ML20091K782
| ML20091K782 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Browns Ferry |
| Issue date: | 08/31/1983 |
| From: | Donich T LAWRENCE LIVERMORE NATIONAL LABORATORY |
| To: | Dunesia Clark Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML17139C189 | List:
|
| References | |
| CON-FIN-A-0250, CON-FIN-A-250, FOIA-84-105 NUDOCS 8406070200 | |
| Download: ML20091K782 (7) | |
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2-4 TECHNICAL EVALUATION OF THE ELECTRICAL, INSTRUMENTATION, AND CONTROL CESIGN ASPECTS OF THE PROPOSED LICENSE AMENDMENT REVISION 1 FOR SINGLE-LOOP OPERATION 0F BROWNS FERRY NUCLLAR PLANTS (Docket No. 50-259) - Unit 1 (Docket No. 50-260) - Unit 2 (Occket no. 50-295) - Unit 3 Terry R. Donich Lawrence Livermore National Laboratory.
Nuclear Systems Safety Program
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FIN A0250 Responsible NRC Individual and Division Dick Clark Division of Licensing THIS DOCUMENT WAS PREPARED PRIMARILY FOR PRELIMINARY OR INTERNAL USE.
IT HAS NOT RECEIVED FULL REVIEW AND APPROVAL.
SINCE THERE MAY BE SUBSTANTIVE CHANGES, THIS DOCUMENT SHOULD NOT BE CONSIDERED FINAL.
0 8406070200 840319 PDR FOIA BELL 84-105 PDR
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f f:07 ICE Tnis recort was prepare: as an account cf work sconsored oy the. agency of tne United States Government.
Neither the United States Severnment nor any agency thereof, or any of their employees, makes any warranty, expressed or implied, or assumas any legal liability or responsibility for any third party's use, or the results of sucn use, of any information, apparatus proouct or process disclosed in this r,eport, or represents that its use by such third
- arty would not infringe privately owned rights.
ABSTRACT This report documents the technical evaluation of the proposed changes to the plant reactor protection system by the licensee of Browns Ferry Nuclear power Stations, Units 1, 2, and 3, to account for single-loop plant operation.
This evaulation is restricted to only the electrical, instrumentation and control design aspects of proposed chan'ges to the plant technical specifications for' single-loop operation beyond 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />'s. The conclusion of the evaluation is that the Browns Ferry Nuclear Power Statiend Y
Ur.its 1, 2, and 3 license amencment for single-loop operaticn has met the review criteria provided sufficient acministrative controls are put.into effect.
FOREWARD This report is supplied as part of the Selected Operating Reacter Issues Program II being conducted for the U. S. Nuclear Regulatory Commission, Office
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of Nuclear Reactor Regulation, Division of Licensing, by Lawrence Livermore
- ational Laboratory.
The U. S. Nuclear Regulatory Commission funded the work under the classification entitled " Selected Operating Reactor Issues Program II, B & R 20 19 10 11 1, FIN No. A-0250.
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INTR 000CT10N By letter to the U.S. Nuclear ' Regulatory Commission (NRC) dated March 4, 1982 [Ref.1], the Tennessee Valley Authority (TVA) submitted informatien to y,/,,, /4-6 support it's propcsed license amencment to operate the Browns Ferry /" Units-1, 2, and 3 (BF-1, SF-2, and BF-3, respectively) with one recirculation loop out of service (i.e. single-loop operation). This information included the licensee's analysis of significant events, which were based on a review of.
accidents and abnormal operational transients, associated with power operations in the single-loop mode provided by General Electric Ccmoany, Nuclear Energy Division (GE-NED), the nuclear steam rJpply system designer.
Conservative assunptions were employed, as ciscussed in the GE-NED report NE00-24236 catec May,1981 [Ref. 2], to ensure that the generic analyses for boiling water reactors (BWR 3 and/or 4) were applicable to the Browns Ferry.
In response to an NRC request, the licensee provided supplemental information in a letter cate_d September 3,1932 [Ref. 3].
The purpose of this report is to document the evaluation of 'th'e electrical, instrumentation, and control (EI&C) cesign aspects of the pr3cosed license untacment change to the Browns Ferry technical specifications. The consideration of proper plant variables, computer models, anc the licensee's cone'usions on core performance and clad temperature are outside the scope of this evaluation. This review was conducted using 10CFR50, Appendix A, l
" General Design Criteria for Nuclear Power Plants" (G.D.C. 20 through 24)
.:Ref. 42 anc ANSI /IEEE Std 279-1971 [Ref. 5: with the following guidance from -
tne NRC staff for the application of Section 4.15 of the ANSI /IEEE Stancard:
Manual switening to the more restrictive setpoint for the APRMs in the reactor protection system is acceptable for BWRs if sufficient administrative controls exist to assure that the more restrictive setpoints are in effect when required by the plant Technical Specifications.
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EVALUATION 40 RECOMMEEATIONS The current Browns Fe*ry Technical Specifications ce not permit-single-loco plant operation at reduced power for = ore tnan 24 nours.
The
= licensee's proposed Technical Specification changes would allow the reactor to coerate at reduced power (not greater than 507.) with one recirculation loop inoperable for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if certain changes are made to the reactor protection systems. Specifically, the changes are to the Average Power Range Monitor ( APRM) scram trip setpoint and the Rod. Block Monitor (RBM) rod block setpoint.
A different flow pattern is established in the vessel during single recirculation loop operation as compared to the normal two loop operation 4
[Ref. 2).
In ' single-loop operation, there is backflow through the jet pumps in the idle loop.
The jet pump core flow measurement system is calibrated only-when both loops are in operation and all jet pumps are in forward flow.
The total core flow is the sum of the measured jet pump flows.
In single loop operation, the measured flow in the back-flowing jet pumps must be subtracted from the flow through the other jet pumps.
Also, the jet pump flow coefficient is diff erent for reverse flow than for forward flow to the jet I
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Because of the different flow rate and. flow path during single recirculation loop operation, the APRM SCRAM trip settings, which are flow-biased according to the. equation in the proposed technical specifications, reouire resetting to protect the reactor from overoower. The red-block setpoint equation is flow-biased in the same way and with the same flow signal as tne APRM setpoint, and must also be modified to provide adequate core p'rotection for a postulated rod withdrawal error.
The manual _ APRM gain adjustment to acconnodate single-loop operation is the only_ change imposed upon the Browns Ferry reactor protection system j
(RPS). This modification adds the term 0.66 W-to the APRM readings to canpensate for backflow through the jet pumps in the idle loop.. The licensee stated that sufficient range exists in the APRM gain settings to make the necessary changes to the RPS for single-loop operation'[Ref. 3].' This change will not cause the RPS to violate General Design Criteria 10 to 24 of 10CFR50 Appendix A.
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~ The -licensee'ined:a ec in Reference 3 :nat ::rocccures cover the APRM gain U
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- a,djustment to accour.: fc-single loop operation.
The procedures are in surveillance instructior (SI) 4.1.5*.15 and 2.1.
Tne Itcensee efd not state what acministrative cen PCls were te be used to assure that the gain
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adjustments are performed-ccrrectly.
We recocynenc tnat the licensee provide NRC1cocumentaticn te ensure the necessary gain adjustments have sufficient administrative controls and are therefore consis'ent with Section 4.15 of t
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CONCLUSION!
sased on Our review of tne informa-ion anc cocuments or videc by the
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licensee in Ref. 1, we conclude tnat the more conservative reactor protection sys:em (RPS) setooint trips for the APRM and R3M will satisfy tne functional recu'rements (i.e., parameters te be monitorec, setpoints, etc.) for single-recirculation-loop operation.
The manual APRM gain adjustment settings,tc acccamodate single loop I
operation is the only change imposed upon the Browns Ferry Nuclear P?d tt 7
r rea::or orotection system (RPS) instrumentation.
This change will no: cause the RPS instrumentation system to violate 10CFR50 Appendix A General Design Criteria 20 through 24 [Ref. 4) or IEEE-279-1971 [Ref. 5) with the exceotion of tne IEEE Standard discussed below.
, Because o'f the backflow through the jet pumos during single-recirculation-loop operation, indications in the control rocm of individual jet-pump flow.and total summed core flow will be misle'ading. We reccomenc that these anomalous control room indications must be corrected or warr'r.;-:agged f or the duration of the single-recirculaticn-loop Operaticr., as rec f ree oy section 4.20 of IEEE Std-27g-1971 [Ref. 5).
We recommend that the licensee provide assurance that manual switching to the mere restrictive setpoint for the APRMs has sufficient acministrative c09 ': s to assure that the more restrictive setpoints are in effect wnen req. ired by tne plant Technical Specificaticns.
We recommend to NRC that upon successful implementation of the above re. commanded actions the proposed licensee amendment for single-recirculation-~
loop operation at Browns Ferry Units 1, 2 and 3 meets the requirements.
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l REFERENCES M:ll:
(1) Tennessee Valley Autfgity Letter (L. M. Mill { to NRC' (Harold Denton),
x4 "In the Matter of the Tennessee Valley Authority", March 4,1982.
(2) General Electric Company, Nuclear Power Systems Division, " Browns Ferry Nuclear Planty Units 1, 2 and 3 Single-Loop Operation", NEDO 24236, May A
1981.
(3) Tennessee Valley Authority Letter (L. M. Mills) to NRC (Harold Denten),
" Single-Loop Operation for Browns Ferry, Response to Questions," dated September 3, 1982.
(4) Code of Federal Regulations, Title 10, Part 50, Appendix A, " General Design Criteria for Nuclear Power Plants",1981.
(5)
IEEE Std-279-1971, " Criteria for Protection Systems for Nuclear Power Generating Stations", dated 1971.
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