ML20032A511
| ML20032A511 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 09/30/1981 |
| From: | Udy A IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20032A510 | List: |
| References | |
| CON-FIN-A-6425, TASK-06-07.A3, TASK-6-7.A3, TASK-RR EGG-EA-5579, NUDOCS 8110300271 | |
| Download: ML20032A511 (11) | |
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EGG-EA-5579 SEPTEMBER 1981 SYSTEMATIC EVALUATION PROGRAM TOPIC VI-7.A.3, ECCS ACTUATION SYSTEM, MILLSTONE NUCLEAR POWER STATION, UNIT NO. 1 I
A. C. Udy U.S. Department of Energy Idaho Operations Office
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This is ao informal report intended for use as a preliminary or working document l
l Prepared for the U.S. Nuclear Regulatory Comission Under DOE Contract No. DE-AC07-76ID01570 FIN No. A6425 g
E S g G idaho 8110300271 811027 PDR ADOCK 05000245 P
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FORM (G4G 398 e um INTERIM REPORT Accession No.
Report No. _ EG_.G-EA-5579 Contract Program or Project
Title:
Electrical, Instrumentation, and Control Systems Support for the Systematic Evaluation Program (III)
Subject of this Document:
Systematic Evaluation Program Topic VI-7.A.3, ECCS Actuation System, Millstone Nuclear Power Station, Unit No. 1 Type of Document:
Informal Report Author (s):
A. C. Udy Drte of Document:
September 1981 Responsible NRC Individual and NRC Office or Division:
Ray F. Scholl, Jr., Division of Licensing This document was prepared primarily for preliminary orinternat use. it has not received full review and appro tal. Since there may be substantive changes, this document should not be considered t;nal.
EG&G Idaho. Inc.
Idaho Falls, Idaho 83415 Prepared for the U.S. Nuclear Regulatory Commission l
Washington, D.C.
Under DOE Contract No. DE AC07-76lD01570 NRC FIN No.
A6425 L
INTERIM REPORT
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. SYSTEMATIC : EVALUATION. PROGRAM TOPIC VI-7.A.3-ECCS ACTUATION SYSTEM
' MILLSTONE NUCLEAR POWER STATION,-UNIT NO.'l' Docket No. 50-245 SeptemberL1981 A. C. Udy EG&G Idaho, Inc.
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ABSTRACT' ThisSEPtechnicalevaluation,fortheMillstoneNuclearPowerStation, Unit No. 1,. reviews the scope and frequency of periodic testing of the,emer--
gency: core cooling system and compares the. required testing against. current.
licensing criteria.
FOREWORD LThis report is supplied as'part of-the " Electrical,. Instrumentation, and' Control Systems Support for, the Systematic Evaluation Program (.II)"
being conducted for the U.S. Nuclear Regulatory Commission, Office of 7
Nuclea'r Reactor Regulation, Division of Licensing by EG&G Idaho,'Inc.,
Reliability & Statisitics Branch.
The U.S. Nuclear Regulatory Commission funded the work under the
-authorizationiB&R 20-10-02-05 FIN A6425.
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.. CONTENTS.
1.0 : INTRODUCTION.....................................................-
C'ITERIA'........................................................
1 2.0 R
'3.0 CORE SPRAY SYSTEM...............................................
2 3~ 1 Description...............................................
2-i 3.2-Evaluation................................................
.2 4.0 LOW PRESSURE COOLANT INJECTION SYSTEf4...........................
3-4.1 Description...............................................-
3 4.2 Evaluation................................................
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5.0 FEEDWATER COOLANT lhJECTION SYSTEML..............................-
4 5.1-Description...............................................
4 5.2 Evaluation................................................
4 6.0 AUTOMATIC PRESSURE RELIEF SYSTEM................................
4 6.1 Description...............................................-
4-6.2 Evaluation.........................................'.......
5 7.0 -
SUMMARY
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8.0 REFERENCES
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'SEP TECHNICAL EVALUATION TOPIC VI-7.A.3 ECCS ACTUATION SYSTEM MILLSTONE NUCLEAR POWER STATION, UNIT NO. 1 1.0- INTRODUCTION
'The objective'of this review-is to determ;ne if all Emergency Core-
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Cooling System (ECCS) components, including pumps and valves, are included in component and system tests, if the scope and frequency.iof: periodic test-ing-are identified, and if the test program meets current-licensing crite-ria.
The.. systems included in the ECCS-are the Core Spray-system,'the Low:
PressureCoolent' Injection (LPCI) system,ctheFeedwaterCoolagtInjection (FWCI) system and the Automatic Pressure Relief (APR) system 2.0 CRITERIA
' General Design Criterion 37-(GDC 37), " Testing of Emergency Core Cool-ing Systems," requires'that:
The ECCS_be designed to permit appropriate periodic pressure and.func-tional testing to assure the operability of the system as a whole~and to verify,_under conditions as close to design as practical, the per-formance of the full operational sequence.that brings the system into
. operation, including operation of applicable portions of the protec-tion system, the transfer between nornal and emergency poyer sources, and the operation of the associated ccoling water system ?
Branch Technical Position ICSB 25, " Guidance far the Interpretaticn of GDC 37 for Testing the Operability of the Emergency Core Cooling System as a Whole," states that:
All ECCS pumps should be included in the system test.3 Regulatory Guide 1.22, " Periodic Testing of the Protection System Actuation Functions," states, in Section 0.1.a, that:
-The periodic tests should duplicate, as closely as practicable, the performanceghatisrequiredoftheactuationdevicesintheeventof an accident.
Standard Review Plan, Section 7.3, Appendix A, "Use of IEEE Standard 279 in the Review of the ESFAS and Instrumentation and Controls of Essential Auxiliary Supporting Systems," states, in Section ll.b, that:
Periodic testing'should duplicate, as closely as practical, the inte-grated performance from the ESFAS systems and their essential auxiliary supporting systems.
If such a " system level" test can be performed only during shutdown, the testing done du' ring power operation must be 1
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. reviewed in detail. -Check that'" overlapping" tests do, in; fact, over-lap from'one test' segment to another. 'For example, closing a circuit' breaker with-the manual ~ breaker control switch may not be adequate to test the ability of the ESFAS to close the breaker.5
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Regulatory Guide 1.22 states, in Section D.4, that:
Where actuated equipment is not tested during reactor operation, it should be shown that:
1.
There is no practical system design that would permit operation of the actuated equipment without adversely affecting the safety or. operability of the plant 2.
The probability that the protection system will fail to initiate the operation of the actuated equipment-_is,.and can be maintained, acceptably low without testing the' actuated equipment during
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reactor operation 3.
Theactuatedgquipmentcanberoutinelytestedwhenthereactor is thut down.
2.0 CORE SPRAY SYSTEM 3.1 Description. The Core Spray System consists-of two independent full-capacity subsystems which are designed to start automatically and pump water f-om the suppression pool (the condensate storage. tank can also be used, but is a manual connection) into the reactor vessel, to be dispersed inside of the inner shroud directly above the fuel region. Each subsystem can be started manually in addition to the automatic initiation by.the-control logic.
Each subsystem has two independent control logic channels. Four low-low reactor water level and four drywell high pressure sensors,
-connected in a one-out-of-two taken twice logic, will initiate the logic channels in both subsystems. Each of the two -logic channels will con'plete the equipment control functions for its respective subsystem. The system.
selector switch permits manual testing of each logic channel without the normal interaction (start) of the other core spray system with the reactor operating.
3.2 Evaluation.
The design of the Core Spray System allows testing, including motors and valves, for each subsystem during reactor operation by opening a test valve to the suppression pool and closing a block valve to the reactor. By use of the system selector switch the other subsystem will be available for use if required.
Independently, the admission valves and, the check isolation valves are also tested.
The Millstone-Unit 1 Technical Specifications 7 require the following tests and surveillance for the core spray system:
Simulated Automatic Actuation Test--each refueling outage a.
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Pump Flow Rate--ofter pump. maintenance and every three months
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Pump:Operabi.lity--once per month-d.
Motor operated valves--once per month e.-
' Core spray header op instrumentation--check, once per day, calibrate and test once per three months.
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~The spray pump space coolers, part of the Turbine Building secondary
-closed cooling water system (cooled by the service water system), are not
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required to be tested by the. unit technical specification.
. The Core Spray System is not tested from the automatic actuation devices through to the establishment.of flow through the test bypass valve during reactor operation.. This is done on a refueling basis. However of.the motor operated valves and each'of the pumps are tested monthly.8.all-The pump tests verifies that eacn pump operates'at rated flow. These over-lapping test demonstrate the operability of the actuated equipment during Y reactor operation. Thus, the required testing for the core spray system does conform with the testing criteria of Regulatory Guide 1.22 and GDC 37.
However, ectuation and operation of the spray pump space coolers is not required by technical specification as GDC 37 requires.
4.0 Low Pressure Coolant Injection System 4.1 Description. The Low Pressure Coolant Injection (LPCI) system consists of pumps, piping, valves and instrumentation to inject coolant f rom the suppression pool into the reactor vessel. Thera are two separate cross-tied loops, each consisting of a heat exchanger and two LPCI pumps.
The cross-tie permits flow from one loop to utilize the heat exchanger of the opposite, loop. The design for'the LPCI system requires a minimum of three of the four pumps and one of the two heat exchangers to be operable.
The two heat exchangers are cooled by independent Station Emergency Service' Water subsystems.
Automatic actuation logic similar to that used for the Core Spray system, and using the same primary detectors, initiates each LPCI loop.
Admission valves are opened, after the LPCI pressure is established, to those recirculation loops that do not have an indicated pipe rupture.
4.2 Evaluaticn.
The testing and surveillance requirements of the Millstone-Unit -1 Technical Specifications for the LPCI system are the same as the Technical Specification ~ testing and surveillance requirements for the Core Spray System, and are, therefore, also in compliance with Regula-tory Guide 1.22 and GC0 37. Additionally, the technical specifications only require that (any) three LPCI pumps be able, together, to establish a flow' rate of 15,000 gallons per minute.
It does not establish that each set of~three LPCI pumps meet this-criterion.
Should a pump fail, the testing does not verify that any three remaining pumps have the capacity required to mitigate the consequences of an accident.
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The Station Emergency Service Water subsystem is required to be tested by Technical Specification: Pump & Valve operability--once per three months and flow rate test--after pump maintenance and every three months. Testing requirements do not verify that this system will be actuated along with the LPCI system. Thus, the GDC 37 requirement for full operational test sequence, including associated cooling water systems, is not complied with.
5.0 FE_EDWATER COOLANT INJECTION SYSTEM 5.1 Description. The Feedwater Coolant Injection (FWCI) system is provided to inject coolant, at a high pressure, to the reactor vessel under small break conditions, where the Core Spray and LPCI systems do not provide sufficient head to inject coolant into the reactor vessel or the core spray spargers.
In the process of injecting coolant, the-FWCI system also reduces the pressure in the reactor vessel. Coolant is from the condensate storage-tank via the turbine condenser hotwell.
The operating conditions for the FWCI system components during a loss of coolant accident are the same as they are normally exposed to, as the FWCI system utilizes components from the feedwater system.
The actuation for the FWCI system is from the feedwater level control, and will initiate FWCI system operation at the reactor vessel water level low-low setpoint. Three condensate pumps, three condensate booster pumps and two reactor feed pumps are part of the FWCI system.
5.2 Evaluation.
In addition to the normal operation of the FWCI system components as part of the feedwater system, the FWCI concensate transfer system is tested, per technical specification requirements, af ter pump maintenance and every three months to verify flow capability at a system head. A simulated automatic actuation test is required at each refueling outage. Additionally, a weekly check of the condensate storage tank water level is required.
Testing of the Turbine Building closed cooling water system and the turbine building secondary closed cooling water system (and associated Service Water system) are not required by Technical Specifications. How-ever, these are normally operated when the FCWI components are utilized as part of the feedwater system.
The required testing and surveillance requirements in addition to the normal operation of the FWCI system components meet the intent of current licensing criteria for the testing of this ECCS actuation' system.
6.0 AUTOMATIC PRESSURE RELIEF SYSTEM 6.1 Description. The Automatic Pressure Relief (APR) system provides automatic blowdown of the reactor pressure upon sensing high drywell pres-sure and low-low reactor water level (in a one-out-of-two taken twice logic) and discharge pressure from either the core spray or the LPCI system. The APR system will not function if the FWCI system is functioning as determined by individual FWCI flow sensing and time delay networks that, if not satis-fied, give a permissive signal to the APR logic. Manual operation of the 4
APR system is also possible, separately from automatic operation.
Exces-sive reactor vessel pressure will also automatically open the pressure relief valves through action of the primary. pilot valves.
Discharge from each of the four valves in either the overpressure blowdown or the small break blowdown, is to the suppression pool.8 The depressurization of the reactor vessel by the APR system permits either the Core Spray system'or the LPCI system to cool _the reactor core following a small break loss of coolant accident.
6.2 Evaluation. The Millstone Unit 1 Technical Specifications require-a simulated automatic initiation once per operating cycle and, with the reactor at low pressure, a verification of manual valve operability. Ovec-lapping control circuitry for these two tests conform to Standard Review Plan 7.3, Appendix A.
The tests are not required during reactor operation since this would affect the operability of the unit. However, there is low probability that the system will fail to operate when required, whether or not testing is done during reactor operation.
7.0
SUMMARY
The review of the rr;erenced material has determined the following in regard to-the Millstone U..it 1 ECCS testing and testability:
1)
The core spray system does not meet all of the current criteria for testability during operation.
2)
The LPCI system does not meet all of the current criteria for testability during operation.
3)
The FWCI system meets the intent of the current criteria for testability during operation.
4)
The APR system meets all of the carrent criteria for testability during operation.
C0 REFERENCES 1.
Millstone Point Nuclear Power Station-Unit No.1, " Final 56fety Analy-sis Report" Amendment 5, dated March 14, 1968.
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2.
General Design Criterion 37, " Testing of Emergency Core Cooling Sys-tem," of Appendix A, " General Design Criteria for Nuclear Power Plants," 10 CFR Part 50, " Domestic Licensing of Production and Utili-zation Facilities."
3.
BTP ICSB 25, " Guidance for the Interpretation of GDC 37 for-Testing the Operability of the Emergency Core Cooling System as a Whole."
4.
Regulatory Guide 1.22, " Periodic Testing of the Protection System Actuation Functions."
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Nuclear Regula+.ory Commission Standard Review Plan, Section' 7.3,-
. Appendix A "Use of-IEEE Standard 279 in the Review of the ESFAS:and Instrumesitation and Controls of Essential Auxiliary Supporting
~ Systems."
. 6.
" Standard Technical Specifications for Westinghouse Pressurized Water Reactors-(PWRs)," NUREG-0452, Revision-3, Fall.1980.
I 7.
Technical Specifications and Bases for Millstone Nuclear Power Plant ^
Unit 1, Appendix A, to Provisional Operating License DPR-21,' Amend-ments i.through 45, dated December 1977.
L8.
Northeast Utilities letter, W. G. Counsil to Directcr of Nuc ear Reactor Regulation, NRC, "SEP Topic.VI-7.A 3, ECCS Actuation System,"
August 25, 1981,-A01825.
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