ML20091K681
| ML20091K681 | |
| Person / Time | |
|---|---|
| Site: | Monticello, Dresden, Pilgrim, Brunswick, Quad Cities, 05000000 |
| Issue date: | 11/10/1976 |
| From: | Baer R Office of Nuclear Reactor Regulation |
| To: | Schwencer A, Ziemann D Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML17139C189 | List:
|
| References | |
| FOIA-84-105 NUDOCS 8406070168 | |
| Download: ML20091K681 (1) | |
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UNITED STATES f
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MEMORANDUM FOR:
A. Schwencer, Chief Operating Reactors Branch-1, DOR D. Ziemann, Chief, Operating Reactors Branch-2, D0R FROM:
'R. Baer, Chief, Raactor Safety Branch, DOR
SUBJECT:
REVIEW OF SINGLE LOOP OPERATION IN BWR'S There are four cu'rrent technical assistance requests dealing with the review of the acceptability of operation of BWR's with a recirculation loop.
The requests and the plants are:
TACS Number Plant 2e 3 None Pilgrim-1 r
6170
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Brunswick-2 31j 6190 Monticello 16a 6184 Dresden 2/3 and 2.3 7,1" oI Quad Cities 1/2 154,t65 This memorandum is being written to discuss the status of these reviews.
Eac5 of these reviews consists of several major aspects; namely, ECCS performance, normal operation, and transients.
The Analysis Branch (AB) of DSS is reviewing the ECCS model and the Reactor Safety Branch (RS) of DOR is reviewing the plant specific aspects for both normal operation and ECCS performance.
The approach agreed upon by the General Electric Company, AB and RS was as follows:
(1) General Electric would submit a topical report documenting the i
ECCS model used for single loop BWR operation.
(2) The Reactor Safety Branch would review the information submitted on Pilgrim 1 and request any additional information required regarding methods of calculations for normal operation and transients.
(3) General Electric would update the Pilgrim 1 docket to reflect item (2).
l (4) The Pilgrim 1 docket would be refere ced for future submittals,
.'and responses to questions on non-ECCS calculational methods for
. lants under current review would also reference the Pilgrim 1 p
docket.
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NOV 111979
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ME'iOPR100M FOR:
T. Ippolito, Chief, Operating Reactors Branch #3, D0R FROM:
G. Lainas, Chief, Plant Systems Branch, D0R
SUBJECT:
BRUNSWICK STEAM ELECTRIC PLANT, UNIT 2 - SAFETY EVALUATION FOR N-1 LOOP OPERATION r
In response to technical assistance request, TAC 6171, enclosed is the Plant Systems Branch Safety Evaluation Report for single loop operation of Brunswick Unit 2.
We find the proposed modifications to the plant for single loop operation as described in Carolina Power and Light Company's submittal to be acceptable.
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G. Lainas, Chief Plant Systems Branch Division of Operating Reactors
Enclosure:
Safety Evaluation Report
Contact:
J. Burdoin X-28128 cc w/ enclosure:
D. Eisenhut n/
B. Grimes l
G. Zech W. Gamill P. Check F.~ Coffman S. Rubin J. Hannon G. Lainas R. Clark D. Tondi J. Burdoin P. Shemanski V. Panciera f J.L n - w - -= g - -
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' SAFETY EVALUATION REPORT N-1 LOOP OPERATION BRUNSWICK STEAM ELECTRIC PLANT UNIT 2 I.
, INTRODUCTION By letter.to the U.S. Nuclear Regulatory Comission (NRC) dated September 3,1976, the Carolina Power and Light Company (CP&L) submitted information to support its proposed license amendment to operate the Brunswick Steam Electric Plant, Unit 2, with one recirculation loop out of service (i.e., single-loop operation).
This infomation represented the licensee's analysis of significant events, based on a review of accidents and abnormal operational transients associated with power operations in the single-loop mode and provided by the nuclear steam supply steam designer (General Electric Company, Nuclear Energy Division (GE-NED)).
Conservative assumptions were employed in the GE-NED Report NED0-21281, dated May 1976, to ensure that its generic analyses for boiling water reactors (BWR) 3/4 were applicable to the Brunswick Steam Electric Plant, Unit 2.
GE-NED submitted an additional report (NED0-20566-2, dated July 1978) of an analytical model for a loss-of-Coolant Accident (LOCA) with one recirculation loop out-of-service which
, is pre:ently under review by the NRC Reactor Safety Branch (RSB).
1 The purpose of this report is to evaluate the E1cetrical, Instrumen-tation, and Centrol (EI&C) design aspects of the proposed license amendment as presented in NEDD-21281 using the following criteria:
IEEE Std-279-1979; the Code of Federal Regulations Title 10, Part 50.46; and Title 10 Put 50, Appendix A and Appendix K.
II.
EVALUATION The enclosed technical evaluation was prepared for us by Lawrence Livemore Laboratory /EG&G as part of our technical assistance program.
III., CONCLUSION The consultant has reviewed Carolina Power and Light Company's submittal for license amendment for single-loop operation of the Brunswick Steam Electric Plant, Unit 2, and concluded that the modifications satisfy the IEEE Std-279-1971 criteria and are acceptable.
i The submittal was based on the analysis in NEDO-21281 perfomed by l
the nuclear steam supply system manufacturer (GE-NED).
The manu-i facturer had, however, not analyzed the perfomance of the Emergency Core Cooling System (ECCS) during single-loop operating conditions.
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A new analysis has been performed by GE-NED for a LOCA with one recirculation loop out-of-service.
This analysis, reported in NEDO 20566-2, includes the ECCS single-loop analysis and was provided to satisfy the. Code of Federal Regulations, Title 10, Part 50, Appendix K.
The consultant also concluded that if an additional review of the EI&C design aspects is required as part of the staff's review of NED0-20566-2, the licensec will be required to update its submittal based on. that new analysis. Such a review, if required, will'be presented as a supplement to the consultant's technical evaluation.
Based on our review of consultant's technical evaluation, we conclude that conceptional design as presented in the licensee submittal and reviewed in the consultant's technical evaluation is acceptable.
However, the licensee's submittal did not include a design of hard-wire modifications (see Section 2.2 of attached technical evaluation) to the reactor protection system that will enable the operator to make setpoint changes frca the front of the nuclear instrument cabinet.
It is, therefore, concluded that before operation in the single-loop mode can be implemented at Brunswick, Unit 2, the licensee must accomplish the aforementioned modifications to the reactor protection system in a manner that satisfies IEEE Stds. 279-1971, 323-1971, and 344-1975.
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TECHNICAL EVALUATION OF THE l
ELECTRICAL, IhSTRUMENTATION, AND CONTROL DESIGN ASPECTS 0F THE PROPOSED LICENSE AMEN 0 MENT FOR SINGLE-LOOP OPERATION OF THE BRUNSWICK STEAM ELECTRIC PLANT, UNIT 2 9
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m, TABLE OF CONTENTS Page 1.
INTRODUCTION.........'..............
1 2.
DESCRIPTION AND EVALUATION OF THE PROPOSED LICENSE AMENDMENT FOR SINGLE-LOOP OPERATION...........'
3 2.1 Description of the Proposed Changes.......
3 2.2 Eval uation of the Proposed Changes........
4 3.
CONCLUSIONS.......................
7 REFERENCES...........................
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TECHNICAL EVALUATION OF THE ELECTRICAL, INSTRUMENTATION, AND CONTROL DESIGN ASPECTS OF THE PRG.0 SED LICENSE AMENDMENT FOR SINGLE-LOOP OPERATION OF THE BRUNSWICK STEAM ELECTRIC PLANT, UNIT 2 (Docket No. 50-324)
James H. Cooper EG&G, Inc., Energy Heasurements Group, San Ramon Operations 1.
INTRODUCTICN I
By letter to the U. S. Nuc' lear Regulatory Commission (NRC) dated September 3, 1976, the Carolina Power & Light Company (CP&L) submitted information to support its pr'oposed license amendment to operate the Brunswick steam electric plant, Unit 2, with one recirculation loop out of service (i.
e., single-loop operation).
This information represented the licensee's analysis of significant events, based on a review of accidents and abnormal operational transients associated with power operations in the si,ngle-loop mode and provided by the nuclear steam supply system designer (General Electric Company, Nuclear Energy Division (GE-NED)).
Conservative assumptions were employed in the GE-NED Report NED0-21;81,2 dated May' 1976, to ensure that its generic analyses for boiling water reactors (BWR) 3/4 were applicable to the Brunswick steam electric plant, Unit 2.
GE-NED submitted an additional report (NE00-20566-2,3 dated July 1978) of an analytical model for a los s-o f-cool ant accident (LOCA) with one recirculation loop out-of-service which is presently under review by the
'NRC Reactor Safety Branch (RSB).
O
The purpose of this report is to eval uate the electrical, instrienentation, and control (EI&C) design aspects of the proposed license 2
4
- amendment as presented in NED0-21281 and using IEEE Std-279-1971 criteria and the Code of Federal Reaulations, Title 10, Part 50.46,5 and Title 10, 6
7 Part 50, Appendix A and Appendix K criteria.
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DESCRIPTION AND EVALUATION OF THE PROPOSED LICENSE AMENCMENT FOR SINGLE-LOOP OPERATION l
2.1 DESCRIPTION
OF THE PROPOSED CHANGES 2
The' licensee states that from its analysis of NE00-21281 the only changes necessary to the reactor protection system (RPS) for single-loop' operation, are:
(1)
Mcdifications to the rod-block setpoints of the rod-block mqnitor (RBM) system.
(2)
Modifications to the SCRAM trip settings of the average power range monitor (APRM) system.
(3)
Reduction of 0.82 in the maximum average planar linear heat generation rate (MAPHLGR) limit for the fuel.
Because of the different flow quantity and different flow path during single-loop operation, the APRM SCRAM trip settings, which are flow-biased according to the equation in t'he technical specifications, require re-setting to protect the reactor from overpower.
The rod-block setpoint equation is flow-biased in the same way and with the same flow signal as the APRM setpoint, and must also be modified to provide adequate local core protection for the postulated rod withdrawal error.
The revised technical specifications propose si ngl e-l oo p operation at reduced safety settings for unlimited periods of time.
The revised technical specifications also pro ose a limit of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in which to reduce the safety settings.
Use of Section 3.4.1.1.a of the standard
'BWR ' technical specifications 8,g)) be required.
Section 3.4.1.1.a st'ates l
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"With one reci rcul ation loop not in operation, (reactor) operation may continue; restore both loops to operation. within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or b at least hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />."g in The nunerical values of the new sett.ings are delineated in the revised technical specifications which accompany the licensee's submittal.1 2.2 EVALUATION OF THE PROPOSED CHANGES The temporary changes in the settings of the trip points for the APRM and RBM must be made in the power-range cabinets in the control room and so must be done with the reactor shut down (1. e., with the mode switch in shutdown or refuel, condition 3, 4, 5) as required by the NRC Branch Technical Position ICSB. 12.9 These adjustments include readjusting the power an't flow potentianeters in each of the six APRM channels and the two RBM channels.
One channel of the multichannel systems will be adjusted at a time and then returned to service.
Before all of the channels are returned to serv ice, the new trip' setpoints will b verified by the instru:nent engineer following the readjustment and testing of the setpoints by the instrument technician.
Two operators will perform functional tests to double check the new setpoints end to check the instrunent's return to an operable condition.
The sequence outlined above shall be written into the plant technical specif'ications.
A permanent installation of the setpoint-change ca'pability must be made in order for the system to satisfy the requirements of Section 4.15 of IEEE Std-279-1971.4 Hard-wire modifications will be required to enable making setpoint changes from the front panel of the power rar.ge cabinet by way of control switches.
The licensee's proposed modifications must be submitted to the NRC staff for review prior to this installation.
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The recirculation-loop equalizer valves must be verified closed and tagged for single-loop power operation, as is the case for two-loop 2
power operation.
The safety analysis in NED0-21281 assumes that these valves remain closed as their effect on a LOCA has not been analyzed.
Recirculation flow must be manually controlled by the operator, as opposed to autanatic control, whenever the system is operating in the single-loop mode, since control stability is improved in the cranual mode.
Manual control is assumed in NEDO-21281.2 The technical specifications will be changed to include this restriction.
Due to the different flow pattern during single-loop operation as described by the licensee, a number of indications in the control rcom will change, such as individual jet-pump flow and total sumed core flow.
Some indications will be only slightly less accurate, but some others will be erroneous.
The control room indications must be corrected prior to single-loop power operation or they must be tagged out-of-service, as appropriate.
.This.is a requirement of Section 4.20 of IEEE Std-279-1971.4 The normal plant confi.guration as described in the final safety analysis report (FSAR)10 includes recirculation-pump start interlocks to prevent an inadvertent cold-water injection into the reactor.
Any recirculation loop that is out-of-service and whose water has cooled must be run in the bypass mode to preheat the water to within 100 F.of the reactor cooling ' water before the water may be valved back to the reactor pr' essure vessel.
The recirculation pump start is interlocked to permit start-up only if the pump discharge valve is closed, the bypass valve is open, and the suction valve is open.
This configuration will limit the amount of cold water which can be transported through the reactor vessel from a cold-loop startup, thereby limiting the effect of a cold-water slug event.
Although interlocks are provided, no credit is taken for their 2
safety functicn ~ in NED0-21281 for single-loop operation since this is not the limiting transient.
The instrument setpoints can be set down to enable operation in the single-loop mode for unrestricted periods.
This mode of operation is desired by the licensee to facilitate more ex'.ensive unschedul ed maintenance without the requirement of keeping the reactor shut down.
It is stipulated that single-loop opcration will, not be a planned mode of operation.
l The new rod-block and trip setpoints vary linearly as a function of recirculation flow rate.
For power increases by rod withdrawal, the RBM rod block must be set to the next higher trip level by manual operator
' action.
The APRM, flow-biased, SCRAM trip follows the new trip curve automatically for both power increases and decreases.
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CONCLUSIONS 4
We conclude that the Carolina Power & Light Company's license amendment submittal for single-loop operation of the Brunswick steam electric plant, Unit 2, satisfies the IEEE 5td-279-1971 criteria and is 2
acceptable.
- The submittal was based on the analysis in NE00-21281 perfonned by the nuclear steam supply system manufacturer (GE-NED).
The manufacturer had, however, not analyzed the performance of the emergency core cooling system (ECCS) during single-loop operating conditions.
A new analysis has been performed by GE-NED for a LOCA with one recirculation loop out-of-service.
This analysis, reported in NEDO 20566-2,3 includes the ECCS single-loop analysis and is in acc'ordance with the Code of Federal Reculations, Title 10, Part 50, Appendix K.7 If an additional review of the El&C design aspects is required as a result of NE00-20566-2,3 the licensee will be required to update its submittal based on that new analysis. The review will then be presented as a supplement to this technical evaluation.
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REFERENCES 1.
CP&L Letter to NRC (8. Rusche:), dated September 3,1976.
2.
General Electric Company, Nuclear Energy Division, Brunswick Steam Electric Plant Unit'2 License Amendment Submittal For
' Single-Loop O eration With the Bypass Flow Holes Plugged and With J
LPSI Modification, NED0-21281 (May 1976).
3.
General Electric Company, Nuclear Energy Division, An Analytical Model For loss-of-Coolant Accident (LOCA) With One Recirculation loop Out-Of-Service, NEDO-20566-2 (July 1978).
4.
IEEE Std-279-1971:
Criteria For Protection Systems For Nuclear Power Generating Stations (n. d.).
5.
Code of Federal Regulations, Title 10, Part 50.46:
Acceptance Criteria For Emergency Core Cooling Systems For Light Water Nuclear Power Reactors (January.1976).
6.
Code of Federal Regulations, Title 10, Part 50, Appendix A:
General Design Criteria For Nuclear Fower Plants (January 1978).
7.
Code of Federal Regulations, Title 10, Part 50, Appendix K:. ECCS Evaluation Models (January 1,1978).
8.
- Company, Standard Boiling Water Reactor Technical Specifications (n. d.).
9.
NRC/RS8, Protection System Trip Point changes for Operation With Reactor Coolant Pump Out of Service,. Branch Technical Position ICSB 12 (n. d.).
10.
- CPSL, Final Safety Analysis Report For Brunswick Stean E15ctric Plant (FSAR)(n.d.).
.