occur.
If necessary, as a result of the location of any critical component, include the effects of adverse containment conditions.on RCP trip reliability.
Describe the basis -for the adverse containment parameters selected.
' Response The reactor coolant pump (RCP) circuit breakers are located in the 13.8 KV switchgear in the Non-essential Switchgear Room. These are used to trip the RCPs from the Main Control Board. Control power is provided from the non-class lE 125 VDC station battery. One battery charger is provided, which is fed from a class.14 diesel generator backed bus. This arrangement provides the most reliable source.of control power available for the non-class IE RCPs. None of these components is affected by the adverse containment conditions.
Therefore, it is concluded that the RCP trip ' function will be operable trhen required.
C.-
Operator Training and Procedures (RCP Trip)
Request 1..
Describe the operator training program for RCP trip.
Include the' general. philosophy regarding the need to trip pumps versus the desire to keep pumps running.
Response
Seabr,ok Station's training program includes training in plant specific emergency procedures that have been developed to dictate when the RCPs are to be tripped or lef t running. As stated earlier in this
. response,.Seabrook Station has elected to use the reactor coolant subcooling margin as a method of providing guidance to the operator
-for initiating a manual trip of the RCP in the event of a small break LOCA.. The low subcooling margin temperature only provides guidance and is not meant to be used alone.
Before actually tripping the RCPs, the operator must also ensure that at least one CCP or SI pump is also running.
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-4 Attachment to SBN-976 (continued)
Request 2.
Identify those procedures which include RCP trip related operations:
^
(a) RCP trip Asing WOG alternate criteria (b) RCP restart (c)~ Decay heat removal by natural circulation (d). Primary system void removal (e) -Use of ' steam generators arith and without RCPs operating
-(f)- RCP trip for other reasons s
Response
The following procedures include RCP trip related. operations:
Procedure Number Proced'ure Title E-0 Reactor Trip or Safety Injection E-0.0 Rediagnosis ES-0.1 Reactor Trip, Response ES-0.2 Natural Circulation Cooldown ES-0.3 Natural Circulation Cooldown with' Steam Void
- 4 in Vessel'(With RVLIS)
ES-0.4 Natural Circulation Cooldown with Steam Void in Vessel (Without RVLIS)
~
E-1 Loss of Reactor or Secondary Coolant ES-1.1 SI Termination ES-1.2-Post-LOCA Cooldown and Depressurization
.ES-1.3 Transfer to Cold Leg Recirculation ES-1.4 Transfer to Hot Leg Recirculation E-2 Faulted Steam Generator Isolation E-3 Steam Generator Tube Rupture ES-3.I' Post SGTR Cooldown Using Backfill ES-3.2 Post SGTR Cooldown Using Blowdown.
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' Attachment to SBN-976
'(continued)
Response (continued)~
Proced'ure Numb' r Procedure Title e
'ES-3.3 Post:SGTR Cooldown Using: Steam Dump ECA-0.0 Loss of All AC Power ECA-0.1
. Loss of All AC Power Recovery Without SI Required-ECA-0.2
' Loss of All AC Power Recovery With SI Required ECA-1.1 Loss of Emergency Coolant Recirculation-ECA-1.2-LOCA Outs'ide Containment
'ECA-2.1 Uncontrolled Depressurizat' ion of All Steam Generators ECA-3.1 SGTR.With Loss of Reactor Coolant - Subcooled.
Recovery Desired
'ECA-3.2 SGTR With Loss of Reactor Coolant - Saturated' Recovery Desired ECA-3.3
.SGTR Without Pressurizer Control FR-S.1
. Response to Nuclear Power Generation /ATWS FR-S.2
' Response to Loss of Core Shutdown FR-C.1 Response to Inadequate Core Cooling
'~
FR-C.2 Response-to Degraded Core-Cooling FR-C.3 Respouse to Saturated Core Cooling Condition FR-H.1 Response to Loss of Secondary Heat Sink FR-H.2 Response to Steam Generator Overpressure FR-H.3 Response to Steam Generator High 14 vel.
FR-H.4 Response to Loss of Stema Dump Capabilities FR-H.5
. Response to Steam Generator Low Level Attachment to SBN-976 (continued)
Response (continued)
Procedure Number Procedure Title FR-P.J-Response to Imminent Pressurized Thermal Shock Conditions FR-P.2 Responsefto Anticipated Pressurized Thermal Shock Conditions FR-Z.1 Response to High Containment Pressure FR-Z.2 Response to Containment Flooding rR-Z.3 Response to High Containment Radiation Level-FR-I.1 Response to High Pressurizer Level FR-I.2 Response to Low Pressurizer Level FR-I.3 Response to Voids in Reactor Vessel-t J..