ML20090A687
| ML20090A687 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Brunswick |
| Issue date: | 04/30/1980 |
| From: | Cooper J EG&G, INC. |
| To: | |
| Shared Package | |
| ML17139C189 | List:
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| References | |
| FOIA-84-105 EGG-1183-4144, NUDOCS 8006260153 | |
| Download: ML20090A687 (13) | |
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,1183-4144 I
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s TECHNICAL EVALUATION OF THE ELECTRICAL, INSTRUMENTATION, AND CONTROL DESIGN ASPECTS OF THE PROPOSED LICENSE AMENDMENT FOR SINGLE-LOOP OPERATION OF THE BRUNSWI.CK STEAM ELECTRIC PLANT, UNIT 2
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Work Perfcrmed for Lawrence Livermore Laboratory under U.S. Department of Energy Contract No. DE-ACO8-76 NVO 1103.
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1183-4144 Energy Measurements Group San Ramon Operations TECHNICAL EVALUATION OF'THE ELECTRICAL, INSTRUMENTATION, AND CONTROL DESIGN ASPECTS OFTHE PROPOSED LICENSE AMINDMENT FOR SINGLE-LOOP OPERATION OFTHE BRUNSWICK STEAM ELECTRIC PLANT, UNIT 2 (DOCKET NO. 50-324) by James H. Cooper Approved for Publication 1//L=
/ Jolin' R.' Radosevic
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Department Manager l
This Document is UNCLASSIFIED j
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' Nicholas 'E.~Broderick l
Department Manager l
Derivative Classifier Work Performed for Lawrence Livermore Laboratory under U.S. Department of Energy Contract No. DE-ACOS-76 NVO 1183.
SAN RAnr. OPERATIONS 3001 C' D r 'CW C#.NvCN 8tC AC
l ABSTRACT This report documents the technical evaluation of the electrical, instrumentation, and control design aspects of the proposed license amend-ment for single-loop operation for the Brunswick steam electric plant, Unit 2.
The review criteria are based on IEEE Std-279-1971 requirements and the Code of Federal Regulations, Title 10, Part 50.46 and Part 50, Appendices A and X requirements for single-loop operation.
This report is supplied as part of the Selected Electrical, Instrtmentation, and Control Systems Issues Program being conducted for the U. S. tiuclear Regulatory Ccmission by Lawrence Livermore Laboratory.
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FOREWORD This report is supplied as part of the Selected El ectrical,
Instrumentation, and Control Systems Issues (SEICSI). Program being conduct-ed for the U. S. Nuclear Regulatory Ccmmission, Office of Nuclear Reactor Regulation, Division of Operating Reactors, by Lawrence Livermore Labora-tory, Field Test Systems Division of the Electronics Engineering Depart-ment.
The U. S. Nuclear Regulatory Commission funded the work under the authorization entitled " Electrical, Instrumentation and Control System
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Support," B&R 20 19 04 031, FIN A-0231.
This work was perfomed by EGaG, Inc., Energy Measurements Group, S'an Ramon Operations, for Lawrence Livermore Laboratory under U. S-Depart-ment of Energy contract number DE-AC08-76NV01183.
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TABLE OF CONTENTS
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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 Evaluation of the Proposed Changes........
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3.
CONCLUSIONS.......................
- 7 3.1 I ntrod uct ion...................
7 3.2 Co nc l us i o n s...................
7 REFERENCES............................
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TECHNICAL EVALUATION OF THE ELECTRICAL, INSTRUMENTATION,. AND CONTROL DESIGN ASPECTS OF THE PROPOSED LICENSE AMENDMENT FOR SINGLE-LOOP OPERATION OF THE BRUNSWICK STE#1 ELECTRIC PLANT, UNIT 2 (Docket No. 50-324)
James H. Cooper EG&G, Inc., Energy Measurements Group, San Ramon Operations 1.
INTRODUCTION 1
By 1.etter to the U. S. Nuclear Regulatory Ccamis'sion (NRC) dated September 3, 1976, the Carolina Power & Light Ccmpany (CP&L) submitted information to support its proposed license amendment to cperate the Bruns-wick steam electric pl ant, 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 abnonnal operational transients associated with power operations in the single-loop mode and provided by the nuclear steam supply system designer (General Electric Company, Nuclear Energy Division (GE-NED)).
Conservative 4
assumptions were employed in the GE-NED Report NED0-21281,2 dated May 1976, to ensure that its generic analyses for boiling water reactors (BWR 3 and/or 4) were applicable to the Brunswick steam electric plant, Unit 2.
GE-NED submitted an additional report (NED0-20565-2,3 dated July 1978) of an analytical model for a loss-of-coolant accident (LOCA) with one recirculation loop out of service which is presently under review by the NRC Reactor Safety Branch (RSB).
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-The purpose of this report is to evaluate the electrical, instru-t tion, and control (EISC) design aspects of the proposed license amend-men a 2
4 ment 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, Part 50, Appendix A and Appendix X# criteria.
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DESCRIPTION AND EVALUATION OF THE PROPOSED LICENSE AMENDMENT FOR SINGLE-LOOP OPERATION 2'.1 DESCRIFTION OF THE PROPOSED CHANGES 2
The licensee states that from its analysis of NED0-21281 the only changes necessary to the reactor protec. tion system (RPS) for single-loop operatio'n, are:
(1)
Modifications to the rod-block setpoints of the rod-block monitor (RBM) system.
(2)
Modifications to the SCRAM trip settings of the average power range monitor (APRM) system.
(3)
P. eduction of 0.82 in the maximum average planar
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linear heat generation rate (FAPHLGR) 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 the technical specifications, require reset-ting to protect the reactor from overpower.
The rod-block setpoint equa-tion 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 single-loop opera-tion at reduced safety settings for unlimited periods of time.
The revised technical specifications also propose 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 EWR 8
technical specifications will be required.
Section 3.4.1.1.a states that "With one recirculation loop not in operation, (rcac-tor) 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 down 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 at.least hot shut-e h
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The numerical values of the new settings are delineated in the revised technical specifications which accompany the licensee's submittal.1 5.2 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 shet' down (i. e., with the mcde 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 and flow potentiometers 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 re-turned to service, the new trip setpoints will be verified by the instru-ment 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 and to check the instrument's return to an operable condition.
The sequence outlined above shall be written into the pl ant technical specifications.
A pemanent installation of the setpoint-change capability 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 range esbinet by way of control switches.
The licensee's proposed modifications must be submitted to the NRC staff for review prior to this installation.
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.
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f Recirculation ficw must be manually controlled by the operator, as opposed to automatic control, whenever the system is operating in the single-loop mode, since control stability is improved in the manual mode.
Manual control is asst:ned in NED0-21281.2 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 room will change, such as individual jet-pump flow and total surmed core flow.
Some indications will be only slightly less accurate, but some others will be The control room indications must be corrected prior to single-erroneous.
loop power operation or they must be tagged out-of-service, as appropriate.
Thi:; is a requirement of Section 4.20 of IEEE Std-279-1971.4 The normal plant configuration 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 recir-culation loop that is out of service and whose water has cooled must be run
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in the bypass mode to preheat the water to within 100 F of the reactor cooling water before the wcter may be valved back to the reactor pressure 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.
Al though interlocks are provided, no credit is taken for their safety 2
function in NED0-21281 for s. ingle-loop operation since this is not the limiting transient.
The inctrument 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 extensive unscheduled mainte-nance without the requirement of keeping the reactor shut down.
It is stipulated that single-loop operation will not be a planned mode of operation.
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i The new red-block and trip setp'oints 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
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3.1 INTRODUCTION
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The initial licensee submittal analyzed in this report was based
[ h on the analysis in NED0-21281 performed by the nuclear steam supply system 2
s manufacturer (GE-NED).
The manufacturer, however, had not analyzed the i.h performance of the emergency core cooling system (ECCS) during single-loop h
operation.
h-j A new analysis was performed by GE-NED for 7
a LOCA with one
= T recirculation loop out of service.
The analysis, reported in NED0 s
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20566-2,3 includes the ECCS single loop analysis.
This NEDO report was
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- ublished subsequent to the initial licensee submittal..
The licensee will submit more detailed information based on the new analysis.
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v 3'.2 CONCLUSIONS
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We conclude that the concept outlined in the Carolir.a Power and h.
'ight Cmpany's initial license amendment submittall is in accordance with o
he intent of IEEE-Std-271-1971.
At the conclusion of this review, the er specific design criteria noted herein were communicated to the licensee, as
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required by the NRC.
The licensee has agreed to address that criteria in f
-he detailed license amendment submittal which was necessitated by the
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"evised GE report.
We recommend that the NRC review the EI&C aspects of
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- be forthecming detailed license amendment submittal for conformance to h
- hese, criteria.
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REFERENCES 1.
CP&L Letter to NRC (B. Rusche), dated September 3,1976.
2.
General Electric Company, Nuclear Energy Division, Brunswick, Steam Electric Plant Unit 2 License Amendment Submittal for Single-Loop Operation With the Bypass Fbw Holes Plugged and with LPSI Modification, hEDO-21281 (May 1976).
3.
General Electric Company, Nucle ~ar 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, Pai+. 50.46:
Acceptance Criteria for Emergency Core Coolin Systems for Light Water Nuclear Power Reactors (January 1976)g 6.
Code of Federal Regulations, Title 10, Part 50, Appendix A:
General Design Criteria for Nuclear Power Plants (January 1978).
7.
Code of Federal Regulations, Title 10, Part 50, Appendix K:
ECCS Evaluation Models (January 1,1978).
8.
Ge' neral Electric Company, Standard Boiling Water Reactor Tech-nicalSpecifications(n.d.).
9.
NRC/RSB, Protection System Trio Point changes for Operation With Reactor Coolant Pumo Out of Service, Branch Technical Position ICSB 12 (n. d.).
10.
CP&L, Final Safety Analysis Report for Brunswick Steam Electric Plant (FSAR) (n. d.J.
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NUCLEAR REGULATORY COMMISSION E
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Docket Nos. 50-325 August 24, 1981 and 50-324 Mr. J. A. Jones Senior Executive Vice President Carolina Power & Light Company 336 Fayetteville Street Raleigh, North Carolina 27602 RE: BRUNSWICX STEAM ELECTRIC PLANT UNITS 1 AND 2
Dear Mr. Jones:
On July 2,1921, we sent a letter to all licensees who have requested approval to operate on a continuing basis at power levels above 50". with only one recirculation loop in the event the other icop is inoperative. You and other SWR licensees received a copy of one of these letters since we expect most BWR facilities would like to have this flexibility.
In the letter we proposed a meeting to obtain a better understanding of what might have caused variations in' jet pump flow and related parameters at Browns Ferry Unit No.1 during single loep operation and how this incident should affect approval of single loop operation at,other facilities.
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You have indicated to your NRC project manager that you are interested in attending the proposed meeting. The meeting will be held at 9:00 A.M.,
Wednesday, September 9,1981 in room P-118, Phillips Building, 7920 Norfolk Avenue, Sethesda, Maryland.
You are requested to advise your project manager of thd people who will be attending this meeting from your organization.
Sincerely, OYa ThomaM. Ippolito, Chief Operating Reactors jiranch #2 Division of Licensing
Enclosure:
Meeting Agenda cc w/ enclosure See next page
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Proposed Peeting with BWR Applicants and 1.icensees.c'n Single toop Operation, i
t Purpose of Meeting:
1.
To determine shat may have ca0 sed the jet _ pump ficw "~ " l and other variations experienced by Srewns Ferry., Unit _1 ;
'during siiiil'5 Ecp' clieration arid - -.
2.
Evaluate whether the Browns Feriy experier.ce should resQlt in oower' limits.for other BWRs operating en a single loop.
Agenda:
1.
Discussien of what may have caused the unexpected variatiens in operating parameters when Browns Ferry Unit i exceeded about 60 percen: rated pcwer while operating with anly one recirculation lo.op.
2.
Discussion of parameters affected at Br:wns Ferry 1 (i.e., jet pump flew, neutron flux, core flow, core pressuredrop,etc.)
3.
Discussion of whether the Browns Ferry 1 experience would be expected at other BWF.s operating en one recirculaticn loop.
If wre safety limits likely.to -
be violated or cause complicatiens with respect to core ' -
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-stability, core ficw spmetry, pump cavitation or damage -
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tc the jet pumps and reacter vessel internais.
4 Discussion of the benefits vs. p:ter,tial pichlems ahd cost of testing single loop operatien in another SWp.
that is instrumented to detect what paracaters are affected.
5.
Evaluation of whether single loop operation at ;o'wer levels about 50 to 55 percent is a safe and prudent means of reactor operation.
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September 23, 1981 Docket Nos. 50-325 and 50-324 Mr. J. A. Jones Senior Executive Vice Fresident Carolina Power & Light Company 336 Fayetteville Street Raleigh, North Carolina 27602 RE:
BRUNSWICK STEAM ELECTRIC PLANT UNITS 1 AND 2
Dear Mr. Jones:
My letter to you of August 24, 1981 infomed you that we were proposing a meeting with licensees who have requested approval to operate on a single recirculation loop. The purpose of the meeting is to-determine what may have caused variations in jet pump flow at Browns Ferry Unit No.1 while operating on a single loop and what impact this should have on approval of other facilities to operate on one loop.
Licensees and applicants who have not requested approval for single loop operation were also invited to the meeting.
As you were advised by your licensing project manager, the meeting scheduled for September 9,1951 had to be postponed to allow more time for analysis of relevant. operating data. We apologize for this incon-venience. The meeting will be held at 9:00 AM on Thursday, October 22, 1981 in Room P-118, Phillips Building, 7920 Norfolk Avenue, Bethesda, Maryland.
It would be appreciated if you would inform your licensing project manager of the number of people who will be attending this teeting from your organization.
Sincerely, 4
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G# _ 60 Thomas ' ' ppolito, Chief Operating Reactors Branch #2 Division of Licensing cc: See Next Page l
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d Mr. J. A. Jones cc:
Richard E. Jones, Esquire,.
Carolina Power & Light Company 536-Fayetteville Street-Raleigh, North Carolina 27602 George F. Trowbridge, Esquire Shaw, Pittman, Potts & Trowbridge 1500 M Street, N. W.
Washington, D. C.
20035 Jchn J. Burney,J Jr., Esquire Burney, Burney, Sperry & Sarefoot 110 North Fifth Avenue Wilmington, North Carolina 28401
?.esident Inspector U. -S. Nuclear Regulatory Commission P. O. Sox 1057 50,uthport, Ncrth Carolina 2S461 5:uth; ort - Brunswick County Library IC9 W. Moore Street Southport, North Carolina 23461 Mr. Charles'R. Dietz 71 ant Manager 4
F. O. Sox 458 Southport, North Carolina 28451
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pc UNITED STATES
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Docket rio. 50-324 October 9,1981 i
Mr.-J. A. Jones Senior Executive Vice President Carolina Power & Light Company 336 Fayetteville Street Raleigh, fiorth Carolina 27602
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Dear Mr. Jones:
RE:
BRUNSWICK STEAM ELECTRIC PLANT UtlIT tt0. 2 We have reviewed your single loop license amendment request dated August 12, 1981.
In that request you assert that the potential effects of a pump seizure accident are bounded by the effects of a Loss of Ccolant Accident (LOCA).
However, the acceptance criteria for a pump seizure are more stringent than the criteria for a LOCA.
Standard Review Plan 15.3.3 (Reactor Coolant Pump Rotor Seizure and Reactor Coolant Pump Shaft Break) requires that for a pump seizure accident any a.ctivity release must be such.that the calculated doses at the site boundary are a sma'll fraction of the 10 CFR Part 100 guidelines. This is significantly more limiting than the corresponding LOCA criteria. We request therefore that you submit an analysis that demonstrates the acceptability of a pump seizure accident during single loop operation.
We will.not be'able to complete action on your request until we have received and reviewed this. additional information.
Sincerely, Q,e.-
Thomas W. Ippolito, Chief Operating Reactors Branch #2 Division of Licensing tc:
See flext Page l
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l Mr. J..A. Jones v-s' cc:
Richard E. Jones, Esquire
' Carolina Power & Light Company 336 Fayetteville Street i
Raleigh, North Carolina 27602 George F. Trowbridge, Esquire Shaw, Pittman, Potts & Trowbridge 1800 M Streat, N. W.
Washington, D. C.
20036 John J. Burney, Jr., Esquire Burney,- Burney, Sperry & Barefoot 310 North Fifth Avenue Wilmington, North Carolina 28401 Resident Inspector U. S. Nuclear Regulatory Commission
. P. O. Box 1057 Southport, North Carolina 28461 Southport - Brunswick County Library 109 W. Moore Street Southport, North Carolina 28461
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Mr. Charles.'R. Dietz
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Piant Manager
.P. O. Box 458 Southport, North Carolina 28461 e
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