ML20008G090
| ML20008G090 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 06/30/1981 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Counsil W NORTHEAST NUCLEAR ENERGY CO. |
| References | |
| TASK-06-07.A3, TASK-6-7.A3, TASK-RR LSO5-81-06-128, LSO5-81-6-128, NUDOCS 8107020218 | |
| Download: ML20008G090 (15) | |
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Mr. W. G. Counsil. Vice President j JUL 0119816 h Nuclear Engineeriny and Operations V'
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Dear Mr. Counsil:
SUBJECT:
SEP TOPIC VI-7.A.3 ECCS ACTUATION SYSTEM DRAFT SER FOR l
MILLSTONE NUCLEAR POWER STATION UNIT 1 l
I is a draft safety evaluatd6n report that is based on Enclosure 2.
This safety evalostion presents the staff position regarding the de-l sign of your facility in this area. is a copy of our contractor's draft evaluation of Systematic Evaluation Program Topic VI-7.A.3.
This assessment compares your facility, as described in Docket No. 50-245, with the criteria currently used by the regulatory staff for licensing new facilities. Please infonn us if your as-built facility differs from the licensing basis assumed in our assess-I ment within 30 days of receipt of this letter.
In addition to correcting any errors of fact that may exist in our evalua-tions you are reouested to provide your justification for not conducting tests of the core spray system and LPCI during reactor operations using the full flow return lines.
l This evaluation will be a basic input to the staff's safety evaluation re-port for this topic for your facility unless you identify changes needed to reflect the as-built conditions at your facility. This topic assess-ment may be revised in the future if your facility design is changed or if NRC criteria relating to this topic are modified before the integrated as-sessment is completed.
Sincerely, Dennis M. Crutchfield, Chief Operating Reactors Branch No. 5 8107020 W Division of Licensing AD:SA:DL*
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Docket No. 50-245 Mr. W. G. Counsil, Vice President Nuclear Engineering and Operatic's dortheast Nuclear Energy Conpany Post Office Box 270 liartford, Connecticut 06101
Dear Mr. Counsil:
SUBJECT:
SEP TOPIC VI-7.A.3 ECCS ACTUATION SYSTEM DR MT SER FOR MILLSTOL NUCLEAR POWER STATION UNIT 1 is a draft safety evaluation report that is based on Enclosure 2.
This safety evaluation presents the staff pcsition regarding the de-sign of your facility in this area. is a copy of our contractor's draft evaluation of Systematic Evaluation Program Topic VI-7.A.3.
This assessment compares your facility, as described in Docket No 50-245, with the criteria currently used by the regulatory staff for licensing new facilities. Please inform us if your as-built facility differs from the licensing basis assumed in our assess--
ment within 30 days of receipt of this letter.
In addition to correcting any errors a fact that may exist in our evalua-tions you are requested to provide your justification for not conducting tests of the core spray system and LPCI during reactor operations using the full flow return lines.
This evaluation will be a basic input to the staff's safety evaluation re-port for this topic for your facility unless you identify changes needed to reflect the as-built conditions at your facility. This topic assess-ment may be revised in the future if your faility design is changed or if NRC criteria relating to this topic are modified before the integrated as-sessment is conpleted.
Sincerely, Dennis M. Crutchfield.. Chief
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Operating Reactors Branch No. 5 Division of Licensing
Enclosures:
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June 30, 1981 Docket No. 245 LS05-81-06-128 L
l Mr. W. G. Counsil, Vic.e President Nuclear Engineering and Operations Northeast Nuclear Energy Company Post Office Box 270 Hartford, Connecticut 06101 j
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Dear Mr. Counsil:
SUBJECT:
SEP TOPIC VI-7.A.3 ECCS ACTUATION SYSTEM DRAFT SER FOR MILLSTONE NUCLEAR POWER STATION UNIT 1 l is a draft safety evaluation report that is based on Enclosure 2.
This safety evaluation presents the staff position regarding the de-l sign of your facility in this area. is a copy of our contractor's draft evaluation of Systematic Evaluation Program Topic VI-7.A.3.
This assessment compares your facility, as described in Docket No. 50-245, with the criteria currently used by the regulatory staff for licensing new facilities. Please infonn us if your as-built facility differs from the licensing basis assumed in our assess-ment withi'n 30 days of receipt of this letter.
In addition to correcting /
)rs of fact that may exist in our evalua-tions you are requested tr.
.me your justification for not conducting tests of the core spray system and LPCI during reactor operations using l
the full flow return lines.
This evaluation will be a basic input to the staff's safety evaluation re-port for this topic for your facility unless you identify changes needed to reflect the as-built conditions at your facility. This topic assess-ment may be revised in the future if your facility design is changed or if NRC criteria relating to this topic are modified before the integrated as-sessment is completed.
l Si cerely, L2/Y Dennis M. Crutchfield, f
Operating Reactors Branc No. 5 Division of Licensing
Enclosures:
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TOPIC:
VI-.7.A.3 ECCS ACTUATION _SYSTE!i I.
INTRODUCTION The ECCS actuation system was reviewed with respect to the i.estability of operability and performance of individual active components of the syst,am and of the entire system as a whole under conditions as close to tae design condition as practical. The purpose of the reviews was to assure that all ECCS components (e.g. valves and pumps) are included l
in the component and system test and to assure that the scope of the periodic testing is adequate and meets the requirements of GDC 37. The l
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technical specifications were also audited for large differences be-s tween the present test requirements and those in the Standard Technical Specifications.
II.
RE'.'IEW CRITERIA The current licensing criteria are identified in Section 2 of EG&G Re-port 0432J "ECCS Actuation System".
III. Related Safety Tonics an'd Interfaces The scope of review for this topic was limited to avoid duplication of effort since some aspects of the review wera perfomed under related topics. Related topics and the subject mv..' are identified below.
Each of the related ' topic reports contain tne acceptance cr f teria and review guidance for its subject matter.
Topic VI-3, " Containment Pressure and Heat Removal Capability."
Topic VI-4, " Containment I.nlation System."
Topic VI-7, " Emergency Core Cooling syn.em."
Topic VI-7.C. "ECCS Single Failure ' Criterion ar.d Requirements for t.ocking, Out Power to Valves Including Independence of Interlocks on ECCS Valves."
Topic VI-9, " Main Steam Isolation."
Topic VI-10.A, " Testing of Reactor Trip System and Engineered Safety Features Including Response Time Testing."
Only Topic VI-10.A. is dependsnt on the present topic information for completion. Response time testing is addressed in Topic VI-10.A.
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IV. REVIEW GUIDELINES The review guidelines are presented in Sections 3, 4, 5, and 6 of Report 0432J.
V.
EVALUATION Report 0432J describes the extent to which the ECCS actuation system can be tested, except for the question of response time testing. The report notes that the testing of the Core Spray and LPCI systens does not satis-fy Regulatory Guide 1.22 nor General Design Criterion 37 because full flow tests from a simulated actuation signal are not required.
The report also notes that the Technical Specifications d'o not spe:ify that each of the four LPCI pump combinations must be capable of deliv-ering 15,000 gpm.
It is not the function of the Technical Specifications to provide operating procedures nor is it the intent of the Technical Specifications to require that any three pumps be able to deliver 15,000 gpm when four pumps are available (because the LPCI is redundant to the Core Spray).
VI. CONCLUSION Based upon our review of our contractor's evaluation, the staff concludes
, that the Millstone 1 plant does not conform to current licensing criteria because system level fall flow testing of the Core Spray and LPCI is not conducted. A suitable change in the Technical Specificaticas should be developed.
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e 0432J SEP TECHNICAL EVALUATION TOPIC VI-7.A.3 ECCS ACTUATION SYSTEM MIL. STONE NUCLEAR POWER STATION, UNIT NO. I Docket No. 50-245 June 1981 A. C. Udy O
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CONTENTS 1
1.0 INTRODUCTION
I 2.0 CRITERIA..........................................................
3 3.0 CORE SPRAY SYSTEM.................................................
3 3.1 Description..................................................
3 3.2 Evaluation...................................................
4.0 LOW PRESSURE COOLANT INJECTION SYSTEM.............................
5 5
4.1 Description..................................................
5 4.2 Evaluation...................................................
6 5.0 FEEDWATER COOLANT INJECTION SYSTEM................................
6 5.1 Description..................................................
7 5.2 Evaluation...................................................
6.0 AUTOMATIC PRESSURE RELIEF SYSTEM..................................
7 7
6.1 Description..................................................
8 6.2 Evaluation...................................................
8 7.0
SUMMARY
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8.0 REFERENCES
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1.0 INTRODUCTION
The objective of this review is to determine if all Emergency Core Cooling System (ECCS) components, including pumps and valves, are included in component and system tests, if the scope and frequency of periodic test-ing are identified, and if the. test program meets current licensing crite-ria. The systems included in the ECCS are t'he Core Spray system, the Low Pressure Coolant Injection (LPCI) system, the Feedwater Coolant Injection, (FWCI) system and the Automatic Pressure Relief (APR) system.I 2.0 CP.ITERIA l
General Design Criterion 37 (GDC 37), " Testing of Emergency Core Cool-ing Systems," requires that:
l The ECCS be designed to permit appropriate periodic pressure and func-tional testing to assure the operability of the system as a whole and l
to verify, under conditions as close to design as practical, the per-l formance of the full operational sequence that brings the system into l
operation, including operation of applicable portions of the protec-tionsystem,thetransferbetweennormalandemergencypogersources, and the operation of the associated cooling water system.
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Branch Technical Position ICSB 25, Guidance for the Interpretation 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 periadic 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 cf the ESFAS and Instrumentation and Controls of Essential Auxi-liary Supporting Systems," states, in Section 11.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 during power operation must be reviewed in detail. Check that "overlappinf 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 got be adequate to l
test the ability of the ESFAS to close the breaker.
Regulatory Guide 1.22 states, in Section 0.4, that:
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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
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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 reactor operation Theactuatedgquipmentcanberoutinelytestedwhenthereactor 3.
is shut down.
3.0 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 from the suppression pool (the condensat,e 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 contrcl logic.
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Each subsystem has two independent control logic channels. Four low-low reactor water level and four diywell high pressure sensors, con-nected in a one-out-of-two taken twice logic, v:lil initiate the logic Each of the two logic channels will complete channels in both subsystems.
The system the equipment control functions for its respective subsystem.
selector switch permits manual testing of each logic channel without the normal interaction (start) of the other core spray system with the reactors 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 3
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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.
7 The. Millstone Unit 1 Technical Specifications require the following tests ano surveillance for the ccre spray system:
a.
Simulated Automatic Actuation Test--each refueling outage b.
Pump Flow Rate--after pump maintenance and every three months c.
Pump Operability--once per month d.
Motor oper?'ed talves--once per month e.
Core spray header op instrumentation--check, once per day, calibrate and test once per three months.
The spray pump space coolers, part of the Turbine Building secondary closed cooling water system (cooled by the service water system), are not required to be tested by the uait technical specification,
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l The Core Spray System is not tested from the automatic actuatf6n devices throuch to the establishment of flow through to the test bypass valve during reactor operation. The FSAR and the Technical Specifications do not establish that the test would adversly affect the safety or the operability of the unit, nor has the licensee established that the proba-bility of core spray failure is acceptably low without regular testing during nuclear operation. Thus, the required testing for the core spray system does not conform with the testing requirements of Regulatory Guide 1.22 and GDC 37.
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4.0 Low Pressure Coolant Infection System 4.1 Descrig 6. The Low Pressure Coolant Infection (LPCI) system consists of pumps, piping, valves and instrumentation to infect coolant from the suppression pool into the reactor vessel. There 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 exhanger 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 tljat used for the Core Spray system using the same primary detectors initiate each LPCI loop. Admission valves are opened, after the LPCI pressure is established, to those recir-b culation loops that do not have an indicated pipe rupture.
i 4.2 Evaluation. 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, als'o not in compliance with
-Regulatory Guide 1.22 and GCD ~7.
Additionally, the technical specifica-tions only require that (any) three LPCI pumps be able, together, to estab-lish a flow rate of 15,000 gallons per minute. It does not establish that each set of three LPCI pumps meet this criteria. Should a pump fail, the l
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testing does not verify that any three remaining pumps have the capacity I
required to mitigate the consequencses of an accident.
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 maintanence 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 FEEDWATER 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 i
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.
l The operating conoitions for the FWCI system components during a loss of coolant accident are the same as they are normally exposed to as the l
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 i
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and two reactor feed pumps are part of the FWCI system.
i 5.2 Evaluation.
In addition to the normal operation of the FWCI system components as part of the feedwater system, the FWCI condensate transfer system is tested, per technical specification requirements after pump maintenance and every three months o verify flow capability at a system head. 'A simulated automatic actestion 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 FWCI components are utili. zed 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) 7 rei.
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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 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 three valves in either the overpressure' blowdown or the small break blowdown, is to the suppression pool.
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 autornatic initiation once per operating cycle and with the reactor at low pressure, a verifi<ation of manual valve operability. These two tests overlap control circuitry that 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 referenced material has determined the following in reguard to the Millstone Unit 1 ECCS testing and testability:
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The core spray system does not meet all of the current require-ments for testability during operation.
2)
The LPCI system does not meet all of the current requirements for testability during operation.
3)
The FWCI system meets.the intent of the current requirements for testability during operation.
4)
The APR system meets all of the current requirements for test-ability during operation.
8.0 REFERENCES
1.
Millstone Point Nuclear Power Station-Unit No. 1, " Final Safety Analy-sis Report" Amendment 5, dated March 14, 1968.
2.
General Design Criterion 37, " Testing of Emergency Core Cooling Sys-tem," of Appendix A, " General Design Criteria of Nuclear Power Plants,"
10 CFR Part 50, " Domestic Licensing of Production and Utilization Facilities."
l 3.
BTP ICS8 25, " Guidance for the Interpretation of GDC 37 for Testing the Operability of the Emergency Core Cooling System as a Whole."
l 4.
Regulatory Guide 1.22, " Periodic Testing of the Protection System Actuation Functions."
5.
Nuclear Regulatory Commission Standard Review Plan, Section 7.3, Appendix A, "Use of IEEE Standard 279 in the Review of the ESFAS and 9
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i Instrumentation and Controls,of Essential Auxiliary Supporting Systems."
6.
Standard Technical Specifications for General Electric Boiling Water Reactors (BWRs), NUREG-0123, Revision 2, Fall 1980.
7.
Technical Specifications and Bases for Millstone Nuclear Power Plant Unit 1, Appendix A, to Provisional Operating License DPR.21, Amend-ments 1 through 45, dated December 1977.
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