ML20009B582
| ML20009B582 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 07/09/1981 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Counsil W NORTHEAST NUCLEAR ENERGY CO. |
| References | |
| TASK-06-10.A, TASK-6-10.A, TASK-RR LSO5-81-07-033, LSO5-81-7-33, NUDOCS 8107160334 | |
| Download: ML20009B582 (18) | |
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Docket No. 50-245 LS05 07-033 N
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Mr. W. G. Counsil Vice President JUL15198 N 2-Nuclear Engineering and Operations
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Northeast Nuclear Energy Company Q
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Dear Mr. Counsil:
SUBJECT:
SEP TOPPC VI-10.A. TESTING OF REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES, INCLUDING RESPONSE TIE TESTING, DRAFT T SAFETY EVALUATION FOR MILLSTONE 1 is our contractor's draft technical evaluation for this topic. is the draft staff safety evaluation that is based upon Enclo-sure 1.
' proposes modifications to the Technical Specifications and some equipment to implement a response time testing program.
Your comments on Enclosures 1 and 2 are requested within 30 days. The need to actually implement these changes will be determined during the integrated safety assessment. This topic assessment 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 assessment is completed.
Sincerely, Dennis M. Crutchfield, Chief Operating Reactors Branch No. 5 Division of Licensing
Enclosures:
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July 9,1981 Docket No. 50-245 LS05 07-033 Mr. W. G. Counsil, Vice President Nuclear Engineering and Operations Northeast Nuclear Energy Company Post Office Box 270 Hartford, Connecticut 06101
Dear Mr. Counsil:
SUBJECT:
SEP TOPIC VI-10.A, TESTING 0F REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES, INCLUDING RESPONSE TIME TESTING, ORAFT T SAFETY EVALUATION FOR MILLSTONE 1 is our contractor's draft technical evaluation for this topic. is the draft staff safety evaluation that is based upon Enclo-sure 1. proposes modifications to the Technical Specifications and some equipment to implement a response time testing program.
Your comments on Enclosures 1 and 2 are requested within 30 days. The need to actually implement these changes will be determined during the integrated safety assessment. This topic assessment 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 assessment is completed.
Sincerely, ct I.
A 1V Dennis M. Crutchfield, Chief Operating Reactors Branch N3. 5 Division of Licensing
Enclosures:
As stated cc w/ enclosures:
See next page l
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MILLSTONE 1 Mr. W. G. Counsil Docket No. 50-245 cc William H. Cuddy, Esquire Connecticut Energy Agency Day, Berry & Howard ATTN: Assistant Director Counselors at Law Research and Policy One Constitution Plaza Development Hartford, Connecticut 06103 Department of Planning and Energy Policy Natural Resources Defense Council 20 Grand Street 91715th Street, N. W.
Hartford, Connecticut 06106 Washington, D. C.
20005 Northeast Nuclear Energy Company ATTN: Superintendent Millstone Plant P. O. Box 128 Waterford, Connecticut 06385 Mr. James R. Hiarelwright Northeast Utilities Service Coapany P. O. Box 270 Hartford, Connecticut 06101 Resident Inspector c/o U. S. NRC
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P. O. Box Drawer KK Niantic, Connecticut 06357 Waterford Public Library Rope Ferry Road, Route 156 Waterford, Connecticut 06385 First Selectman of the Town of Waterford Hall of Records 200 Boston Post Road Waterford, Connecticut 06385 John F. Opeka Systens Superintendent Northeast Utilities Service Company P. O. Box 270 Hartford, Connecticut 06101 U. S. Environmental Protection Agency.
Region I Office l
ATTN: EIS COORDINATOR JFK Federal Building Boston, Massachusetts 02203 e
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l SEP TECHNICAL EVALUATION 8
1 TOPIC VI-10.A f
TESTING 0F REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES i
MILLSTONE NUCLEAR POWER STATION, UNIT N0. 1 i
i Docket No. 50-245 f
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CONTENTS
1.0 INTRODUCTION
I 2.0 CRITERIA...........................................................
1 3.0 REACTOR TRIP SYSTEM................................................ 4 3.1 Description.................................................. 4 3.2 Evaluation.................................................... 5 4.0 STANDBY LIQUID CONTROL SYSTEM...................................... 8 4.1 Description................................................... 8 4.2 Evaluation.................................................... 8 5.0
SUMMARY
12
6.0 REFERENCES
12 TABLES 1.
Comparisons of Millstone Unit 1 RPS instrument surveillance requirements with BWR Standard Technical Specification requirements...................................................... 6 2.
Standby liquid control system and associated system surveillance requirements.....................................................
10 9
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SEP TECHNICAL EVALUATION TCPIC VI-10.A TESTING 0F REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES MILLSTONE NUCLEAR POWER STATION, UNIT NO. 1
1.0 INTRODUCTION
The objective of this review is to determine if all Reactor Trip System (RTS) components, including pumps and valves, are included in component and system tests, if the scope and frequency of periodic testing is adequate, and if the test program meets current licensing criteria. The review will also address these same matters with respect to the Standby Liquid Control System (SLCS) as a typical example of all Engineered Safety Feature (ESF) systems.
2.0 CRITERIA General Design Criterion 21 (GDC 21), " Protection System Reliability and Testability," states, in part, that:
The protection system shall be designed to permit periodic testing of its functioning when the reactor is in operation, including a capability to test channels independently to determine failure and losses of redundancy that may have occurred.
l Regulatory Guide 1.22, " Periodic Testing of the Protection System Actuation l
Functions," states, in Section D.l.a, that:
The periodic tests should duplicate, as closely as practicable, the performance that is required of the actuation devices in the event of an accident; and further, in Section 0.4, states that:
When actuated equipment is not tested during reactor operation, it should be shown that:
l 1
a.
There is no practicable system design that would permit operation of the actuated equipment without adversely affecting the safety or operability of the plant, b.
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, and c.
The actuated equipment can be routinely tested when the r2 actor is shut down.
IEEE Standard 338-1977, " Periodic Testing of Nuclear Power Generating Station Class lE Power and Protection Systems," states, in part, in Section 3:
Overlap testing consists of channel, train, or load-group verification by performing individual tests on the various components and subsystems of the channel, train, or load group. The individual component and subsystem tests shall check parts of adjacent subsystems, such that tne entire channel, train or load group will be verified by testing of individual components or subsystems.3 and in part in Section 6.3.4:
Response time testing shall be required only on safety systems or sub-l systems to veri'y that the response times are within the limits of the l
overall response times given in the Safety Analysis Report.
i Sufficient overlap shall be provided to verify overall system response.
The response-time test shall include as much of each safety system, from sensor input to actuated equipment, as is practicable in a single test. Where the entire set of equipment from sensor to actuated equip-ment cannot be tested at once, verification of system response time shall be accomplished by measuring the response times of discrete 2
O portions of the system and showing that the sum of the response times of all is within the limits of the overall system requirement.
In addition, the following criteria are applicable to the ESF: General Design Criterion 40 (GDC 40), " Testing of Containment Heat Removal System,"
states that:
The containment heat removal system shall be designed to permit appro-priate periodic pressure and functional testing to assure:
a.
The structural and leaktight integrity of its components, b.
The operability and performance of the active components of the system.
c.
The operability of the system as a whole and under conditions as close to the design as practical, the performance of the full operational sequence that brings the system into operatio1, including operation of applicable portions of the protection systems, the transfer between normal and emergency power sources, and the operation of the associated cooling water system.4 GDC 38, " Testing of Emergency Core Cooling ~ystens," CDC 43, " Testing of Containment Atmosphere Cleanup Systems and GDC 46, " Testing of Cooling Water System," are similar.
l Standard Review Plan, Section 7.3, Appendix A, "Use of IEEE Standard 279 in the Review of the ESFAS and Instrumentation and Controls of l
Essential Auxiliary Supporting Systems," states, in Section ll.b, that:
i Periodic testing should duplicate, as closely as practical, the inte-grated performance required from the ESFAS, ESF 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 " overlapping" tests do, in fact, overlap from one test segment to another. For example, 3
l
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.0 3.0 REACTOR TRIP SYSTEM (RTS) 3.1 Description. The system is made up of two independent logic channels, each having subchannels of tripping devices. Each subchannel has an input from at least one independent sensor, monitoring each of the crit-ical parameters.
The output of each pair of sabchannels is canbined in a one-out-of-two logic: that is, an input in either one or both of the independent subchan-nels will produce a logic channel trip. Both of the other two subchannels are likewise combined in a one-out-of-two logic, independent of the first logic channel. The outputs of the two logic channels are combined in two-out-of-two arrangement so that they must be in agreement to initiate a scram. An off-limit signal in one of the two subchannels in one of the logic channels must be confirmed by any other off-limit signal in one of the two subchannels of the remaining logic channel to provide a reactor scram.
During normal operation, all vital sensor and trip contracts are closed, and all sensor relays are operated energized. The control rod pilot scram valve solenoids are energized, and instrument air pressure is applied to all scram valves. When a trip point is reatned in any of the monitored parameters, a contact opens, de-energizing a relay which controls a contact in one of the two subchanne's. The opening of a subchannel con-tact de-energizes a scram relay which opens a contact'in the power supply to the pilot scram valve solenoids supplied by its logic channel. To this point, only one-half the events required to produce a reactor scram have occurred. Unless the pilot scram solenoids supplied by the other logic channel are de-energized, instrument air pressure will continue to act on the scram valves and operation can continue. Once a single channel trip is initiated, contacts in that scram relay circuit open and keep that circuit de-energized until the initiating ~ parameter has returned within operating 4
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limits and the reset switch is actuated manually.
It should be noted that each control rod has individual pilot scram solenoids for each channel and an individual air-operated scram valve. A normally-closed switch is pro-vided in each logic channel pilot scram solenoid circuit. This allows each rod to be manually scrammed (tested) by opening both logic channel switches and de-energizing the pilot scram solenoids. This type of test would pro-vide the required overlapping test of the RTS.
The parameters (sensors) which are required to initiate reactor scram are listed in Table 1.
However, the only instruments included in this table are those required to prevent exceeding the fuel claading integrity limits during normal operation or operational transients. These are described in Table VII-l of the plant FSAR and listed in Tables 4.1.1 and 4.1.2 of the f.illstone Nuclear Power Station Technical Specifications for Unit 1.
For example, the condenser low-vacuum sensors are connected to the RPS trip system and can initiate a -eram.
3.2 Evaluation. The Millstone 1 RTS is designed to allow overlap-ping tests from actuating device through the control rods. The design allows individual channel tests from sensors though pilot scram valves while the reactor is in operation and the overlapping rod scram tests during refueling. Although one or more rod scram valves may fail during reactor operation, the channel tests will verify that no common mode fail-ure will occur and sufficient pilot valves will operate to shut down the reactor.
Table I shows the present Millstone 1 RTS instrument surveillance requirements, including frequency. The table also shows the current licen-sing requirements for General Electric boiling water reactors as listed in the Standard Technical Specifications. The tests shown only involve single l
channels testing (half-scram).
It should be noted that Technical Specification Table 4.1.2 does not require channel calibration for main steam-line isolation valve closure or turbine stop valve closure parameters, although the Millstone Technical Specification requirement for Unit 1 in Section 2.1.2.8 requires that a 10%
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REQUIREMENTSgF MILLSTONE UNIT 1 RPS INSTRUMENT SURVEILLAN CDMPARISONS TABLE 1.
WITH,8WR STANDARD TECHNICAL SPECIFICATION REQUIREMENTS (STS)0 Channel Channe}
Functiogal Channel C
Check Test Calibration Millstone Millstone Millstone Instrument Channel Unit 1 STS Unit 1 STS Unit i STS High reactor pressure NA NA Q*
M Q
R High drywell pressure NA NA Q*
M Q
Q Low reactor water D
D Q*
M Q
R level High water level in
- A NA Q*
M Q
R scram discharge Condenser low vacuum NA NA Q*d NA R
NA Main steam-line iso-NA NA Q*
M NA R
lation valve closure Turbine stop valves NA NA Q*
M NA R
closure Manual scram NA NA Q*-
M NA NA Turbine control valve NA NA Q*
M NA Q
fast closure Average power range NA S
Q sue Q
W/SA monitor (APRM) flow biased high flux APRM-reduced high flux NA S
Sue 500 Q
W/SA Intermediate range NA S
sue
-Sue R
R monitor (IRM)
High steam line S
W Q*
W Q
R radiation Reactor mode switch NA NA R
R NA NA in shutdown position 6
O TABLE 1.
(continued)
FREQUENCY NOTATION Notation Frequency lotation Frequency S
At least once per R
At least once per refueling 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> outage (18 months)
D At least once per NA Not applicable 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> SA At least once per 184 days W
At least once per SU Prior to start up 7 days M
At least once per SD Prior to shutdown 31 days Q
At least once per Q*
Baseo on unsafe failure rate 3 months data and reliability analysis.
Not less than one-month or greater than three months, a.
A qualitative determination of acceptable operability by observation of channel behavior during operation. This determination shall include, where possible, comparison of the channel with other independent channels measuring the same variable.
b.
Injection of a simulated signal into the channel to verify its proper response including, where applicable, alarm and/or trip initiating action.
c.
Adjustecit of channel output such that it responds, with acceptable range and accuracy, to known values of the parameter which the channel measures. Calibration shall encompass the entire channel, including equip-ment actuation, alarm, or trip.
d.
Consists of injecting a simulated electrical signal into the measurement channel.
e.
Maximum test frequency is once per week.
valve closure initiate scram. Additionally the time delay of 260 msec for the Turbine Control Valve Fast Closure is not verified.
The Standard Technical Specifications fcr General Electric boiling water reactors (page 3/4 3-1, paragraph 4.3.1.2) require the logic system 7
function test and siriulated automatic operation at least every 18 months.
Available information indicates that the overlapping system test is not performed at Millstone, Unit 1.
As can be seen in Table 1 the following channels are not subjected to a channel check as frequently as required for present-day licensing:
APRM--Flow biased ".igh flux APRM--Reduced high Flux IRM The following channels ace not subjected to a channel functional test at frequently as required for present-day licensing:
High Reactor Pressure High Drywell Pressure Low Reactor Water Level High Water level in scram discharge Main Steam Line Isolation Valve Closure Turbine Stop Valves Closure Manual Scram Turbine Control Valves Fast Closure APRM--Flow biased high flux High Steam Line Radiation The following channela are not calibrated at least as frequently as required for present-dry licensing:
APRM--Flow biased high flux APRM--Reduced high flux Main Steam Line isolation valve closure Turbine Control Valve Fast Closure Turbine Stop Valves Closure In Section 3.1 of the Millstone 1 Technical Specifications, 100 milliseconds is stated as the required limit to the response time
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between any channel trip and the de-energization of the scram s'lenoid o
relay. Response time testing to verify that the channel response time does not exceed this requirement,is not in evidence in the Technical Specifications.
4.0 STANDBY LIQUID CONTROL SYSTEM 4.1 Description. The standby liquid control sytem is designed to insert a sodium pentaborate (or equivalent poison) solution to render and maintain the reactor subcritical even when the control rods are all fully withdrawn. The equipment consists of an unpressurized solution storage tank, a pair of positive displacement pumps, either of which has full capacity to perform the system function, two explosive actuated shear plug valves, a poison sparger ring and associated valves, piping and instrumen-tation. A complete description is in Section VI-7.2 of the plant FSAR.
The storage tank is heated to prevent particulate formation. The discharge cf each pump is protected by a pressure relief valve that discharges back to the storage tank. Pilot light indication of circuit continuity for the explosive valves is provided. A single key controlled switch will start a pump and open associated valves. Both sets of valves and pumps are not operated simultaneously, however, the valves for both pumps may be open. A test tank and a supply of demineralized water are provided for testing.
The FSAR indicates that testing is done in two parts. One part determines the ability of the pump to develop flow and suction from the storage tank. The system is afterwards flushed to prevent boron precipitation. Another test uses demineralized water to show that water can be delivered into the reactor vessel. This test requires replacement of the explosive charges in the shear plug valves.
4.2 Evaluation. Table 2 shows the current testing requirements for the standby liquid control system and associated systems. The following surveillence is not done at least as frequently as required for present day licensing:
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TABLE 2.
STANDBY LIQUID CONTROL SYSTEM SURVEILLANCE REQUIREMENTS Frequency Millstone Surveillance Regt f rements Unit 1 STS 1.
Solution temperature within limits.
Da D
2.
Solution Volume is greater than specified.
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3.
Heat traced pump suction piping is greater than N/A 0
or equal to 700F.
4.
Start both pumps and raci.
' ate demineralized Mc g
water to the test tank.
5.
Verify the continuity cf the explosive charges.
M 6.
Solution chemical analysis.
M/M M/M M
7.
Verify valve position and that they are not locked, sealed or otherwise secured.
8.
Initiating one loop using demineralized water R
R and replacement of the explosive charga.
9.
Verify minimum flow requirement against reactor M
R vessel head pressure.
- 10. Demonstrate relief valve setpoint and that it Rd R
does not operate during recirculation test to the test tank.
R/M
- 11. Verify piping from the storage tank to the reactor vessel is not blocked.
R
- 12. Demonstrate that the storage tank heaters are operable.
a.
Minimum temperature is not specified.
b.
Minimum volume is not specified.
c.
Flow rate required to be 32 gpm while the FSAR design requires 40 gpm.
The technical specifications do not require testing of both pump loops.
Pressure not specific. A second requirement 4.4.A.2b recirculates solution from and to the storage tank at least once in 18 months for both systems.
d.
Non-operation during recirculation test is not required.
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Verification of the continuity of the explosive charges.
Valve position and that they are not locked, sealed or otherwise secured.
Verification that pump suction piping is not blocked.
Demonstration that the storaga tank heaters are operable.
Millstone 1 does not have heat traced piping in the Standby Liquid Control System, therefore this requirement is not applicable.
The Millstone 1 technical specifications do not agree with the present standard technical specifications further in that:
1.
The minimum volume of solution is not specified, 2.
The mininum solution temperature is not specified, 3.
The relief valves are not verified to not operate under normal system operating pressure, and 4.
Both pump loops are not specifically tested monthly (Item 4).
One loop could be tested all the time while the other loop is not tested.
Further, it is apparent that Millstone 1 has only one heater in the solution storage tank, whereas present requirements are for two.
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5.0
SUMMARY
The Technical Specifications for Millstone Unit I were compared with the Standard Technical Specifications for current Boiling Water Reactor licensing.
It was found that, for the reactor trip system, three signals are not subjected to a channel check, ten signals are not subjected to a channel functional test and five channels are not calibrated as frequently as required in the standard technical specifications.
(See Section 3.2.)
Additionally, the channel respor.se time between channel trip and the de-energization of the scram relay is not required to be tested.
For the Standby Liquid Control System, selected as typical of ESF systems, surveillance requirements were less frequent (or non-existent) than required in the standard technical specificati0ns in four requirements.
Four additional requirements do not conform with the standard technical specification while the frequency of surveillance does.
(See Section 4.2.)
6.0 REFERENCES
1.
General Design Criterion 21, " Protection System Reliability and Test-ability," of Appendix A, " General Design Criteria of Nuclear Power Plants," 10 CFR Part 50, " Domestic Licensing of Production and Utili-zation Facilities."
2.
Regulatory Guide 1.22, " Periodic Testing of the Protection System Actuation Functions."
3.
IEEE Standard 338-1975, " Periodic Testing of Nuclear Power Generating Station Class 1E Power and Protection Systems."
4.
General Design Criterion 40, " Testing of Containment Heat Removal Systems," of Appendix A, " General Design Criteria of Nuclear Power Plants," 10 CFR Part 50, " Domestic Licensing of Production and Utili-zation Facilities."
5.
Nuclear Regulatory Commission 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."
6.
Standard Technical Specifications for General Electric Boiling Water Reactors (BWRs), NUREG-0123, Ravision 2, Fall 1980.
12
7.
Millstone Point Nuclear Power Station-Unit No.1, "Tinal Safety Analysis Report," Amendment 5, dated March la, 1968.
8.
Technical Specifications.and Basec, for Millstone Nuclear Power Plant Unit 1, Appendix A, to Provisional Operating License DPR-21, Amendments 1 through 45. dated December 1977.
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e TOPIC: VI-10.A. TESTING OF REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES, INCLUDING RESPONSE TIME TESTING I.
INTRODUCTION The purpose of this Topic is to review the reactor trip system (RTS) and engineered safety features (ESF) test program for verification of RTS and ESF operability on a periodic basis and to verify RTS and ESF response time in order to assure the operability of the RTS and ESF.
Response times should not exceed those assumed in the plant accident analyses. Accordingly, the test program of the RTS and ESF was reviewed in accordance with the Standard Review Plan, including applicable Branch Technical Positions.
II.
REVIEW CRITERIA The review criteria are presented in Section 2 of EG&G Report 0400J,
" Testing of Reactor Trip System and Engineered Safety Features."
III. RELATED SAFETY TOPICS AND INTERFACES Topic VI-7.A.3 discusses the question of testing protection systems under conditions as close to design condition as practical. There are no topics that are dependent on the present topic information for their completion.
IV.
REVIEW GUIDELINES Review guidelines are presented in Section 2 of Report 0400J.
V.
EVALUATION Millstone 1 does not comply with the current licensing criteria, because the systems required to protect the public health and safety are not tested as frequently as operating experience has indicated to be desirable and because response time testing is not conducted.
VI.
CONCLUSION It is the staff's position that the design of systems which are required for safety shall include provisir for periodic verification that the minimum perf)rmance of instruments and control is not less than that which was assumed in the safety analyses. The bases for this position are Gen-eral Design Criterion 21, Section 3.9 of IEEE Std 279-1971, and IEEE Std 338-19i/. Therefore, the licensee should implement a program for response time testing of all reactor protection system (including engineered safe-ty features systems such as containment isolation).
Furthermore, the present Technical Specifications should be revised to reflect the higher test frequency of the current Standard Technical Specifications.
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