ML20041G031
| ML20041G031 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 02/28/1982 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-0420, NUREG-0420-S02, NUREG-420, NUREG-420-S2, NUDOCS 8203190075 | |
| Download: ML20041G031 (32) | |
Text
NUREG-0420 Supplement No. 2 Safety Evaluation Report related to the operation of Shoreham Nuclear Power Station, Unit No.1 Docket No. 50-322 Long Island Lighting Company U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation February 1982
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NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in N RC publications will be available from one of the following sources:
- 1. The NRC Public Document Room,1717 H Street, N.W.
Washington, DC 20555 l
- 2. The NRC/GPO Sales Program, U.S. Nuclear Regt.. :ory Commission, Washington, DC 20555 l~
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Referenced documents available for insocction and copying for a fee from the NRC Public Docu-ment Room include NRC correspondence and internal NRC memoranda: NRC Office of Inspection and Enforcement bulletins, circulars, information notices, inspection ar.d investigation notices; Licensee Event Reports; vendor reports and correspondence; Commission papers;and applicant and l
licensee documents and correspondence.
The following documents in the NUREG series are available for purchase froni the NRC/GPO Sales Program: formal NRC staff and contractor reports, NRC-sponsored conference proceedings, and l
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GPO Pnnted copy pnce. N. 7_5
,I NUREG-0420 c
Supplement No. 2 l
1 Safety Evaluation Report related to the operation of Shoreham Nuclear Power Station, Unit No.1 Docket No. 50-322 l
Long Island Lighting Company l
l U.S. Nuclear Regulatory s
Commission l
Office of Nuclear Reactor Regulation February 1982 s
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TABLE OF CONTENTS 6'
1 Page 1
INTRODUCTION AND GENERAL DISCUSSION.........../......'.............
1-1 1.1 Introdution.....(..............
1-1
- 1. 7 Outstanding Issues,'
1-2
[6 ENGINEERED SAFETY FEATURES......
6-1 6.2.1.7 Steam Bypass of the Su pression Pool.
6-1 6.3 Emergency Core Cooling Sy, stem.............................
6-1 9
AUXILIARY SYSTEMS....
9-1 9-1 9.4.1 Control Room Air Conditioning System...
4e 9.5' ' Fire ProtUction System..
9-1 fi3 CONDliCT OF OPERATIONS.............
13-1 2
'I3.5.2 Operating and Maintenance Procedures.
13-1 n
4 15
-ACCIDENI ANALYSIS...............
15-1
-e 15-1 7 anticipated Iransients Without Scram.
15.3
/
18 P5 VIEW 8Y THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 18-1 TMI-2REQObEtENTS.........................
22-1 22
't I.D.1 Control Room Design Review.
22-1 II.B.4
' Degraded Core Training................
22-3 II.F.1.
Additional Accident-Monitoring Instrumentation...
22-4 APPENDIX A ~.< ERRATA TO,THE SER....................................
A-1 APPENDIX B ERRATA T0 THE SSER #1...............
B-1 i
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l Shoreham SSER #2 i
1 INTRODUCTION AND GENERAL DISCUSSION 1.1 Introduction The Nuclear Regulatory Commission's Safety Evaluation Report (NUREG-0420) on the application by Long Island Lighting Company (LILC0 or applicant) to operate the Shoreham Nuclear Power Station was issued on April 10, 1981.
Supplement No. 1 to the Shoreham Safety Evaluation Report was issued on September 9, 1981.
The purpose of this supplement is to provide:
(1) the report of the Advisory Committee on Reactor Safeguards, (2) revisions to sections of the Safety Evaluation Report where further discussion or changes are in order, and (3) the results of our review of letters submitted by the applicant to address certain outstanding issues identified in Section 1.7.
The information provided in these letters must be acceptably documented in Amendments to the Final Safety Analysis Report prior to licensing.
Each of the sections in this supplement is numbered the same as the section of the Safety Evaluation Report that is being updated.
The discussions in this report are supplementary to and not in lieu of the discussions in the Safety Evaiuation Report, except where specifically noted.
Copies of this report are available for public inspection at the Commission's Public Document Room, 1717 H Street NW, Washington, D.C. and at the Shoreham-Wading River Public Library, Route 25A, Shoreham, New York 11786.
Copies of this report are also available for purchase from the sourpes indicated on the inside front cover.
The NRC Project Manager assigned to the operating license application for Shoreham is Dr. Robert A. Gilbert.
Dr. Gilbert may be contacted by calling (301) 492-7000 or writing to the following address:
Dr. Robert A. Gilbert Division of Licensing U.S. Nuclear Regulatory Commission Washington, DC 20555 This Safety Evaluation Report is a product of the NRC staff.
The folluing NRC staff members and consultants contributed to this report:
Raj K. Anand - Auxiliary Systems Engineer J. D. Behn Consultant - Gage-Babcock Thyagaraja Chandrasekaran - Nuclear Engineer James W. Clifford - Operational Safety Engineer S. Crowell - Consultant - Battelle Pacific Northwest Laboratories L. Gefferding - Consultant - Battelle Pacific Northwest Laboratories Vincent A. Deliso - Reactor Safeguards Analyst Richard Eckenrode - Human Factors Analyst Mel B. Fields - Containment Systems Engineer Michael J. Goodman - Engineering Phychologist Shoreham SSER #2 1-1
l Charles S. Hinson - Health Physicist William T. LeFave - Auxiliary Systems Engineer M. Morganstern - Consultant - Battelle Pacific Northwest Laboratories R. Shikiar - Consultant Battelle Pacific Northwest Laboratories Summer B. Sun - Nuclear Engineer Richard J. Urban
.0perational Safety Engineer 1.7 Outstanding Issues In Section 1.7 of the Safety Evaluation Report, we identified 61 outstanding issues which were not resolved at the time of issuance of the Safety Evaluation Report.
In this report we discuss the resolution of a number of these items previously identified as open.
The items identified in Section 1.7 of the Safety Evaluation Report are listed below with the status of each item.
If the item is discussed in this supplement, then the section where the item is discussed in this report is identified.
The resolution of the remaining outstanding issues will be discussed in future supplements to the Safety Evaluation Report.
Item Status Section 1.
Pool Dynamic Loads Awaiting further information 2.
Masonry Walls Resolved 3.
Piping Vibration Test Program -
Resolved Small Bore Piping / Instrumentation Lines 4.
Piping Vibration Test Program -
Resolved Safety Related Snubbers 5.
LOCA Loadings on Reactor Vessel Resolved Supports and Internals 6.
Downcomer Fatigue Analysis Resolved pending confirmation 7.
Piping Functional Capability Criteria Resolved 8.
Dynamic Qualification Awaiting second audit 9.
Environmental Qualification Awaiting audit
- 10. Seismic and LOCA Loadings Resolved pending confirmation
- 11. Supplemental ECCS Calculations with Resolved with NUREG-0630 Model-license condition
- 12. ODYN-Generic Letter 81-9 Resolved Shoreham SSER #2 1-2 4
Item Status Section 13.
NUREG-0619, Feedwater Nozzle and Control Rod Return Line Cracking Resolved Generic Letter 81-11 14.
Jet Pump Holddown Beam Resolved
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15.
Inservice Testing of Pumps and Valves Res'olved pending confirmation 16.
Leak Testing of Pressure Isolation Resolved Valves 17.
SRV Surveillance Program Resolved 18.
NUREG-0313, Revision 1 Resolved 19.
Preservice Inspection Resolved pending confirmation 20.
Appendix G - IV.A.2.a Resolved 21.
Appendix G - IV.A.2.c Resolved 22.
Appendix G - IV. A.3 Resolved 23.
Appendix G - IV.B Resolved 24.
Appendix H - II.C.3.b Resolved 25.
RCIC Re',olved 26.
Suppression Pool Bypass Staff Position 6.2.1.7 l
27.
Steam Condensation Downcomer Lateral Resolved Loads 28.
Steam Condensation Oscillation and Resolved pending Chugging Loads confirmation 29.
Quencher Air Clearing Load Resolved 30.
Drywell Pressure History Resolved 31.
Impact Loads on Grating Resolved 32.
Steam Condensation Submerged Drag Resolved pending Loads confirmation 33.
Pool Temperature Limit Resolved Shoreham SSER #2 1-3 u
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Item Status Section 34.
Quencher Arm and Tie-Down Loads Resolved 35.
Containment Isolation Resolved 36.
Containment Purge System Resolved pending confirmation 37.
Secondary Containment Bypass Resolved l
Leakage 38.
Fracture Prevention of Containment Resolved l
Pressure Boundary 39.
Emergency Procedures Awaiting LILCO submittal l
40.
LOCA Analyses Resolved 41.
LPCI Diversion Resolved 42.
Flow Meter Resolved 43.
Loss of Safety Function Resolved l
After Reset 44.
Level Measurement Errors Resolved l
45.
Fire Protection Resolved 9.5.3.2 46.
ole Bulletin 79-27 Resolved pending i
confirmation 47.
Control System Failures Resolved pending confirmation l
l 48.
High Energy Line Breaks Resolved pending confirmation 49.
D.C. System Monitoring Resolved 50.
Low and/or Degraded Grid Resolved Voltage Condition 51.
Fracture Toughness of Steam Resolved
& Feedwater Line Materials 52.
Management Organization Awaiting further information l
53.
Emergency Planning Under review Shoreham SSER #2 1-4
Item Status Section 54.
Security Resolved I
55.
Q-List Resolved 56.
Financial Qualifications Resolved 57.
TMI-2 Requirements Shift Technical Advisor Resolved with license condition Shift Supervisor Administrative Resolved Duties Shift Manning Resolved i
Upgrade Operator Training Resolved Training Programs - Operators Resolved per. ding confirmation i
Revise Licensing Examinations Resolved Organization & Management Resolved Procedures for Transients & Accidents Resolved 13.5.2 Shif t Relief & Turnover Procedures Resolved Shift Supervisor Responsibilities Resolved Control Room Access Resolved i
Dissemination of Operatirq Resolved Experiences Verify Correct Performance of Resolved Operating Activities
=
Vendor Review of Procedures Resolved pending 13.5.2 l
confirmation l
Emergency Procedures Resolved pending 13.5.2 i
confirmation Control Room Design Review Resolved with I.D.1 license condition Training During Low-Power Testing Resolved Reactor Coolant System Vent:
Resolved Shoreham SSER #2 1-5 l
Item Status Section 2
Plant Shielding Resolved Postaccident Sampling Staff position Degraded _ Core Training Resolved II.B.4 Hydrogen Control Resolved Relief & Safety Valves Resolved pending confirmation j
Valve Position Indication Resolved Dedicated Hydrogen Penetrations Resolved Containment Isolation Dependability Awaiting further information Accident-Monitoring Instrumentation Resolved pending confirmation Resolved II.F.1 Resolved f
Resolved Resolved II.F.1 Resolved Inadequate Core Cooling _
License condition IE Bulletins Item 5-Resolved pending confirmation i
Item 10 Resolved pending confirmation Item 22 Resolved Item 23 Resolved i
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- Bulletins & Orders Task Force t
Item 3 Resolved l
Shoreham SSER #2 1-6 E
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Item Status Section Item 13 Resolved pending confirmation Item 15 Resolved Item 16 Resolved pending confirmation l
Item-17 Resolved 4
Item 18 Resolved i
. Item 21 Resolved Item 22 Resolved Item 24 Resolved Item 25 Resolved c
Item 27 Resolved Item 28
~
Awaiting further e
information Item 30 Resolved Item 31 Resolved ir Item 44 Resolved Item 45 Resolved pending confirmation Item 46 Resolved Emergency Preparedness-Short Term Under review Upgrade Emergency Support Facilities Under review j
Emergency Preparedness - Long Term Under review Primary Coolant Outside Containment Resolved Improved Iodine Monitoring Resolved i'
. Control Room Habitability Resolved pending confirmation 58.
Reactor Vessel Materials Toughness Resolved l
Shoreham SSER #2 1-7
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e Item-Status Section 59.
Resolved Generic Letter 81-07
- 60.. Station Blackout -
Resolved pending Generic Letter 81-04 confirmation 61.
Scram System Piping Awaiting further information
- 62.
Remote Shutdown System Awaiting further information In addition to the above outstanding issues, another concern has been identi-fied with regard to Shoreham.
We are seeking assurance that the facility has been designed and constructed in accordance with the application.
This concern is a-result of the situation at Diablo Canyon where errors were uncovered in design drawings and subsequent construction items.
We will meet with the applicant on March 15, 1982 to develop a program to assure proper design and construction prior to plant operation.
1 Shoreham SSER #2 1-8
6 ENGINEERED SAFETY FEATURES 6.2.1.7 Steam Bypass of the Suppression Pool In the first supplement to the Safety Evaluation Report, it was reported that the resolution of the steam bypass capability for the limiting small break accident (which results in a low pressure differential between the drywell and wetwell for the maximum period of time possible) would be a License Condition to be completed before the start of the first low pressure test of the drywell.
The basis for this License Condition was that the issue was an acceptance criterion for the low pressure test which would not take place until the first refueling outage.
Further study of this issue by the staff has shown that the preoperational high pressure test is not designed to shus that the low pressure bypass capability 4
(A/,/k) is confirmed.
We require that the acceptance criterion (equal to or less than 10% of the bypass capability) for the limiting small break be demon-strated before the fuel load date.
This can be accomplished by showing that the leekage from the high pressure test meets this criterion, or by performing a separate test at low pressure (equal to the hydrostatic head in the downcomers).
Since the applicant has not agreed to use our position on steam bypass capa-bility for Mark Il containments and since this issue must now be resolved before the fuel load date for the plant, the low pressure steam bypass cap-ability is now considered an open item on the Shoreham Docket.
We are currently in the process of reviewing the analysis supplied by the applicant and will require that this issue be resolved before the fuel load date.
6.3 Emergency Core Cooling System We recently reopened our review of core spray distribution based on information provided by the Advisory Committee on Reactor Safeguards on Japanese spray distribution tests of a simulated BWR/5 configuration in steam using a 60 i
sector test facility.
The test data show that central bundles receive low core spray flow due to maldistribution.
Although no specific data are available, we
'have also been told that the 360 tests by the Japanese with 5/6 of the spray nozzles blocked gave similar results to the 60 sector tests.
This information is of concern since credit is taken for core spray heat transfer using a minimum spray to each bundle in the General Electric (GE) Emergency Core Cooling System (ECCS) Evaluation Model.
This results in a heat transfer 2
coef ficient of 1.5 Btu /hr-ft - F for core spray heat transfer, which is the minimum value specified in Appendix K to 10 CFR Part 50.
The Japanese data are not the first to show low spray flow to some sections of BWR cores.
Thus, as described above, the staff has previously considered the effect of low core spray flow to. individual channels on calculated peak clad temperature (PCT).
In our evaluation of NE00-20566 Amendment 3, " General Electric Company Analytical Model for Loss of-Coolant Analysis in Accordance O
Shoreham SSER #2.
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with 10 CFR 50 Appendix K - Effect of Steam Environment on Core Spray Distri-bution," it was concluded that minimum spray flow to any channel following a LOCA would not be less than half of the design flow that was demonstrated to be available by tests and calculations.
The tests and calculations did not include steam effects on nozzle spray patterns and flow rate.
Based on measurements of minimum bundle spray flow for each BWR size and type for one sparger only, in air, the minimum flow for BWR/2 through BWR/5 designs was calculated to be 1.3 times the flow necessary to remove decay heat by vapori-zation (reference flow).
Thus, the steam effects on spray distribution would not result in less than 0.65 times the minimum reference flow (or 1.3 times with both spray spargers operating).
BWR FLECHT data from APED-5529, " Core Spray and Core Flooding Heat Transfer Effectiveness in a Full-Scale Boiling Water. Reactor Bundle," June 1979, show little degradation in heat transfer for
. flow as low as 0.38 times the reference flow, or approximately one gpm.
As far as we have been told, the minimum flow observed for any bundle in the Japanese 60 sector tests was one gpm.
The heat transfer coefficients in GE's ECCS Evaluation Model are based on the FLECHT data, and a minimum bundle flow of onegpmwouldjustifytheheattransfercoefficientforcorespraycooling (1.5 Btu /hr-ft - F) used in that Model.
During the BWR core spray injection, spray injected in the upper plenum will either be distributed to the core or bypass the core and drain to the lower plenum region, which results in a rapid bottom reflood rate.
Presently, credit is not taken for this rapid bottom reflood effect in the GE ECCS Model.
Any liquid in excess of the minimum required for core spray heat transfer is assumed lost from the system and does not contribute to the reflood.
Prelimi-nary results from the 30 SSTF Counter Current Flow Limiting (CCFL) tests performed in Lynn, Massachusetts show that the spray flow injected in the upper plenum actually drains to peripheral bundles and increases the bottom reflood l
rate.
In response to our request, GE presented results for reanalysis of limiting BWR/4 and BWR/5 cases to assess the effect of no core spray cooling on the peak clad temperature, assuming that the core spray coolant drains to the l
lower plenum and increases the reflood rate as observed in the Lynn tests.
The l
calculated peak clad temperature did not exceed the 10 CFR 50.46 limit of j
2200 F with no credit taken for the spray cooling effect.
i We conclude that the new preliminary and incomplete information from the Japanese tests does not pose a safety concern for BWR/4 and BWR/5 reactors, including the Shoreham reactor, for the following reasons:
(a) Core spray flow maldistributions resulting in flows on the order of one gpm per bundle (apparently consistent with those obtained in the Japanese 60 sector tests) would remain consistent with the core spray cooling assumptions employed in the present GE ECCS Evaluation Model.
(b) New analyses performed by GE have shown that for limiting BWR/4 and BWR/5 cases with core spray assumed to flow down peripheral channels to increase the reflood rate as observed'in the Lynn test, the calculated peak clad temperature did not exceed the 10 CFR 50.46 limit of 2200 F with no credit taken for the spray cooling effect.
Shoreham SSER #2 6-2
9 AUXILIARY SYSTEMS 9.4.1 Control Room Air Conditioning System In our Safety Evaluation Report, we stated that all control room air condition-ing system outside air intakes and exhausts are tornado missile protected.
By letter dated November 13, 1981, we were informed by the Shoreham Resident Inspector that the control room air conditioning system east air intake is not tornado protected and that all the piping from the east air intake is not in a tornado protected structure.
l The east air intake is a remote intake located in the radwaste building and therefore penetration of missiles via this air intake will not affect safe plant shutdown and will not prevent operation of the control room air condition-ing system since the air intake located in the control building is tornado missile protected.
Since only one air intake is necessary for operation of the control room air conditioning system, protection of the remote air intake is not required.
Many plants have only one air intake.
Therefore, our previous conclusion that the control room air conditioning system is acceptable remains unchanged.
9.5 Fire Protection System 9.5.3.2 Fire Doors and Dampers 1
In Supplement No. 1 to the Safety Evaluation Report, we stated that certain l
areas of the plant contained motorized 1 -hour fire dampers in which the l
motorized assembly, including cables, are not U.L. listed.
We were concerned I
that the unlisted assemblies would prevent the fire dampers from performing l
their function.
By letters dated September 25, 1981 and October 13, 1981, the applicant pro-vided additional information.
The installation has been modified to include solenoid and motor circuits approved by U.L.
As a result, we now conclude that the fire dampers, as modified, meet the design guidelines of Section D.1.j of Appendix A to BTP ASB 9.5-1, " Guidelines for Fire Protection for Nuclear Power Plants," and are, therefore, acceptable.
Based on our review, we conclude that the Shoreham fire protection program will meet the technical requirements of Appendix R to 10 CFR Part 50, when committed modifications have been completed, meets the guidelines of Appendix A to BTP. ASB 9.5-1, meets the requirements of General Design Criterion 3, and is, therefore, acceptable, e
Shoreham SSER #2 9-1
13 CONDUCT OF OPERATIONS 13.5.2 Operating and Maintenance Procedures A.
General A review has been conducted of the applicant's plan for development and imple-mentation of operating and maintenance procedures.
The review was conducted to determine the adequacy cf the applicant's program for assuring that routine operating, off-normal, and emergency activities are conducted in a safe manner.
The following description and evaluation is based on information contained in the applicant's Final Safety Analysis Report (FSAR) and the applicant's response to NUREG-0737, " Clarification of TMI Action Plan Requirements."
In determining the acceptability of the applicant's program, the criteria of NUREG-0800, Standard Review Plan, Section 13.5.2 were used.
The revi,ew consis-ted of an evaluation of (1) the applicant's procedure classification system for procedures that are performed by licensed operators in the control room, and the classification for other operating and maintenance procedures; (2) the l
applicant's plan for completion of operating and maintenance procedures during the initial plant testing phase to allow for correction prior to fuel loading; (3) the applicant's program for compliance with the guidance contained in Regulatory Guide 1.33, Rev. 2, March 1978 regarding the minimum procedural requirements for safety-related operations; (4) compliance with the guidance contained in ANSI N18.7-1976/ANS 3.2; and (5) the applicant's program for compliance with TMI Item I.C.1, " Guidance for the Evaluation and Development of Procedures for Transients and Accidents," for the development of Emergency Operating Procedure Guidelines.
B.
Operating and Maintenance Procedure Program The applicant has committed in the FSAR to a program in which all activities l
are to be conducted in accordance with detailed written an1 approved procedures meeting the requirements of Regulatory Guide 1.33, Rev. 2, March 1978, " Quality Assurance Program Requirements (0peration)," and ANSI N18.7-1976/ANS 3.2.
The l
applicant uses the following categories of procedures fcc those operations performed by licensed operators in the control room:
General Operating Procedures System Operating Procedures l
Emergency Operating Procedures Alarm Response Temporary Procedures Other procedures include the following areas:
Initial Test Maintenance Instrument and Control Systems Shoreham SSER #2 13-1
Surveillance Fmergency Plan Health Physics Chemistry Reactor Engineering
-Plant Security Radioactive Waste Management Our review disclosed that the applicant's program for use of operating and maintenance procedures meets the relevant requirements of 10 CFR 50.34, and is consistent with the guidance provided in Regulatory Guide 1.33 and ANSI N18.7-1976/ANS 3.2.
Therefore, we concluded that the applicant's program is acceptable.
C.
Reanalysis of Transients and Accidents; Development of Emergency Operating Procedures In letters of September 13 and 27, October 10 and 30, and November 9, 1979, the Office of Nuclear Reactor Regulation required Licensees of operating plants, applicants for operating licenses and licensees of plants under construction to perform analyses of transients and accidents, prepare emergency procedure guidelines, upgrade emergency procedures, and to conduct operator retraining (see also item I.A.2.1).
Emergency operating procedures are required to be consistent with the actions necessary to cope with the transients and accidents analyzed.
Analyses of transients and accidents were to be completed in early 1980 and implementation of procedures and retraining were to be completed three months after emergency procedure guidelines were established; however, some difficulty in completing these requirements has been experienced.
Clarifica-tion of tN scope of the task and appropriate schedule revisions were included in NUREG-0737, Item I.C.1.
Pending staff approval of the revised analysis and guidelines, the staff will continue the pilot monitoring of emergency operating procedures described in Task Action Plan Item I.C.8 (NUREG-0660).
The adequacy of the BWR Owners' Group Guidelines will be identified for each near term operating license (NT0L) during the emergency operating procedure review.
In a submittal dated June 30, 1980, the BWR Owners' Group provided a draft of the generic guidelines for Boiling Water Reactors.
The guidelines were developed to comply with Task Action Plan Item I.C.1(3) as clarified by NUREG-0737 and incorporated the requirements of short term reanalysis of small break loss-of-coolant accidents and inadequate core cooling (Task Action Plan Items I.C.1(1) and I.C.1(2)).
In a letter dated October 21, 1980, from D. G. Eisenhut to S. T. Rogers, the staff indicated that the generic guidelines prepared by General Electric and the BWR Owners' Group were acceptable for trial implementation at the Shoreham Nuclear Power Station.
Additional infor-mation was requested by the staff and was submitted by the Owners' Group on January 31, 1981.
This additional information is still under review prior to the staff making a final conclusion on the acceptability of the guidelines for implementation on all Boiling Water Reactors.
The guidelines are st'll considered acceptable for trial implementation'at the Shoreham Nuclear Power Station.
Based on our review of the emergency operating procedures developed from the BWR Owners' Group Guidelines and our observation of the procedures being implemented on a simulator and in a walk-through in the control room, we have Shoreham SSER #2 13-2
concluded that the guidelines have been adequately incorporated into the procedures.
This fulfills the requirements of Item I.C.1 of NUREG-0737.
In accordance with NUREG-0737, Item I.C.7, NSSS vendor review of the low power testing, power ascension testing, and emergency operating procedures is neces-sary to further verify adequacy of the procedures.
This requirement must be met before issuance of a full power license.
The NSSS vendor, General Electric Corporation, will review the startup tests and emergency operating procedures prior to these procedures being implemented.
The startup tests encompass the low power testing and the power ascension testing phases.
The applicant has committed to ensuring these reviews are complete prior to fuel load.
The staff must review the applicant's resolution of vendor comments to confirm vendor review and implementation of vendor comments into the procedures.
The staff will confirm that this review is completed prior to issuance of a full power license.
In accordance with NUREG-0737, Item I.C.8, correct emergency procedures as necessary based on the NRC audit of selected plant emergency operating pro-cedures (e.g., small-break LOCA, loss-of-feedwater, restart of engineered safety features following a loss of ac power and steam-line break).
This action will be completed prior to issuance of a full power license.
The staff and personnel froni ': sttelle Pacific Northwest Laboratories reviewed the procedures forwarded by the applicant to the NRC to ensure that the pro-cedures were consistent with the plant's design, the BWR Owner's Group guide-lines, and incorporated applicable human factors considerations.
The review resulted in two pages of general comments and numerous specific detailed comments on the procedures.
The general comments included human factors consideration on the use of standard logic format, procedure identification, interaction with reon-emergency procedures, inconsistency between emergency procedures and control room displays and the inadequacy of the graphs that were included in the procedures.
The specific comments include clarification and the locations of caution statements, the inclusion of action steps in cautions, the need for the addition of specific information to reduce operator judgments such as the preferred sequence for starting various systems, the need to add decision points to aid operator actions, and numerous references to changing words and using standard logic format to clarify action steps.
A meeting was held with the applicant on September 16, 1981, to discuss the results of the review.
During the meeting many of the comments were resolved by incorporating the recommended changes.
On October 16, 1981, a simulator exercise was held at the Limerick Training Center.
Operators used the revised emergency operating procedures to respond to simulated transients and accidents.
Scenarios were designed to require the concurrent use of procedures and transition among procedures.
The scenarios varied from minor transients to accidents involving multiple system failures.
The simulated transients and acidents included:
1)
Loss of feedwater from leaks or breaks in feed lines, faulty valve opera-tion, and pump failure.
Shoreham SSER #2 13-3
2)
Various initiating events followed by failure of various injection systems (e.g., RCIC, HPCI, LPCI) when needed for level control, level restoration and containment control.
3)
Turbine trip followed by a reactor trip.
4)
Failure of off-site power with subsequent failure of a diesel generator.
5)
Stuck open relief valves resulting in loss of Reactor Pressure Vessel Water inventory and emergency conditions in containment.
All of the emergency operating procedures were tested in responding to the simulations.
The review team observed the exercises and discussed them in detail with the operators.
Special emphasis was placed on the need to use written ee.orgency procedures and evaluating the clarity and usability of the procedures.
Several changes were made to the procedures as a result of the exercises and subsequent discussions.
The changes involved sequencing of steps, labeling to help locate specific steps, and clarifying priorities of actions.
On October 17, 1981, the team of reviewers that had participated in the simula-tor exercises conducted a walk-through of the emergency operating procedures in the control room.
The operators were presented with the initiating event (an intermediate-size break), with the desired sequence of steps.
The operators then walked through the scenario, while the team of reviewers evaluated the operators' use of the procedures, the interaction of the operators with the control panels, and the interaction between the operators.
The entire sequence was discussed in detail with the control room operators and the plant operations staff at the conclusion of the simulated event.
The effective. manner in which L
the operators used the emergency operating procedures indicates that they are clear, properly sequenced, and compatible with the control room and its equip-ment.
During the review, it was noted that:
- 1) some plant specific data were not available and noted by a "(Later)", 2) the graphs referenced in the procedures need revision to-improve their usability, and 3) there are a few additional changes required in the procedures as noted during the simulator exercises.
The applicant has committed to incorporate the plant specific data when they are available and to make the agreed to changes to the procedures and graphs.
The staff will verify that the missing data and changes have been included in j
the procedures before issuance of an operating license.
l Shoreham SSER #2 13-4
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4 15 ACCIDENT ANALYSIS 2
.15.3 Anticipated Transients Without Scram We stated in the Safety Evaluation Report that the applicant agreed to develop an emergency procedure for an ATWS event.
The Shoreham ATWS procedure was reviewed by members of the NRC staff and contractor personnel from Battelle Pacific Northwest Laboratories and comments were discussed with the operations personnel.
Based on our evaluation, we conclude that the Shoreham ATWS procedure provides an acceptable basis for licensing and interim operation of Shoreham pending the outcome of the proposed rulemaking on ATWS in accordance with General Design Criteria 10, 15, 26, 27, and 29 of 10 CFR Part 50 Appendix A.
The staff has recommended to the Commission that rulemaking be used to determine any future modifications necessary to resolve ATWS concerns and the required schedule for implementation of such modifications.
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Shoreham SSER #2 15-1
18 REVIEW BY THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS The Advisory Committee on Reactor Safeguards performed a review of the application for an operating license for the Shoreham Nuclear Power Station.
A Subcommittee toured the facility on April 30, 1981 and held a meeting in Washington, D.C. on September 30, 1981.
On Oct oer 15, 1981, the full Committee considered the Shoreham application 6. 'ts 258th meeting in Washington, D.C.
A copy of the Committee's report to Chairman Palladino is presented on the following pages.
We have considered the comments and recommendations made by the Committee in their report.
The Committee specifically mentioned LILC0's safety review committees, training program, and remote shutdown system.
The Committee also referred to LILC0's commitment to evaluate the final generic load definition of NUREG-0808 against the load specification used in the interim evaluation (NUREG-0487).
These matters are currently under review by the staff and the results of our review will be provided in a future supplement to the Safety Evaluation Report.
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Shoreham SSER #2 18-1
i p ut o
UNITED STATES g
NUCLEAR REGULATORY COMMISSION o
,E ADVISORY COMMITTEE ON REACTOH SAFEGUARDS 0,,
WASHINGTON, D. C. 20555
- ,g
,,s October 19, 1981 Honorable Nunzio J. Palladino Chaiman U.S. Nuclear Regulatory Commission Washington, DC 20555
SUBJECT:
REPORT ON THE SH0REHAM NUCLEAR POWER STATION UNIT 1
Dear Dr. Palladino:
During its 258th meeting, October 15-17, 1981, the Advisory Committee on Reactor Safeguards completed its review of the application of the Long Island Lighting Company (LILCO) for a license to operate the Shoreham Nuclear Power Station Unit 1.
A Subcommittee meeting was held in Washington, D.C. on Sept-ember 30, 1981 to consider this project. A tour of the facility was made by me'bers of the Subcommittee on April 30, 1981. During its review, the Com-mittee had the benefit of discussions with representatives of the Applicant and the NRC Staff. The Committee also had the benefit of the documents listed. The Committee reported on the construction permit application for this plant in a letter to AEC Chaiman Glenn T. Seaborg dated Decenber 18, 1969.
The Shoreham plant is located on Long Island in the Town of Brookhaven, Suffolk County, New York, about 55 miles east-northeast of downtown New York Ci ty.
It uses a GE BWR-4 nuclear steam supply system with a rated power level of 2436 MWt and a Mark II pressure suppression containment with a de-sign pressure of 48 psig. The Shoreham plant is one of three included in the Mark II Owners Group lead pisnt program. The NRC Staff has completed its review of the lead plant program and has issued NUREG-0487 and Supple-ments I and II, " Mark II Containment Lead Plant Program Load Evaluation and Acceptance Criteria." The NRC Staff has concluded that Shoreham satisfies these criteria.
In addition, LILCO has committed to evaluate the final ge-neric load definition of NUREG-0808, " Mark II Containment Program Load Eval-uation and Acceptance Criteria," against the load specification used in the interim evaluation (NUREG-0487).
Subject to satisfactory completion of this work, the NRC Staff has found the Shoreham containment acceptable. We concur in this finding.
LILCO described the management organization and the technical personnel available for operation of the Shoreham plant. Because of LILCO's lack of BWR operating experience, the NRC Staff is requiring that the control room staff and senior plant management be provided with advisors who have sub-stantial BWR operating experience. We concur with the NRC Staff that sup-plemental personnel experienced in BWR operation are needed until adequate operating experience is developed by the LILC0 staff.
18-2
f Dr. Nunzio J. Palladino October 19, 1981 LILCO described three safety review committees which will be a pemanent part of the Shoreham organization. We believe that these committees should include some expertise from sources outside LILCO's or its contractors' or-ganizations to provide balanced professional judgment on matters that could affect public health and safety. LILC0 should organize the planned safety review committees as soon as practical so they will have time to develop an understanding of plant related safety matters prior to plant operation.
LILCO also described its program and philosophy for training of personnel.
The initial training that the operations staff has received using a contractor-run training organization appears adequate. However, LILCO should establish an in-house training program to be maintained on a continuing basis so that operational skills are enhanced.
LILCO has initiated a Shoreham plant assessment based on probabilistic risk assessment techniques.
The Applicant's assessment effort in this area will provide a valuable addition to his operational knowledge.
An outstanding issue in the NRC Staff's Safety Evaluation Report dated April 1981 and Supplement 1 to that report dated September 9,1981 involves the re-mote shutdown system. The NRC Staff is concerned that a single, random fail-ure in the instruments and controls of systems controlled from the remote panel or in the systems themselves may prevent the remote shutdown panel from per-fonning its function. This item should be resolved in a manner satisfactory to the NRC Staff.
The Committee wishes to be kept informed.
The NRC Staff has identified other outstanding issues in its Safety Evaluation l
Report. We believe that these outstanding issues can be resolved and recom-mend that this be done in a manner satisfactory to the NRC Staff before op-eration at full power.
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We believe that if due consideration is given to the recommendations above, and subject to satisfactory completion of construction, staffing, and pre-operational testing, there is reasonable assurance that Shoreham Nuclear Power Station Unit I can be operated at power levels up to 2436 MWt without l
undue risk to the health and safety of the public.
Sincerely, J. Carson Mark l
Chairman
[
References:
1.
Long Island Lighting Company, "Shoreham Hu: lear Power Station Unit 1 Final Safety Analysis Report," Volumes 1-16 and Amendments 1-40.
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Dr. Nunzio J. Palladino October 19, 1981 2.
U.S. Nuclear Regulatory Commission " Safety Evaluation Report Related to the Operation of Shoreham Nuclear Power Station Unit 1," NUkEG-0420, dated April 1981 and Supplement No. I dated September 9,1981.
3.
U.S. Nuclear Regulatory Commis; ion, NUREG-0487, " Mark II Containment Lead Plant Program Load Evaluation and Acceptance Criteria," dated October 1978 with Supplements 1 and 2 dated September 1980 and Febru-ary 1981.
4.
U. S. Nuclear Regulatory Commission, NUREG-0808, " Mark II Containment Program Load Evaluation and Acceptance Criteria," dated August 1981 5.
Letter, SNRC-629, B. R. McCaffray, Manager, Long Island Lighting Company (LILCO), to Advisory Committee on Reactor Safeguards, regarding response to requests for information made at the ACRS Subcommittee meeting on Shoreham Station Unit 1 of September 30, 1981, dated October 13, 1981.
6.
Letters from J. P. Novarro, Project Manager, LILCO to Harold R. Denton, Director, Office of Nuclear Reactor Regulation (NRR), NRC, dated July 17, 1981; June 29,1981; June 15,1981; May 29,1981; May 29,1981; May 28, 1981; May 27,1981; May 27,1981; May 27,1981; May 21,1981; May 15, 1981; May 15,1981; May 15,1981; May 12,1981; April 22,1981; April 15, 1981; March 16,1981.
i 7.
Letters from B. R. McCaffrey, Manager, LILOO, to Harold R. Denton, Director, NRR, NRC dated August 18, 1981; August 7,1981; July 31,1981; July 23,1981; July 22,1981; July 21,1981; July 20,1981.
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4 18-4
22 TMI-2 REQUIREMENTS
'.. D.1 Control Room Design Review Discussion In Supplement No. 1 to the Safety Evaluation Report, we stated that a Human Factors Engineering Branch (HFEB) team conducted a Control Room Design Review / Audit (CRDR/A) at the Shoreham site between March 30 and April 3, 1981.
The review team identified many human engineering discrepencies (HEDs) which are documented in our CRDR/A report transmitted to the applicant on May 10, 1981.
The applicant reviewed these findings and proposed corrective actions or further study for each.
Representatives of the NRC's Region I office also reviewed the findings and proposed dispositions and determined that the applicant's responses to five of the findings were unacceptable.
Based on our consideration of the findings identified by Region I, we have revised our evaluations as set forth below:
CRDR/A Report Finding 1.2 The security console is unnecessary in the control room.
It obstructs movement and obstructs the view of the MXP panel from the operator's desk and the reactw control console.
LILCO's response was to examine the degree of interference between the security console and the HVAC and MXP panels during the Detailed Control Room Design Review (DCRDR) to be conducted in accordance with NUREG-0700,
" Guidelines for Control Room Design Reviews," issued September 1981.
The discrepancy addresses visual interference of the security console with the MXP panel.
This is not considered to be nearly as safety critical as having two totally unrelated functions monitored and controlled from the same general location.
We agree with Region I in that the control room operators and security operators will unavoidably interfere with each other in the normal performance of their jobs.
This is expected to be especially true during an abnormal condition of either function and may be emphasized with the proposed location of the Safety Parameter Display System (SPDS).
It is our position that the security console should be relocated outside of the controlled access space of the control room.
We require that LILC0 re-examine this condition and submit a proposed solution that will eliminate the possibility of any physical, visual,' or auditory interference by the security console or its operator on control room operations.
CRDR/A Report Findings 7.3 On Panels 601 and 602, the Rosemont SRV pressure indicators are too far from the SRV auto-depressurization controls to be read.
I Shoreham SSER #2 22-1
7.6 The Safety Valve Temperature Indicator / Recorder, which provides positive indications of open safety relief valves, is on a back panel behind Panel 601.
7.7 The safety relief valves located on Panel 602 have corresponding annunciator tiles located remotely on annunciator panel G on Panel MCB.
LILC0 proposed no action on 7.3 and 7.6 and we previously recommended they be re-examined during the DCRDR.
On item 7.7 LILC0 agreed to relocate controls using the guidance of NUREG-0700.
The SRV/ ADS controls, displays, recorders and annunciators are currently scattered on panels 601, 602, 614, MCB, and MXB. We consider the safety significance of the individual discrepencies found in our review to be much less than the scattered layout of the overall system.
LILCO has agreed to relocate controls to solve the control / annunciator discrepency in item 7.7.
We believe that relocation of only part of the system components would not solve the major problem and that LILC0 should consider integration of the complete functional group and propose a new layout design for our review and approval.
CRDR/A Report Finding 9.16 Use of flashing yellow "-99.9" to indicate that data should be ignored is confusing.
LILCO's response was that the use of flashing yellow "-99.9", to indicate an unknown analog value, has been reviewed and approved by operations personnel.
This convention is also consistent with G.E. Software.
Multiple unknown analog values on most displays during the audit was due to the fact that a large number of analog points are currently awaiting checkout by startup personnel.
During plant operation the occurrence of unknown values will be infrequent.
No additional LILC0 action will be taken.
The following information has been received since this item was closed in SER Supplement No. 1:
(1)
"-99.9" is a legi'4 mate value (e.g., reactor water level) which can appear on the CRT under normal circumstances.
(2) Legitimate normal numeric values can appear on the CRT in the color yellow.
(3) Flashing of a numeric value on the CRT indicates that the value is out of the normal range.
Thus a flashing yellow "-99.9" does not have a unique meaning and could be misinterpreted, especially when its occurrence will be infrequent.
It-is our position that LILC0 re-examine this problem and develop a new solution for our review and approval.
Shoreham SSER #2 22-2
On page C-1 of SER Supplement No. 1, we stated that LILC0's response to Item 1.1 was, "The Watch Engineers' office will be completed prior to loading fuel.
Visual and voice contact will be verified."
In a letter from l
J. P. Novarro to Harold R. Denton dated June 11, 1981 (SNRC-585), the applicant deleted the last sentence in its response to Item 1.1.
We find this revision unacceptable and require the applicant to verify visual and-voice contact prior to fuel load, using.the guidance of NUREG-0700.
Conclusion We believe that adequate solutions to the findings described in this report are necessary to improve the safety of the Shoreham Nuclear Power Static-control room.
The applicant shall reevaluate the findings and submit proposed corrective actions and a schedule for implementing the actions for our review and approval 120 days prior to issuance of the operating license.
Approved corrective actions will be required to be implemented prior to exceeding 5 percent power operations.
We conclude that with the implementation of the corrective actions described in SER Supplement No. 1 and with acceptable solutions to the findings of this report, the potential for operator error leading to serious consequences as a result of human factors considerations in the control room will be sufficiently low to permit safe operation of the Shoreham Nuclear Power
. Station.
II.B.4 Degraded Core Training Discussion and Conclusion The applicant committed in its letter dated October 2, 1981, to implement a degraded' core training course.
The course covers 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> of training to control or mitigate an accident in which the core is severely damaged.
This material will be incorporated in the FSAR by an amendment from the applicant.
This course will be given to all shift technical advisors and. operations l
personnel, from the Plant Manager to and including licensed operators, prior to fuel loading.
t Managers and technicians in instrumentation and controls, health physics, and chemistry, will be given training ccmmensurate with their responsibilities during accidents which involve severe core damage.
i Based on this commitment, we have concluded that'the Long Island Lighting Company has met our requirements for training personnel in the use of installed plant systems to control or mitigate an accident in which the core is severely damaged.
This. training program will be in general agreement with the " Training Guidelines for Recognizing and Mitigating the Consequences of Severe Core Damage" from the Institute of Nuclear Power Operations, Document Number STG-01, Revision 1, dated January 15, 1981.
The applicant'has committed.to' complete the training of al.1 operating personnel, prior to fuel load, in the use of installed systems to monitor and control accidents in which the core may be severely damaged.
- Shoreham SSER #2 22-3
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II.F.1 Additional Accident-Monitoring Instrumentation, Sampling and Analysis of Plant Effluents In Supplement No. 1 of the Safety Evaluation Report, we stated that the applicant's procedures to sample the station vent effluent for radioiodines and particulates indirectly (by separately sampling the inputs from the tur-bine and the radwaste buildings to the station vent) during those accidents when the sampling media is inaccessible would be acceptable only if the applicant provides adequate justification for his contention that no accessible location for the sampling media is feasible, consistent with the requirements of NUREG-0737 and ANSI N 13.1-1969.
By submittal dated January 7, 1982, the applicant committed to install the post-accident mid/high level. station vent exhaust monitor and associated sampling provisions in the turbine building.
The instrumentation will be accessible during an accident and will provide direct sampling capability for the station vent exhaust.
The instrumentation will include provisions for isokinetic sampling and compliance with ANSI N 13.1-1969 to assure representa-tive sampling.
The installation and calibration of the monitor and sampler will be completed prior to fuel loading.
Based on our review of the above submittal, we conclude that the applicant has provided an acceptable method to continuously sample the station vent effluent for radioiodines and particulates during an accident.
A post-implementation review of the installed system, detailed drawings, and procedures for system operation will be performed., Containment Water Level Monitor We stated in Supplement No. 1 to the Safety Evaluation Report that the water level monitor proposed by the applicant did not have the necessary range.
The applicant has subsequently provided a water level monitor design which has an adequate range. We require this instrument to be installed and functional before fuel load.
Subject to applicant's compliance with the above requirement, we find this issue to be resolved.
Shoreham SSER #2 22-4
APPENDIX A ERRATA TO THE SAFETY EVALUATION REPORT Page 4-22, Line 12 Delete Technical specifications will be modified to require that the core flow be checked at least every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the average power range monitor flow biased scram be recalibrated at least oace per month.
and replace with Technical Specifications will require that the core flow be checked every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Page 4-22, Line 41 Following day insert a period and delete the remainder of the sentence.
Page 12-5, Line 25 Following monitors insert a period and delete the remainder of the sentence.
Page 15-7, Line 19 Delete the s from stripping.
Page 15-7, Line 22 Make Source plural.
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Shoreham SSER #2 A-1
APPENDIX B ERRATA TO SER SUPPLEMENT NO. 1 Page 9-7, Line 19
- elete all fire and replace with these.
Page 9-10, Line 28' -
Delete the period and insert for which credit.is taken in the shutdown scenario associated with each fire segment.
Page 9-11, Line 19 Delete objective and replace with capability.
Page 9-11, Line 20 Delete in event of and replace with by providing controls and instrumentation which can be isolated from.
Page 13-38, Line 34 -
Delete RCO and replace with ROC.
Page 13-38, Line 39 -
Add Reactor Engineer.
Page 22-68, Line 41 -
Delete 0373 and replace with 0737.
Page C-26, Line 2 Delete Labels will be located on the recorder to provide clear readability to the operator and replace with Labels located below controls & recorders will i
be reevaluated for adequacy prior to fuel load.
Changes necessitated as a result of the reevaluation will be effected prior to fuel load.
I I
1 Shoreham SSER #2 B-1
NRC roRu 335
- 1. REPORT NUMBER (Assegrwaby DDC)
U.S. NUCLEAR REGULATORY COMMISSION o,,,
NUREG-0420 BIBLIOGRAPHIC DATA SHEET Supplement No. 2
- 0. TITLE AND SUBTITLE lAdd Voturne No.. st appeconatel
- 2. (Leave blankt Safety Evaluation Report Related to the Operation of the Shoreham Nuclear Power Station, Unit No.1
- 3. RECIPIENT 3 ACCESSION NO.
- 7. AUTHORG)
- 5. D ATE REPORT COMPLE TED M ON TH l YEAR Februa ry 1982 9 PE RFORMING ORGANIZATION N AME AND MAILING ADDRESS (inclue l'en Codel DATE REPORT ISSUED U. S. Nuclear Regulatory Commission uoN7" lvEAR Office of Nuclear Reactor Regulation Februa ry 1982 Washington, D. C.
20555 6 (Leave b< anal
- 8. (Leaveplank)
- 12. SPONSORING ORGANIZATION NAME AND M AILING ADDRESS (include 2,p Codel p
Same as 9 above
- 11. CONTRACT NO.
- 13. TYPE OF REPORT PE RIOD COV E RE D (Inclus,ve datesl Safety Evaluation Report
- 15. SUPPLEMENTARY NOTES 14 (Leave o/ mal Pertains to Docket No. 50-322
- 16. ABSTR ACT C00 words or less)
Su ppl emei. No. 2 to the Safety Evaluation Report of Long icknd Lighting Company's application for a license to operate the Shoreham Sclear Power Station, Unit 1, located in Suffolk County, New York, has been prepared by the Office of Nuclear Reactor Regulation of the U. S. Nuclear Regulatory Commission. This supplement reports the status of certain items that has not been resolved at the time of publication of the Safety Evaluation Report.
17 KEY WORDS AND DOCUMENT AN ALYSIS 17a. DESCRiPTORS 17n. IDENTtrtE RS/OPEN ENDED TE RMS 18 AV AILABILITY STATEMENT 19 SECURITY CLASS (Th,s reporrl 21 NO OF PAGES Unlimited
_U_nC1 " M i f kd 20 SECURITY CL ASS f Thes papI
- 22. PReCE S
PdRC F ORM335 67 771 l
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