ML20031G116
| ML20031G116 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 09/30/1981 |
| From: | Udy A EG&G, INC. |
| To: | Scholl R Office of Nuclear Reactor Regulation |
| References | |
| CON-FIN-A-6425, REF-GTECI-B-24, REF-GTECI-EL, REF-GTECI-ES, TASK-06-04, TASK-6-4, TASK-B-24, TASK-OR, TASK-RR EGG-EA-5564, NUDOCS 8110210121 | |
| Download: ML20031G116 (6) | |
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EGG-EA-5564 SEPTEMBER 1981 SYSTEMATIC FVALUATION PP0 GRAM TOPIC VI-4, ELECTRICAL, ANA Dr2S INSTRUMENTATION AND CONTROL ASPECTS OF THE OVERRIDE OF g
CONTAINMENT PURGE VALVE ISOLATI0fi, LA CROSSE BOILING d[
WATER REACTOR NRC leSearc.i ancTecinica' [afC/[Dk
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This is an Informal report inten.- d for use as a preliminary or working document Prepared for the U.S. Nuclear Regulatory Commissio:.
Under DOE Contract No. DE-AC07-76ID01570 FIN No. A6425 g
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FORM EG&C We (Rev 1179)
INTERIM REPORT Accession No.
Report No.
EGG-EA-5564 Contract Program or Project
Title:
Electrical, Instrumentation, and Control Systems Support for the Systematic Evaluation Program (II)
Subject of this Document:
Systematic Evaluation Program Topic VI-4, Electrical, Instrumentation and Control Aspects of the Override of Containment Purge Valve Isolation, La Crosse Boiling Water Reactor Type o' Document:
Inforrul Report Author (s):
A. C. Udy h;RC Researci anc1%cinica' D te of Document:
[ggg*
p0f Septee.bce 1981 R;sponsible NRC Individual and NRC Office or Division:
Ray F. Scholl, Jr., Division of Licensing This docur nt was prepared primarily for preliminary orinterra use. lt!,as not received full review and approval. Since there may be substantive changes, this document s' ould n
not be considered final.
I EG&G Idaho, Inc.
Idaho Falls, Idaba 83415 l
Prepared for the
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U.S. Nuclear Regulatory Commission Washington, D.C.
Under DOE Contract No. DE-AC07 761D01570 NRC FIN No.
A6425 INTERIM REPOPT 1
I 0386J SYSTEMATIC EVALUATION PROGRAM TOPIC VI-4 ELECTRICAL, INSTRUMENTATION, AND CONTROL ASPECTS OF THE OVERRIDE OF CONTAINMENT PURGE VALVE ISOLATION LA CROSSE BOILING WATER REACTOR Docket No. 50-409 l.
September 1981 A. C. Udy
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Reliability and Statistics Branch Engineering Analysis Division r
EG&G Idaho, Inc.
i NRC Researc1 and "ecanica Assistance Report 9-9-81 9
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ABSTRACT This SED technical evaluation, for the La Crosse Boiling Water Reactor, resiewt the electrical, instrumentation and control system design aspects of the containinent ventilation isolation system and other related engineered safety feature systems.
FOREWORD This report is supplied as part of the " Electrical, Instrumentation, and Control Systems Support for the Systematic Evaluation Program (II)"
being conducted for the U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Division of Licensing by EG&G Idaho, Inc.,
Reliability & Statistics Branch.
The U.S. Nuclear Regulatory Commission funded the work under the authorization B&R 20-10-02-05 FIN A6425.
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CONTENTS
1.0 INTRODUCTION
1 2.0 EVALUATION OF THE LA CROSSE BOILING WATER REACTOR 2
2.1 Review Guidelines.........................................
2 2.2' Containment Ventilation Isolation Circuits Design Description...............................................
3 2.3 Containment Ventilation Isolation System Design Evaluation-4 2.4 Other Related Engineered Safety Feature System Circuits 5
3.0
SUMMARY
,5
4.0 REFERENCES
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L SYSTEMATIC EVALUATION PROGRAM TOPIC VI-4 ELECTRICAL, INSTRUMENTATION, AND CONTROL ASPECTS OF THE OVERRIDE OF CONTAINMENT PURGE VALVE ISOLATION LA CROSSE BOILING WATER REACTOR
1.0 INTRODUCTION
Based on the information supplied by the Dairyland Power Cooperative.
(DPC), this report addresses the electrical, instrumentation, and control systems design aspects of the Containment Ventilation Isolation (CVI) system and other related Engineered Safety Feature (ESF) functions for the La Crosse Boiling Water Reactor.
Several instances have been reported where the automatic closure of the containment ventilation or purge isolation valves would not have occur-red because the safety actuation signals were manually overridden or blocked during normal plant operations. Lack of proper mdnagement con-trols, procedural inadequacies, and circuit design deficiencies contributed to these instance:
These events also brought into question the mechanical opardi?ity of the valves themselves. These events were determined by the Nuclear Regulatory Comission (NRC) to be an Abnormal Occurrence (#78-05) and accordingly, were reported to Congress.
The NRC is now reviewing the electrical override aspects of contain-me:t purging and venting for all operating reactors. On November 28, 1978, the NRC issued a letter, " Containment Purgirg During Normal Plant Opera-tion"I to all Boiling Water Reactor and Pressurized Water Reactor licensees. This required a review of these systems by the licensee. DPC 2
3 responded on February 1, 1979. On February 20, 1981, DPC provided additional information requested by the NRC. The Final Safety Analysis Rer. ort (FSAR) and a letter of January 14, 1980,4 also contain design
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information reviewed for this report. DPC comments of August 26, 19815 were also evaluated for this eport.
1
2.0 EVALUATION OF THE LA CROSSE BOILING WATER REACTOR i
2.1 Review Guidelines.
The intent of this evaluation is to determine if the actuating signals for the ESF equipment meet the following NRC requirements:
1.
Guideline No. 1--In keeping with the requirements of Generai Design Criteria 55 and 56, the over-ridea of one type of safety actuation signal (e.g., radiation) should not cause the blocking of any other type of safety actuation signal (e.g.,
pressure) for those valves that have no function besides containment isolation.
2.
Geideline No.
--Sufficient physical features (e.g.,
key lock switches) are to be provided to facilitate adequate administrative controls.
3.
Guideline No. 3--A system level annunciation of the overridden status should be provided for every safety system impacted when any override is actit:3.
Additionally, this review uses the following NRC design guidelines:
1.
Guideline No. 4--Diverse signals should be provided to initiate $ solation of the containment ventil? tion system. Specifically, containment high radiation, safety injection actuation, and containment high pres-sure (where containment high pressure is not a portion of safety injection actuation) should automatically initiate CVI.
2.
Guideline No. 5--The instrumentation and control systems l
provided to initiate the ESF should be designed and qualified as safety grade equipment.
3.
Guideline No. 6--the overriding c: resettinga of the ESF actuation signal should r.ot cause any valve or damper to change position.
a.
The following definitions are given for clarity of use in this evaluation:
Override:
the signal is still present, and it is blocked in order to
' perform a function contrary to the signal.
Reset: the signal has come and gone, and the circuit is being cleared in order to return it to the normal condition.
l 2
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Guideline 6 in this review applies primarily to other related ESF systems because implemtntation of this guideline for containment isolation will be reviewed by the Lessons Learned Task Force, based on the recommen-dations in NUREG-0578, Section 2.1.4.
When containment isolation is not involved, consideration on a case-by-case basis of automatic valve' reposi-tioning upon reset may be considered acceptable. Acceptability would be dependent upon system function, design intent, and suitable operating procedures.
2.2 Containment Ventilation Isolation Circuits Design Description.
The containment purge and vent isolation valves use solenoid-operated air pilot valves. Loss of power or air will cause the isolation valves'to close. Automatic closure of the containment purge inlet and outiet isolation valves will occur on any of the following conditions :'
l.
High reactor containment building pressure (5 psig).
2.
High primary system pressure (1325 psig).
3.
Low reactor water level.
4.
High Radiation.
The four-inch vent header valve will also close on all of the above signals.5 During operation, these valves are closed. OPC has indicated,
3 that these signals are derived from safety grade equipment. SEP Topic III-12, " Environmental Qualification," will verify that they are.
The high primary system pressure signal can be bypassed for calibration by means of a test sw!tch on the front of the ' amplifier drawer of each channel. The amplifiar drawers are located in tile vertical contrai panel in the control room.
If both drawer test switches are in the " Test" position at the same time, or if either switch is opertted without first operating the key operated scram bypass switch, a reactor scram will result. The scram signal from each drawer cannot be bypassed at the same time. The use of either drawer test switch or the key operated scram bypass switch is annunicated.5 3
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2.3 Containment Ventilation Isolation System Design tvaluation.
Guideline 1 requires that no signal override can prevent another safety actuation signai from functioning.
The La Crosse Station complies with this guideline.
Guideline 2 requires that reset cnd override switches have physical provisions to aid in the administrative control of these switches. The key-locked bypass switch, previously mentioned, ;omplies with this guide-line. DPC installed a locked cover over all tha reset switches that are not key operated.5 The high primary pressure test switches, since they are used in conjunction with the key-locked byphss switch, also comply with this guideline.
Guideline 3 requires system level annunciation whenever an override affects the performance of a safety system. The use of the override is dnnunciated in Conformance with this guideline.
Guideline 4 requires that isolation of the CVI valves be actuated by several diverse signals. This requirement is met. Manua? actuation of either the high pressure core spray or the alternate core spray systems will not actuate closure af the CVI valves.
Guideline S requires that isolation actuation signals be deri.'ed from safet/ grade equipment. DPC indicates that the isolation actuation signa'ls a l qualifieG to opc: ate in their normal environment. Should isolation be necessary, it would be accomplis *.ed prior to any significant exposure to radiation, temperature or pressure. Ste Topic III-12 will further examine the environmental qu lifications of this equipment.
Guideline 6 requires that no reset of isolation logic will automati-cally open che isolation valves. OPC indicates thd no valve or damper I
will change position when a containment isolation signal is overridden or reset.3 4
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2.4 Other Related Engineered Safety Feature System Circuits. A review of other related ESF circuits was also made. No other manual overrides have been identified in the review of the material submitted for this audit.
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3.0
SUMMARY
The NRC issued a letter. " Containment Purging During Normal Plant Operation," which requested DPC to review purging requirements, controls, ar.d procedures for purging at the La Crosse station.
The electrical, instrumentation, and control design aspects of the containment ventilation isolation. valves for La Crosse Unit I were evaluated -
using the design guidelines stated in Section 2.1 of this report. These guidelines are satisfied. However, automatic isolation will not occur for maraal operation of the high pressure core spray or the alternate core spray systems. The NRC should determine if this is acceptable.
Both redundant CVI 20-inch valves are' located inside containment.
This is not presently acceptable per General Design Criteria 56 which requires one valve to be inside containment and the other valve outside containment.- The NRC should determine the continued acceptability of this exemption to this General Design Criteria.
4.0 REFERENCES
I i
1.
NRC/ DOR letter (A. Schwencer) to DPC and all BWR and PWR licensees,
" Containment Purging During Normal Plant Operation," dated November 28, 1978.
1 2.
DPC letter, F. Linder to the Director of Nuclear Reactor Regulation, NRC, " Containment Purging During Normal Plant Operation," February 1, 1979, LAC-6104.
3.
DPC letter, F. Linder to the Director of Nuc' ear Reactor Regulation, I
NRC, " Bypass and Reset of Engineered Safety Features," February 20, 1981, LAC-7379.
2 4.
DPC letter, F. Linder to the Director of Nuclear Reactor Regulation, i
NRC,-"IE Bulletin No. 79-08. Additional Information-Fvents Relevant-to BWR's Identified During Three Mile Island Incident,". January 14, 1980, t AC-6732.
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MC letter, F. Linder to the Director of Nuclear Reactor Regulation, NRC, "SEP Topic VI-4, Override of Containment Purge Isolation and other ESF Actuation Signals (La Crosse), August 26,1981, LAC-7761.
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