ML20151W259
| ML20151W259 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 07/31/1988 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-0822, NUREG-0822-S01, NUREG-822, NUREG-822-S1, NUDOCS 8808230426 | |
| Download: ML20151W259 (70) | |
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NUREG-0822 Supplement No.1 1
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- Integrated Plant Safety Assessment
! Systematic Evaluation Program I Oyster Creek Nuclear Generating Station GPU Nuclear Corporation and Jcrsey Central Power & Light Company i
Docket No. 50-219 i
U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation July 1988 p ' ' %,,
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NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available from one of the following sources:
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- 1. The NRC Public Document Room,1717 H Street, N.W.
Washington, DC 20555
- 2. The Superintendent of Documents, U.S. Government Printing Of fice, Post Of fice Box 37082, Washington, DC 20013 7082
- 3. The National Technical Information Service, Springfield, VA 22161 Although the listing that follows represents the majority of documents cited in NRC publications, it is not intended to be exhaustive.
Referenced documents available for inspection 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 and investigation notices:
Licensee Event Reports; vendor reports and correspondence; Commission papers; and applicant and licensee documents and correspondence.
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NUREG 0822 Supplement No.1 c:::
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Integrated Plant Safety Assessment Systematic Evaluation Program Oyster Creek Nuclear Generating Station GPU Nuclear Corporation and Jersey Central Power & Light Company Docket No. 50-219 U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation July 1988 f.= a.,,,
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ABSTRACT The U.S. Nuclear Regulatory Commission (NRC) has prepared Supplement I to the final Integrated Plant Safety Assessment Report (IPSAR) (NUREG-0822), under the scope of the Systematic Evaluation Program (SEP), for the Oyster Creek Nuclear Generating Station, located in Ocean County, New Jersey, and operated by GPU Nuclear Corporation and Jersey Central Power and Light Company (colicensees).
The SEP was initiated by the NRC to review the design of older operating nuclear power plants to reconfirm and document their safety.
This report documents the review completed under SEP for those issues that required refined engineering evaluations or the continuation of ongoing evaluations subsequent to issuing the Final IPSAR for the Cyster Creek plant.
The review has provided for (1) an assessment of the significance of differences between current technical pos,itions on selected safety issues and those that existed when the Oyster Creek plant was licensed, (2) a basis for deciding how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety.
The final IPSAR and its supplement will form part of the bases for considering the conversion of the existing provi-sional operating license to a full-term operating license, iii
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TABLE OF CONTENTS P_ag ABSTRACT..............................................................
iii ACRONYMS AND INITIALISMS..............................................
xi 1
INTRODUCTION......................................................
1-1 2
TOPICS THAT REQUIRED REFINED ENGINEERING ANALYSIS OR CONTINUATION OF ONG0ING EVALUATION................................
2-1 2.1 Topic II-3.B, Flooding Potential and Protection Requirements; Topic II-3.B.1, Capability of Operating Plant To Cope With Design-Basis Flooding Conditions; Topic II-3.C, Safety-Related Water Supply (Ultimate Heat Sink (UHS)) (NUREG-0822, Section 4.1).................................................
2-1 2.1.1 Condensate Water Pumps (NUREG-0822, Section 4.1(1))...
2-2 2.1.2 Canal Water Level Instrumentation (NUREG-0822, S e c t i o n 4.1 ( 3 ) ).......................................2-2 2.1.3 Protection During Internal Floodin Section 4.1(7))..................g (NUREG-0822, 2-3 2.2 Topic III-1, Classification of Structures, Components, and Systems (Seismic and Quality) (NUREG-0822, Section 4.2)......
2-3 2.3 Topic III-2, Wind and Tornado Loadings (NUREG-0822, Section 4.3).................................................
2-4 2.3.1 Reactor Building Steel Structure Above the Operatin Floor (NUREG-0822, Section 4. 3.1)................. g 2-4 2.3.2 Ventilation Stack (NUREG-0822, Section 4.3.2).........
2-5 2.3.3 Effects of Failure of Nonseismic Category I Structures (NUREG-0822, Section 4.3.3)...........................
2-5 2.3.4 Roof Decks (NUREG-0822, Section 4.3.6)................
2-6 2.3.5 Intake Structure, Oil Tanks, and Diesel Generator Building (NUREG-0822, Section 4.3.7)..................
2-6 2.3.6 Load Combinations (NUREG-0822, Section 4.3.8).........
2-7 2.3.7 Control Room (NUREG-0822, Section 4.3)................
2-7 2.3.8 Architectural Components (NUREG-0822, Section 4.3)....
2-7 2.4 Topic. III-3. A. Effects of Hi h Water Level on Structures 0
(NUREG-0822 Section 4.4)....................................
2-8 2.4.1 Hydrostatic Loads (Short-Duration) (NUREG-0822, Section 4.4(2)).......................................
2-8 2.5 Topic III-4.A. Tornado Missiles (NUREG-0822, Section 4.6)....
2-8 2.5.1 Emergency Diesel Generators and Fuel Oil Day Tank (NUREG-0822, Section 4.6.1)...........................
2-8 l
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TABLE OF CONTENTS (Continued) 1 fagg 2.5.2 Mechanical Equipment Access Area (NUREG 0822, Section 4.6.2).......................................
2-8 2.5.3 Condensate Storage Tank, Torus Water Storage Tank, and Service Water and Emergency Service Water Pumps (NUREG-0822,Section4.6.4)...........................
2-9 2.6 Topic III-4.8, Turbine Missiles (NUREG-0822, Section 4.7)....
2-9
- 2. 7 Topic 111-4.0, Site-Proximity Missiles (Including Aircraft) 2-10 (NUREG-0822. Section 4.8)....................................
4 2.7.1 Aircraft Hazards (NUREG-0822, Section 4.6.2)..........
2-10 l
2.8 Topic III-5.B. Pipe Break Outside Containment (NUREG-0822, 2-11 Section 4.10)................................................
J 2.8.1 Emergency Condenser Isolation (NUREG-0822, i
Section 4.10(2))......................,...............
2-11 i
2.9 Topic III-6, Seismic Design Considerations (NUREG-0822 j
Section 4.11).........................................
2-12 i
i 2.9.1 Piping Systems (NUREG-0822. Section 4.11(1))..........
2-12 2.9.2 Mechanical Equipment (NtlREG-0822, Section 4.11(2))....
2 13 i
2.9.3 Electrical Equipment (NUREG-0822, Section 4.11(3))....
2-13 t
1 2.9.4 Qualification of Cable Trays (NUREG-0822, Section 4.11(5))......................................
2-14 1
2.10 Topic III-7.B. Design Codes, Design Criteria, Load Combina-tions, and Reactor Cavity Design Criteria (NUREG-0822, 2-14 Section 4.12)................................................
2.11 Topic III-10.A Thermal-Overload Protection for Motors of Motor-Operated Valves (NUREG-0822, Section 4.14).............
2 15 1
2.11.1 Thermal-Overload Bypass (NUREG-0822, Section 4.14(1)).
2-15 i
2.12 Topic V-5, Reactor Coolant Pressure Boundary (RCPB) Leakage Detection (NUREG-0822, Section 4.16).........................
2-15 i
2.12.1 Leakage Detection Systems (NUREG-0822, Section 4.16.1).......................................
2-16 i
l 2.13 Topic V-11.A, Requirements for Isolation of High-and Low-Pressure Systems (NUREG-0822, Section 4.19)..............
2-17 2.14 Topic VI-4, Containment Isolation System (NUREG-0822, 2-18
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Section 4.22)................................................
2.14.1 Remote Manual Valves (NUREG-0822, Section 4.22.2).....
2-18 2.15 Topic VII-1.A. Isolation of Reactor Protection System From f
Non-Safety Systems, including Qualification of Isolation 2-18 I
Devices (NUREG-0822 Section 4.27)...........................
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TABLE OF CONTENTS (Continued)
P,agg 2.15.1 Flux Monitoring Isolation (NUREG-0822, Section 4.27(1))......................................
2-18 2.16 Topic VII-1.8, Trip Uncertainty and Setpoint Analysis Review of Operating Data Base (NUREG-0822, Section 4.28).....
2 19 2.17 Topic VIII-3.8, DC Power System Bus Voltage Monitoring and Annunciation (NUREG-0822, Section 4.32)..................
2 JU 2.18 Topic IX-5, Ventilation Systems (NUREG-0822, Section 4.34)...
2 z1 2.18.1 Core Spray and Containment Spray Pump Ventilation (NUREG-0822, Section 4.34(3)).........................
2-21 2.18.2 Battery, Motor Generator, and Switchgear Room Ventilation (NUREG-0822, Section 4.34(4)).............
2-21 3
TOPICS RESOLVED BY CHANGES TO PLANT TECHNICAL SPECIFICATIONS OP PROCEDURES........................................................
3-1 3.1 Topic 11-3.B. Flooding Potential and Protection Requirements; Topic II-3.B.1, Capability of Operating Plants To Cope With Design-Basis Flooding Conditions; Topic 11-3.C, Safety-Related Water Supply (Ultimate Meat Sink (UHS)) (NUREG-0822, Section 4.1).................................................
3-1 3.1.1 Isolation Condenser Flooding (NUREG-0822 Section 4.1(4)).........................,..............
3-1 3.1.2 Low Water Level Shutdown (NUREG-0822, Section 4.1(5)).
3-2 3.2 Topic V-5, Reactor Coolant Pressure Boundary (RCIB) Leaka Detection (NUREG-0822, Section 4.16)....................ge 3-3 3.2.1 Operability Requirements (NUREG-0822, Section 4.16.2).
3-3 3.3 Topic V-6, Reactor Vessel Inte Section 4.17)............... grity (NUREG-0822, 3-4 3.4 Topic V-12.A, Water Purity of BWR Primary Coolant (NUREG-0822 Section 4.20)...................................
3-4 3.5 Topic VI-7.A.3, Emergency Core Cooling System Actuation System (NUREG-0822, Section 4.23)............................
3-5 3.6 Topic VI-10.A, Testing of Reactor Trip and Engineered Srfety Features, Including Response-Time Testing (NUREG-0822, Section 4.20)................................................
3-6 3.6.1 Instrumentation for Reactor Trip System (RTS) Testin (NUREG-0022, Section 4.26.2).......................g 3-6 3.7 Topic IX-5, Ventilation Systems (NUREG-0822, Section 4.34)...
3-7 3.7.1 Restoration of Ventilation (NUREG-0822 Section 4.34(1))......................,................
3-7 vii
TABLE OF CONTENTS (Continued)
Pag _e 3.8 Topic XV-16, Radiological Consequences of Failure of Small Lines Carrying Primary Coolant Outside Containment (NUREG-0822, Section 4.36)...................................
3-7 Topic XV-19, Loss-of-Coolant Accidents Resulting From Spectrum of Postulated Pipe Breaks Within the Reactor Coolant Pressure Boundary (NUREG-0822, Section 4.38).........
3-8' aAR TOPIC RESOLUTIONS CONFIRMED BY NRC REGION I 0FFICE..........
4-1 4.1 Topic II-3.B, Flooding Potential and Protection Requirements; Topic II-3.B.1, Capability of Operating Plants To Cope With 4
Design-Basis Flooding Conditions; Topic II-3.C, Safety-Related Water Supply (Ultimate Heat Sink (UHS)) (NUREG-0822, Section 4.1).................................................
4-1 i
4.1.1 Isolation Condenser Flooding (NUREG-0822, Section 4.1(4)).......................................
4-1 4.1.2 Hurricane Flooding of Pumps (NUREG-0822, Section 4.1(6)).......................................
4-2 4.1.3 Roof Drains (NUREG-0822, Section 4.1(9))..............
4-2 4.2 Topic III-3.C, Inservice Inspection of Water Control Structures (NUREG-0822, Section 4.5).........................
4-2 4.2.1 Intake Structure Trash Racks and Intake Screens.
(NUREG-0822, Section 4.5.2)...........................
4-2 4.2.2 Inspection Program (NUREG-0822, Section 4.5.4)........
4-3
- 4. 3 Topic V-5, Reactor Coolant Pressure Boundary (RCPB) Leakage Detection (NUREG-0822, Section 4.16)..........................
4-3 4.3.1 Operability Requirements (NUREG-0822, Section 4.16.2).
4-3 4.4 Topic V-10.8, Residual Heat Removal System Reliability (NUREG-0822, Section 4.18)...................................
4-5 4.5 Topic V-12.A, Water Purity of BWR Primary Coolant (NUREG-0822, Section 4.20)..................................
4-5 4.6 Topic VI-1, Organic Materials and Postaccident Chemistry (NUREG-0822, Section 4.21)...................................
4-6 4.6.1 Organic Materials (NUREG-0822, Section 4.21.1)........
4-6 4.7 Topic VI-4, Containment Isolation System (NUREG-0822, Section 4.22)................................................
4-6 4.7.1 Locked-Closed Valves (NUREG-0822, Section 4.22.1).....
4-6 4.8 Topic VI-7.C.1, Appendix K - Electrical Instrumentation and Control Re-Reviews (NUREG-0822, Section 4.25)............
4-6 viii
TABLE OF CONTENTS (Continued) 4.8.1 AC Automatic Bus Transfers (NUREG-0822, Section 4.25(1))......................................
4-7 i
4.9 Topic VII-1.A, Isolation of Reactor Protection System From Non-Safety Systems, Including Qualification of Isolation Devices (NUREG-0822, Section 4.27)...........................
4-7 4.9.1 Reactor Protection System (RPS) Protective Trip O!UREG-0822, Section 4. 27(2)).........................
4-7 4.10 Topic VII-3, Systems Required For Safe Shutdown (NUREG-0822, Section 4.30)................................................
4-7 4.11 Topic VIII-2, Onsite Emergency Power Systems (Diesel Generator) (NUREG-0822, Section 4.31)........................
4-8 4.11.1 Diesel Generator Annunciators (NUREG-0822, Section 4.31(1))......................................
4-8 4.11.2 Diesel Generator Trip Bypass (NUREG-0822, Section 4.31(2))......................................
4-8 4.12 Topic VIII-3.B, DC Power Systems Bus Voltage Monitoring and Annunciation (NUREG-0822, Section 4.32)..................
4-8 4.13 Topic IX-5, Ventilation Systems (NUREG-0822, Section 4.34)................................................
4-9 4.13.1 Restoration of Ventilation (NUREG-0822, Section 4.34(1))......................................
4-9 5
REFERENCES.......................................................
5-1 APPENDIX NRC STAFF CONTRIBUTORS AND CONSULTANTS TABLE 2.1 Integrated Assessment Sunnary.....................................
2-23 2
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m ACRONYMS AND INITIALISMS
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APGRMS airborne particulate and gaseous radiation monitoring system APRM average power range monitor ASME American Society of Mechanical Engineers BTP branch technical position BWR boiling-water reactor CFR Code of Federal Regulations ESF engineered safety feature (s)
FSAR final safety analysis report FTOL full-term operating 'icense G0C general design criterion (a)
GE General Electric gpm gallons per minute HELB high energy line break HVAC heating, ventilating, and air conditioning IC isolation condenser IEEE Institute of Electrical and Electronics Engineers IPSAR integrated plant safety assessment report IRM intermediate range monitor LAI licensing actica item LC0 limiting condition (s) for operation LOCA loss-of-coolant accident LPZ low population zone MCC motor control center mph mile (s) per hour MSIV main steam line isolation valve MSL mean sea level NMPC Niagara Mohawk Power Corporation NRC U.S. Nuclear Regulatory Commission PETA plant engineering task assignment PMH probable maximum hurricane POL provisional operating license RCPB reactor coolant pressure boundary RPS reactor protection system RTS reactor trip system xi
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scfh standard cubic feet per hour.
SEP Systematic Evaluation Program
-SER safety evaluation report SRP Standard Review Plan SQUG Seismic Qualification Utility Group TS technical specification (s)
TSSIIL technical specification supporting installed instrumentation list UHS-ultimate heat sink USI unresolved safety issue s
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INTEGRATED PLANT SAFETY ASSESSMENT REPORT SUPPLEMENT NO. 1 l
SYSTEMATIC EVALUATION PROGRAM OYSTER CREEK NUCLEAR GENERATING STATION 1 INTRODUCTION The Systematic Evaluation Program (SEP) was initiated by the U.S. Nuclear Regulatory Commission (NRC) to review the designs of older operating nuclear power plants in order to reconfirm and document their safety.
The review pro-vides (1) an assessment of the significance of differences between current technical positions on safety issues and those that existed when a particular plant was licensed, (2) a basis for deciding how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety.
The results of the SEP review of the Oyster Creek plant were published in NUREG-0822, the Final Integrated Plant Safety Assessment Report (IPSAR), dated January 1983.
The review compared the as-built plant design with current review criteria in 137 different areas defined as "topics." During the review, 54 of the topics were deleted from consideration in the SEP because a review was being conducted under other programs (unresolved safety issues or Three Mile Island Action Plan tasks), the topic was not applicable to the Oyster Creek plant, or the items to be reviewed under that topic did not exist at the site, l
Of the original 137 topics, 83 were, therefore, reviewed for Oyster Creek; of these 43 met current criteria or were acceptable on another defined basis.
From the review of the 40 remaining topics, certain aspects of plant design were found to differ from current criteria.
These 40 topics were considered in the integrated assessment of the plant, which consisted of evaluating the safety significance and other factors of the identified differences from current design to arrive at decisions on whether modification was necessary from an over-all plant safety viewpoint.
To arrive at these decisions, engineering judgment was used as well as the results of a limited probabilistic risk assessment study.
In general, the staff's positions in the integrated assessment fell into one or more of the following categories:
(1) equipment modification or addition, (2) procedure development or Technical Specification changes, (3) refined engineering analysis or continuation of ongoing evaluation, and (4) no modifi-cation necessary.
Table 4.1 of the IPSAR summarizes the staff's integrated assessment positions and documents the licensee's agreement with these positions.
For those positions classified as either Category (1) or (2), the IPSAR lists the scheduled completion dates agreed upon by the staff and the licensee.
Region I has verified or is verifying the implementation of these positions.
For those positions classified as Category (3), the licensee has provided the results of the ongoing evaluation to the staff for review.
The purpose of this 1-1
supplement to the IPSAR is to provide the staff's evaluation of the Category (3) issues and to summarize the status of all actions to be implemented as a result of the "EP review.
The Oyster Creek plant is one of the four SEP plants that has not received a full-term operating license (FTOL).
Therefore, a safety evaluation report (SCR) to support the conversion of the provisional operating license (POL) to i
an FTOL will be prepared.
The SER will consist of the IPSAR, the IPSAR supple-ment, a consideration of major plant modifications that have been made and substantive regulations adopted since the POL was issued, and the unresolved safety issues and Three Mile Island Action Plan issues.
In this supplement, Section 2 provides the topics that required refined engi-neering analysis or the continuation of ongoing evaluation.
Section 3 provides the topics that were resolved by changes to plant Technical Specifications or proc dures.
Section 4 provides the resolutions of IPSAR topics that were con-firmed by the NRC Region I office.
Section 5 provides the references, o!Ser than NRC documents or correspondence between the NRC staff and the licensee or others, cited in this supplement.
The appendix lists the NRC staff contributors and consultants.
1-2
2 TOPICS THAT REQUIRED REFINED ENGINEERING ANALYSIS OR CONTINUATION OF ONG0ING EVALUATION The licensee has submitted an evaluation for each of the issues that required refined engineering analysis or further evaluation.
The staff reviewed these submittals and concluded that either the licensee met current criteria, the evaluation was acceptable on another defined basis, or corrective action or further analysis would be required.
Factors considered in reaching this con-clusion include the perceived safety significance of the difference from cur-rent licensing criteria, a qualitative assessment of the financial and exposure costs to make a modification, and, to a lesser extent, implemenution impact and schedule.
Also considered were any applicable risk perspectives, developed for the integrated assessment and described in the IPSAR, and related corrective actions proposed by the licensee as part of the integrated assessment or as a result of the subsequent evaluations.
A brief discussion of each of the outstanding issues is presented in the follow-ing sections.
Each evaluation references the more detailed license evaluation and staff topic evaluation.
The status of each of these issues is summarized in Table 2.1 along with the status of all SEP issues for the Oyster Creek Nuclear Generating Station.
2.1 Topic II-3.B Flooding Potential and Protection Requirements; Topic II-3.B.1, Capability of Operating Plants To Cope With Design-Basis Flooding Conditions; Topic II-3.C, Safety-Related Water Supply (Ultimate Heat Sink (VHS)) (NUREG-0822, Section 4.1)
General Design Criterion (GDC) 2 in Title 10, Pact 50 of the Code of Federal Regulations (10 CFR 50) as implemented by Sections 2.4.2, 2.4.5, 2.4.10, and 2.4.11 of the Standard Review Plan (SRP, NUREG-0800) and NRC Regulatory Guides 1.59 and 1.27, requires that structures, systems and components important to safety be designed to withstand the effects of natural phenomena such as flooding.
The safety objective of these topics (II-3.8, 11-3.8.1, and II-3.C) is to verify that operating procedures and/or system design provided to cope with the design-basis flood are adequate.
The site grade elevation is 23 feet mean sea level (MSL).
During its review of
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the hydrology-related topics, the staff identified the following flooding eleva-tions as defined by current licensing criteria:
l probable maximum hurricane - 22 feet MSL probable maximum precipitation - 23.5 feet MSL l
1 As a result of these flooding levels, the staff identified nine issues pertaining to the following in the IPSAR:
(1) condensate transfer pumps, (2) plant operat-ing limits on canal water level in the Oyster Creek Technical Specifications l
(TS), (3) canal water level instrumentation, (4) makeup isolation condenser water sources, (5) plant operating limits in the TS on water level at the ser-vice water intake, (6) procedures for a flood, (7) protection during internal 2-1
flooding, (8) hydrostatic loads on buildings, and (9) reactor and turbine build-ing parapets and scuppers, Issues (2), (4), (6), and (9) were resolved by commitments made by the licensee for specific plant modifications or changes in plant procedures.
These are discussed in Section 4.
Issue (8) is discussed in Section 2.4.
Issue (5) is discussed in Section 3.1.
Issues (1), (3), and (7) are discussed in Sections 2.1.1, 2.1.2, and 2.1.3, respectively.
2.1.1 Condensate Water Pumps (NUREG-0822. Section 4.1(1))
In Section 4.1(1) of the IPSAR, the staff concluded that two condensate transfer pumps are essential to charge the emergency condenser with cooling water during a hurricane-induced flood.
Because the motors of both of these pumps are powered from the same engineered safety features bus, a single failure of the power bus would disable both condensate transfer pumps.
The staff also stated that, in conjunction with the resolution of Topic III-4.A (see Section 4.6.4 of the IPSAR), the licensee had committed to provide a port-able pump to provide cooling water in the event of a loss of cooling resulting from tornado-missile damage.
The staff concluded that this diverse means of cooling was sufficient to alleviate the need for redundant power for the conden-sate transfer pumps.
Therefore, backfitting was not recommended.
In a letter dated July 3, 1985, the licensee proposed to use a main core spray pump to supply the isolation condenser.
This would be accomplished by connec-ting a temporary hose to one of the core spray system loops and routing the hose to the isolation condenser.
Both the water supply (suppression chamber) and the components would be protected from potential tornado missiles and ex-ternal flooding.
In a letter dated August 14, 1987, the licensee stated that through a detailed field walkdown and line-loss analysis of an existing system interconnection between the core spray and condensate and demineralized water transfer systems, it was determined that the existing plant configuration en-sures that makeup water can be supplied to the isolation condenser.
The staff has not completed its review regarding this matter. When it completes its re-view, the staff will document the results in a supplement to the IPSAR.
2.1.2 Canal Water Level Instrumentation (NUREG-0822, Section 4.1(3))
In IPSAR Section 4.1(3), the staff concluded that water level instrumentation in the intake canal was inadequate and there was no water level measurement in the discharge canal.
Accordingly, the staff recommended that automatic water level instrumentation be provided in both canals, with measurement indication in the control room, so that the operator would be able to implement emergency shutdown procedures when the specified flooding levels occurred.
Because these i
instruments are not intended for postaccident monitoring, they need not neces-sarily be safety grade.
The staff also stated that the licensee had committed to install an automatic water level gage, with a remote readout, in the intake canal.
Another water level gage in the discharge canal was not necessary because flooding conditions could be identified from the intake canal measurement.
This modification would be coordinated with other modifications being considered by the licensee l
for canal monitoring, including upgrading the existing gages, and the installa-l tion would be completed by the end of the Cycle 11 refueling outage.
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In its letter dated April 21, 1986, the licensee asked to cancel-its commitment to install an automatic water level gage in the intake canal with a remote read-out in the control room.
The licensee proposed. revising Station Procedure 2000-ABN-3200.31, "High Winds," to require a plant shutdown when the water level at the intake structure cannot be verified to be less than elevation 4.5 feet MSL.
This was acceptable to the staff and was documented in the staff's safety
'l evaluation da',ed November 28, 1986.
The former NRC project manager also veri-j fied that this shutdown requirement had been added to Procedure 2000-ABN-3200.31.
This closes out this SEP issue.
Low water level in the intake structure is discussed in Section 3.1.2 of this supplement.
i 2.1.3 Protection During Internal Flooding (NUREG-0822, Section 4.1(7))
In Section 4.1(7) of the IPSAR,.the staff stated that protection against internal flooding of structures caused by local probable maximum precipitation should be provided to a flood level of 23.5 feet MSL.
The licensee should verify that all entrance levels were above this level.
The southwest door of the offgas building may flood even though the sill is at 23.5 feet MSL because of the con-figuration of contours near the door.
The staff stated further that the licensee had proposed to evaluate the conse-quences of flooding in the offgas building and would confirm that no other en-trance level was below 23.5 feet MSL.
By letter dated June 6,1983, the licensee stated that all sill and entry flood elevations are at or above 23 feet, 6 inches MSL for the reactor building, the turbine building, and the new and old radwaste buildings and, thus, modifications were not required.
However, the licensee's review indicated that two entrances in the diesel generator building are at elevation 23 feet MSL, which could expose the enclosed switchgear cabinets to flooding.
The licensee proposed to construct a 6-inch-high asphalt dike at the above two entrances to prevent surface water from entering during the next operating cycle.
Further, the licensee stated that a review of contour maps of the rite had shown no indication of contours that might impound water at the southwest door of the offgas building end, therefore, modifications were not required.
In a letter dated June 23, 1983, the staff found the proposed corrective actions acceptable and sufficient to resolve this SEP issue.
Region I staff will verify that the modifications discussed above have been completed.
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- 2. 2 Topic III-1, Classification of Structures, Components, and Systems l
(Seismic and Quality) (NUREG-0822, Section 4.2) 10 CFR 50 (GDC 1), as implemented by Regulatory Guide 1.26, requires that struc-tures, systems, and components important to safety be designed, fabricated, erected, and tested to quality standards commensurate with the importance of safety functions to be performed.
The codes used for the design, fabrication, erection, and testing of the Oyster Creek plant were compared with current codes.
In Section 4.2 of the IPSAR, the staff stated that it had identified several i
systems and components for which the licensee was unable to provide imformation 2-3 l
to justify a conclusion that the quality standards imposed during plant con-struction met quality standards required for new facilities.
The staff did not identify any inadequate components.
However, because of the limited informa-tion on the components involved, the staff was unable to conclude that for code and standard changes deemed important to safety, the Oyster Creek plant met current requirements.
The staff further stated that the licensee had agreed to complete the evaluations described in IPSAR Section 4.2 and to incorporate the results in the Final Safety Analysis Report update, which must be submitteo within 2 years after completion of the SEP review (10 CFR 50.71 (e)(3)(ii)).
If the results of the licensee's evaluations indicated that facility modifications were required, they would be reported in a licensee event report.
The licensee indicated that it would provide this information.
When the infor-mation is received, the staff will evaluate it and present its finding in a sup-plement to the IPSAR.
2.3 Topic III-2, Wind and Tornado Loadings (NUREG-0822, Section 4.3) 10 CFR 50 (GDC 2), as implemented by SRP Sections 3.31 and 3.32 and Regulatory Guides 1.76 and 1.117, requires that the plant be designed to withstand the effects of natural phenomena such as wind and tornadoes.
In Section 4.3 of the IPSAR, the staff identified several structures important to safety that had not met current licensing criteria regarding their ability to resist tornadoes.
2.3.1 Reactor Building Steel Structure Above the Operating Floor (NUREG-0822, Section 4.3.1)
In Section 4.3.1 of the IPSAR, the staff concluded that the capacities it had calculated were lower (differential pressure induced by a windspeed of 61 miles per hour (mph)) than those required by the site-specific tornado-imposed loads.
The staff also indicated that the licensee was analyzing these structures to determine capacities and would provide the results and identify proposed correc-tive actions to the NRC staff.
l In a letter dated February 2, 1983, the licensee prnvided supporting calculations to justify its conclusions in its letter of May 7, 1981.
The staff, with assistance from the Franklin Research Center staff, reviewed the supporting calculations and did not agree with the limiting windspeed provided by the licensee.
In a safety evaluation report (SER) dated March 8, 1986, the staff concluded that the licensee should (1) determine the capability of the structure with appropriate considerations as presented by the staf f in Sec-tion IIIA of the SER and (2) evaluate potential modifications that would increase the plant's capability to withstand severe wind and tornado loads.
During meetings at the Oyster Creek station on February 2 to February 6, 1987, the licensee stated that it would provide the information concerning this matter.
When this information is received, the staff will evaluate it and report its finding in a supplement to the IPSAR.
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2.3.2 Ventilation Stack (NUREG-0822, Section 4.3.2)
In Section 4.3.2 of the IPSAR, the staff concluded that the stack capacities it had calculated were lower (164-mph windspeed) than those required by the site-specific tornado-imposed loads.
Failure of the stack could affect the integrity of seismic Category I structures because the stack is in close proximity to these structures.
The staff also stated that the licensee was performing an analysis of the stack and a probabilistic evaluation of tornado (or high-wind)-induced stack failure and its consequences and that it would identify any necessary cor-rective actions and submit the results of the analysis to the staff.
By letter dated February 2,1983, the licensee submitted the results of its analysis and showed that the stack was capable of withstanding a ISO-mph wind load.
The licensca concluded that the 180-mph wind load corresponds to an exceed-ance probability of 1x10 6/ year, which is sufficiently low to make the installa-tion of modifications unwarranted.
In an SER dated March 8, 1986, the staff noted that 180 mph corresponds to a probability of exceedance of approximately 5x10-6/ year using the NRC estimate of tornado hazard at Oyster Creek.
The staff also concluded that, considering the various means of plant shutdown avail-able, the conditional probability of core damage given stack failure was accept-ably small.
The stack is capable of withstanding 180 mph (5x10 6/ year) if reso-nance does not occur.
Therefore, the staff concluded that no further evaluation of the stack was warranted.
The staff also concluded that the issue of tornado loads in conjunction with wind loads for the stack was resolved.
This closes out this SEP issue.
2.3.3 Effects of Failure of Nonseismic Category I Structures (NUREG-0822, Section 4.3.3)
In Section 4.3.3 of the IPSAR, the staff stated that the licensee would evaluate the turbine building capacity and the effect of its failure on other structures (e.g., the control roon) and that it would identify any necessary corrective actions and submit the results of the evaluation to the NRC staff.
The licensee provided the results of its evaluation in a submittal dated March 13, 1984.
The licensee's analysis modeled the turbine building as two-and three-dimensional frames and analyzed them using a computer code.
As a result of the analysis, the licensee concluded that for load combinations involving loads such as dead load and snow in combination with wind, the turbine building would remain stable for load conditions involving an 80-mph wind loading.
The licensee also concluded that failure of the roof purlins by the roof deck /
purlin connections would not cause turbine building failure.
In its safety evaluation dated March 8, 1986, the staff concluded that, overall, the structural system of the turbine building is capable of resisting reasonably high levels of loading.
It also concluaed that no further evaluation of the turbine building was warranted.
This closes out this SEP issue.
The issue of wind-load combinations is addressed in Section 3.12 of this supplement.
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2.3.4 Roof Decks (NUREG-0822, Section 4.3.6)
In Section 4.3.6 of the IPSAR, the staff stated that the licensee had indicated that the roof deck of the reactor building could withstand a 280-mph wind and a 0.68 psi differential pressure and that the roof of the diesel generator build-ing coulo withstand a 300-mph wind and a 2 psi differential pressure.
The staff also stated that the licensee would evaluate the capacity of the roof deck of the turbine building.
In a letter dated March 13, 1984, the licensee provided an evaluation of the roof decks of these structures in its analysis of the capacities of the reactor build-ing above the operating floor and the turbine building.
In its safety evaluation dated March 8,1986, the staff addressed the roof decks as a part of its evaluation of the reactor building above the operating floor and the turbine building.
As discussed in Section 2.3.1 of this supplement, the evaluation of the reactor building above the operating floor is not complete and therefore this issue is considered open.
With regard to the roof deck of the turbine building, as discussed in Section 2.3.3 of this supplement, the staff concludes that the overall structural system of the turbine building is capable of resisting high levels of loading and considers the turbine building issue resolved.
2.3.5 Intake Structure, Oil Tanks, and Diesel Generator Building (NUREG-0822, Section 4.3.7)
In Section 4.3.7 of the IPSAR, the staff stated that it did not have sufficient information to be able to conclude that these structures had enough capacity for the pcstulated wind and tornado loadings and that the licensee would submit an analysis of these structures.
In its safety analysis report dated May 7, 1981, 'he licensee concluded that the intake structure and the diesel generator ano oil tank vaults were capable of withstanding a 300-mph wind and a 2.0 psi depressurization load.
By letters dated February 2 and October 25, 1983, the licensee provided supporting calcu-lations for the values given in the report.
In its safety evaluation of March 8, 1986, the staff concluded that the windspeed ratings of these structures as determined by the licensee were valid.
- However, the licensee had not provided evaluations so that the staff could evaluate the effects of tornado-missile loads in load combinations involving tornado wind loads for the diesel generator and oil tanks.
The staff further stated that the major portions of the diesel generator and oil tank vaults are substantial reinforced concrete structures with a roof thickness t
of 1 foot 0 inch and wall thicknesses of 1 foot 6 inches.
Although the licensee has not evaluated missile loads in combination with wind loads, the thicknesses of the structure's roof and walls are such that it is expected that they will afford a substantial amount of protection.
The c.apacity to resist missile and wind loads simultaneously would be less than 300 mph as reported for wind acting alone; however, even if the resistance reduces to a windspeed such as 120 mph, the probability of exceedance is approximately 5x10 */ year, which is low.
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In its safety evaluation of March 8, 1986, the staff also stated that in response to the tornado-missile issue, the licensee had committed to provide a portable pump in a protected area and hose connections to a protected water supply to be used in conjunction with the isolation condenser to achieve hot shutdown.
- Thus, even if the oil tank and diesel generators should be damaged, safe shutdown still I
could be achieved.
The staff concluded from this that analysis to determine if the oil tank and diesel generator vaults can withstand tornado-missile loads in combinations involving tornado wind loads is not warranted.
As discussed in Section 2.1.1, the staff has not completed its review of the licensee's proposed method of supplying makeup water to the isolation condenser.
When it completes its review, the staff will document the results in a supple-ment to the IPSAR.
2.3.6 Load Combinations (NUREG-0822, Section 4.3.8)
In Section 4.3.8 of the IPSAR, the staff stated that as a result of the topic review, it was unable to determine if straight wind loads (not tornado loads) were combined with other loads (i.e., snow loads, operating pipe reaction loads, and thermal loads).
The staff also stated that the licensee had stated that recent analyses had included these loads.
These analyses were to be submitted to the staff in conjunction with Topic III-7.B (see IPSAR Section 4.12).
The staff's evaluation of this matter is provided in Section 2.10 of this supplement.
2.3.7 Control Room (NUREG-0822, Section 4.3)
In its SER dated September 1, 1982, the staff stated that the licensee should provide a description of the methods and sample calculations used to qualify the control building.
By letter dated February 2, 1983, the licensee provided supporting calculations for control room capacities.
The licensee concluded that the control room north wall was capable of resisting a 160-mph tornado wind and a 0.53 psi depressurization load.
The balance of the control room was capable of resisting a 300-mph tornado wind and a 2.0 psi pressure differential.
The 160-mph wind and a 0.53 psi pressure drop correspond to a probability of exceedance of approximately 1x10 h/ year.
In its safety evaluation dated March 8, 1986, the staff concluded that the wind-speed ratings developed by the licensee were valid.
The licensee's February 1, 1983, submittal also noted that the control room could not withstand the tornado-missile load in conjunction with the tornado wind load.
No assessment of the effects of failure of the wall have been provided.
The staff concluded that the licensee should demonstrate that failure of the wall would not prevent safe plant shutdown or should propose corrective actions.
The licensee is evaluating this matter and will submit the results to the staff.
The staff will review the licensee's evaluation when it is submitted and document the results in a supplement to the IPSAR.
2.3.8 Architectural Components (NUREG-0822, Section 4.3)
In its safety evaluation dated March 8, 1986, the staff stated that the licensee should verify that architectural components, such as rollup doors, were not lo-cated so that damage to required equipment could occur.
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The licensee is evaluating this matter and will submit the results to the staff.
The staff will review the licensee's evaluation when it is submitted and document the results in a supplement to the IPSAR.
2.4 Topic III-3.A, Effects of High Water Level on Structures (NUREG-0822, Section 4.4) 10 CFR 50 (GDC 2), as implemented by SRP Section 3.4 and Regulatory Guide 1.59, requires that plant structures be designed to withstand the effects of flooding.
' 4.1 Hydrostatic Loads (Short-Duration) (NUREG-0822, Section 4.4(2))
In Section 4.4(2) of the IPSAR, the staff concluded that the licensee should demonstrate that safety-related structures would remain functional for a short-term hydrostatic loading and could resist flotation for water levels up to 22 feet.
By letter dated July 1, 1983, the licensee provided results of its analyses for the reactor building, the turbine building, the diesel generator building, and the new radwaste building.
On the basis of the results, the licensee con-cluded that the structures were adequate to resist the loadings.
In its evaluation issued on February 23, 1984, the staff concluded that on the basis of the factors of safety obtained against flotation, the adequacy of the subgrade walls, and the adequacy of bearing capacity, the Oyster Creek facility can adequately withstand a groundwater level of elevation 23 feet MSL.
This closes out this SEP issue.
2.5 Topic III-4.A, Tornado Missiles (NUREG-0822, Section 4.6) 10 CFR (GDC 2), as implemented by Regulatory Guide 1.117 prescribes structures, systems, and components that should be designed to withstand the effects of a tornado, including tornado missiles, without loss of capability to perform their safety functions.
In Section 4.6 of the IPSAR, the staff identified several structures and compo-nents that were vulnerable to tornado missiles.
j 2.5.1 Emergency Diesel Generators and Fuel Oil Day Tank (NUREG-0822, Section 4.6.1)
In Section 4.6.1 of the IPSAR, the staff stated that the licensee had concluded that the diesel generators were not necessary for safe shutdown because makeup water could be provided to the isolation condenser by diesel-driven fire water pumps and by de power to the main steam relief valves.
The staff also indicated that the licensee had agreed to evaluate the potential for and consequences of i
tornado-missile damage to the diesel generator building.
The status of this issue is discussed in Section 2.3.5 of this supplement.
2.5.2 Mechanical Equipment Access Area (HUREG-0822, Section 4.6.2)
In Section 4.6.2 of the IPSAR, the staff identified several components (e.g.,
motor control centers (MCC-DC-1 and MCC-1AB 21B), control rod drive hydraulic filter, isolation fill piping, and containment spray valve) in the vicinity of 2-8
the mechanical equipment access opening of the reactor building that were poten-tial targets for missiles penetrating the access doors.
These components had not been considered in the staff's original evaluation.
The staff also stated that the licensee had agreed to evaluate the potential for and consequences of tornado-missile impact on components in this area and provide protection, if necessary.
By letter dateo September 16, 1983, the licensee provided an analysis of tornado-missile risk for Oyster Creek.
This analysis included the development of an annual tornado windspeed exceedance curve.
In a letter dated December 27, 1983, the staff stated that it had independently calculated a probability distribution for high winds and tornadoes for the Oyster l
l Creek site and found that non-tornado-wind frequency was higher than the licen-see's values.
Therefore, the staff requested that the licensee evaluate the con-sequences of wind generated missiles (from windspeeds less than 125 mph) to de-termine whether such missiles contribute significantly to the damage of targets.
In letters dated October 15, 1984, and June 7, 1985, the licensee provided the requested information.
The staff has not completed its review of this infor-mation. When the review is completed, the staff will document the results in a supplement to the IPSAR.
2.5.3 Condensate Storage Tank, Torus Water Storage Tank, and Service Water and Emergency Service Water Pumps (NUREG-0822, Section 4.6.4)
In Section 4.6.4 of the IPSAR, the staff stated that the licensee's position was that the condensate storage tank and torus water storage tank were not required to accomplish safe shutdown because the plant could be safely shut down using one of the two service water pumps or any of the four emergency service water pumps and that backfitting was not required because the pumps were redundant.
The staff also indicated that redundancy was not acceptable protection from tor-nado missiles.
Therefore, it was the staff's position that the licensee provide protection for sufficient systems and components to ensure a safe shutdown in the event of damage from tornado missiles.
The staff also stated that the licensee had agreed to provide a portable pump in a protected area and hose connections to a protected water supply and to provide procedures that specified the conditions for and use of this equipment.
The staff found this action acceptable.
However, as discussed in Section 2.1.1 of this report, the licensee now proposes to use an existing system inter-connection between the core spray and condensate and denineralized water transfer systems to achieve safe shutdown of the plant.
The staff has not completed its review of this matter.
When the review is completed, the staff will document the results in a supplement to the IPSAR.
2.6 Topic III-4.B. Turbine Missiles (NUREG-0822, Section 4.7)_
10 CFR 50 (GDC 4), as implemented by Regulatory Guide 1.115 and SRP Section 3.5.1.3, requires that structures, systems, and components important to safety be appropriately protected against dynamic effects, which include missiles.
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One means of providing adequate protection is assurance of a low probability of failure of the turbine at design or destructive overspeed.
This assurance arises in part from inspection of the turbine discs and testing and inspection of stop and control valves at regular intervals.
In Section 4.7 of the IPSAR, the staff concluded that the licensee should (1) perform a volumetric inspection of the turbine during the Cycle 10 outage and, on the basis of the results of that inspection, propose an inspection frequency (2) describe the monitoring program for main steam control valves and reheat control valves and justify why these valves should not be cycled individ-ually to a fully closed position on a weekly basis In a letter dated May 17, 1984, the licensee described inspections performed in April and June 1983 by General Electric (GE) of the shrunk-on wheels from low-pressure rotors LPA, LPB, and LPC.
Visual and ultrasonic e~aminations were performed.
Indications on the wheel bores and keys were found by ultrasonic examination. GE and the licensee concluded that the indications did not affect the structural integrity of the wheels and keyways and, as a result, GE recommende( that another ultrasonic inspection be performed after approxi-mately 6 years of additional operation.
The licensee committed to conduct the inspection within 6 years of operation, which is the schedule typically recom-mended by the vendor for its turbines.
In a letter dated December 8, 1983, the licensee described the valve monitoring program.
The four individual turbine stop valves are closed fully on a daily basis.
The six reheat stop valves and six intercept valves are individually brought to full closure once a week.
In its SER dated August 21, 1986, the staff concluded that the licensee had proposed a turbine inspection schedule based on a previous inspection and on vendor recommendations.
The testing meets the intent of staff criteria, that is, to verify the ability of the stop and control valves to close to prevent turbine overspeed, even though full-closure testing of the control valves is not practical.
Therefore, the staff concludes that the licensee's response to IPSAR Section 4.7 is acceptable.
This closes out this SEP issue.
2.7 Topic III-4.0, Site-Proximity Missiles (Including Aircraft) (NUREG-0822, Section 4.8) 10 CFR 50 (GDC 4), as implemented by SRP Sections 3.5.1.5, 3.5.1.6, and 2.2.3, requires that structures, systems, and components important to safety be appro-priately protected against the effects of missiles that may result from events and conditions outside tha nuclear power unit.
2.7.1 Aircraft Hazards (NURt M BE, Section 4.8.2)
In Section 4.8.2 of the IPSAR, the staff concluded that because there are sev-eral airports near the site, the licensee should address the potential for or consequences of aircraft impact.
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The licensee submitted its analysis of the probability of an aircraft strike on the plant in a letter dated March 4,1983.
The staff reviewed the licensee's submittal and issued its evaluation dated May 3, 1983.
In that evaluation, the staff concluded that the licensee's analysis was performed in accordance with current criteria and that because the aircraft strike probabilities are extremely low, aircraft traffic does not pose a significant threat to the Oyster Creek plant.
Therefore, this issue is considered resolved.
2.8 Topic III-5 B, Pipe Break Outside Containment (NUREG-0822, Section 4.10) 10 CFR 50 (GDC 4), as implemented by SRP Sections 3.6.1 and 3.6.2 and Branch l
Technical Positions (BTPs) MEB 3-1 and ASB 3-1 (NUREG-0800), requires, in part, that structures, systems, and components important to safety be designed to accommodate the dynamic effects of postulated pipe ruptures.
The safety objective for this topic review is to ensure that if a pipe should break out-side the containment, the plant can be safely shut down without a loss of containment integrity.
2.8.1 Emergency Condenser Isolation (NUREG-0822, Section 4.10(2))
In Section 4.10(2) of the IPSAR, the staff stated that the emergency condenser steam lines have two automatic isolation valves outside and adjacent to the drywell.
A break between these valves with a failure of the first isolation valve or a pipe break between the second valve and the condenser resulting in pipe whip so that the isolation valves would not close would both result in a loss-of-coolant accident outside the containment.
The physical arrangement and space availability preclude installation of restraints.
In addition, it is not practical to install an isolation valve inside the drywell.
The licensee's inservice inspection of these lines was in accordance with Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code).
The licensee committed to submit a reanalysis of the emergency condenser piping along with an evaluation of leakage detection and a schedule for any necessary corrective actions in February 1983.
In a letter dated October 16, 1984, the licensee provided the report entitled "Crack Growth and Leak Rate Assessment of the Oyster Creek Emergency Condenser System Piping Outside Containment Below the 95 Foot Elevation." The analysis I
indicated that the leak rates from postulated cracks are sufficiently high so that visual monitoring is an acceptable method of leak detection.
The licensee further stated that sufficient time exists to take appropriate actions (i.e.,
i shut down or isolate the affected condenser) between the time of leak detection and the time that a crack would grow to an unstable length.
In January 1986, Niagara Mohawk Power Corporation (NMPC) notified the NRC of a j
failure mode involving the drywell penetrations at Nine Mile Point.
Loads cal-culated for the penetrations resulting from a postulated high energy line break (HELB) in the process piping within the penetrations were determined to exceed those for which the penetrations were designed.
These higher loads resulted from the use of a more accurate analysis model that included the effects of both l
pressure and momentum.
The licensee became aware of the NMPC analysis and in February 1986 voluntarily initiated an investigation of the Oyster Creek dry-well penetrations.
The results of this investigation were discussed with the 2-11
._.,_,_.,.,__-..-_.y y
staff in a meeting held on August 22, 1986 (the NRC minutes of the meeting are dated October 1, 1986).
During that meeting and in Licensee Event Report No.86-024, dated October 17, 1986, the licensee stated that four penetrations for the isolation condenser (IC) piping were below the updated Final Safety Analy-sis Report design criteria and would fail if a guillotine rupture of the pipe occurred within the penetration.
As a result of several teiephone discussions with the staff, the licensee provided additional information in letters dated September 17 and November 25, 1986.
In the latter letter, the licensee committed (1) to resolve the problem associated with the four piping penetrations in accordance with the NUREG-0313, Revision 2 ("Technical Report on Material Isolation and Processing Guidelines for BWR Coolant Pressure Boundary Piping") requirements regarding welds inside these penetrations and SEP Topic III.5.B regarding the two containment isola-ion valves outside the containment on the IC steam lines and (2) to operate Oyster Creek with additional limiting conditions for operation in regard to the reactor coolant leakage within the drywell.
In a letter dated December 24, 1986, the staff stated that it had reviewed the licensee's letter and data on the HELB within the IC penetrations through the drywell and, as discussed in the safety evaluation dated December 24, 1966, the staff concluded that operation of Oyster Creek during operating Cycle 11 was acceptable.
The modifications to the four piping penetrations and the IC piping at the 75-foot elevation outside the containment will be completed dur-ing the Cycle 13R outage, which will start in 1991.
In its letter of December 24, 1986, the staff also stated that, as the licensee had explained in its letter dated November 25, 1986, completing the modification during the Cycle 13R outage was contingent on finalizing the design of the penetrations and completing the engineering for modifications
]
in time for the outage.
The staff will be involved in this activity with the licensee because this design will involve NUREG-0313, Revision 2, and SEP j
Topic III-5-B.
When it receives the licensee's proposed resolution of this matter, the staff l
will review the information and report the results in a supplement to the IPSAR.
- 2. 9 Topic III-6, Seismic Desian Considerations (NUREG-0822, Section 4.11) 10 CFR 50 (GDC 2), as implemented by SRP Sections 2.5, 3.7, 3.8, 3.9, and 3.10 and SEP review criteria (NUREG/CR-0098, "Development of Criteria for Seismic Review of Selected Nuclear Power Plants"), requires that structures, systems, and components important to safety be designed to withstand the effects of natural phenomena such as earthquakes.
In Section 4.11 of the IPSAR, the staff identified the areas discussed in the following sections as needing further evaluation.
2.9.1 Piping Systems (NUREG-0822, Section 4.11(1))
In Section 4.11(1) of the IPSAR, the staff stated that the licensee should per-form analyses on a sampling basis (e.g., analyses of two randomly selected pip-ing systems) of piping systems 2 1/2 to 10 inches in diameter as well as submit j
i 2-12
i.he reanalysis of the control rod drive system to the site-specific spectra, including information on the building model and floor response spectra.
In addition, the licensee should verify the design adequacy of piping supports for the main steam and feedwater lines.
The licensee provided several submittals responding to the staff's require-ments as specified in Section 4.11(1) of the IPSAR.
The staff reviewed this information and by letter dated January 9, 1986, provided a draft technical evaluation report that identified the areas where additional information was i
needed.
This matter was discussed at a meeting on April 24, 1986 (meeting sum-mary dated May 19, 1986), and, in a letter dated June 24, 1986, the licensee provided the additional information.
The staff is reviewing this information.
When its review is completed, the I
staff will document the results in a supplement to the IPSAR.
2.9.2 Mechanical Equipment (NUREG-0822, Section 4.11(2))
In Section 4.11(2) of the IPSAR, the staff required the licensee to demonstrate that the control rod drive hydraulic units and associated tubing supports as well as the reactor vessel internals had sufficient capacity to maintain integ-rity followinq the safe shutdown earthquake (SSE).
In a letter date January 20, 1983, the licensee submitted a report entitled "Reanalysis of the Control Rod Drive Return System Piping Considering Axial U-Bolt Restraint and Site Specific Spectra." The licensee stated that the reanalysis demonstrated that the control rod drive return system piping stresses were within code-allowable values for the Oyster Creek site-specific seismic spectra.
In a letter dated January 24, 1983, the licensee submitted a report entitled "0yster Creek Seismic Analysis of Reactor Vessel Internals." The licensee stated that the analysis was performed to address questions raised by the NRC staff during the review of SEP seismic considerations.
The staff is reviewing this information.
When its review is completed, the staff will document the results in a supplement to the IPSAR.
2.9.3 Electrical Equipment (NUREG-0822, Section 4.11(3))
In Section 4.11(3) of the 1PSAR, the staff stated that it was concerned that the licensee had not demonstrated the structural integrity of the panels (load path from an internally mounted element to anchorage support systems).
The licensee proposed to perform an evaluation of the load path for at least two typical cabinets.
In a letter dated March 13, 1984, the licensee submitted its evaluation and results of a seismic analysis of two types of safety-related equipment at Oyster Creek:
4160-volt switchgear and 460-volt unit substation cabinets.
Anchor adequacy and internal load path evaluations for both types of the cabinets were conducted.
The staff reviewed this information and, in a letter 4
dated January 8, 1986, provided a draft technical evaluation report that i
i identified the areas where additional information was required.
This matter was discussed at a meeting on April 1986 (meeting summary dated May 19, 1986),
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_l
and the licensee provided additional information in a letter dated June 24, 1986.
The staff is reviewing this information.
When its review is completed, the staff will document the results in a supplement to the IPSAR.
2.9.4 Qualification of Cable Trays (NUREG-0822, Section d.11(5))
The staff is concerned that safety-related cable trays may not be able to withstand the postulated seismic loads.
The SEP Owners Group has conducted tests on typical cable tray configurations found in nuclear power plants.
One report summarizing the test results was submitted to the staff in April 1983; l
a second report containing cable tray evaluation criteria and guidelines developed from the tests was submitted in August 1983.
On February 19, 1987, the NRC issued Generic Letter 87-02, "Verification of Seismic Adequacy of Mechanical and Electrical Equipment in Operating Reactors, Unresolved Safety Issue (USI) A-46."
The generic letter provided guidance for the resolution of USI A-46 and requested that licensees submit a schedule for final resolution of the issue using that guidance.
In a letter dated October 9,1987, the Seismic Qualification Utility Group (SQUG), of which GPU is a member, stated that it was developing the "Generic Implementation Procedure for Verification of Seismic Adequacy-of Nuclear Plant Equipment" for use by its members.
Part 1 of the procedure will provide SQUG's positions regarding Generic Letter 87-02.
Part 2 will be a detailed technical document containing criteria and associated guidance for the resolu-tion of USI A-46.
Part 3 will consist of a series of training seminars to be sponsored by SQUG.
In a letter dated November 19, 1987, the staff advised SQUG that its letter of October 9, 1987, was acceptable to meet the December 1, 1987, reporting deadline as set forth in the April 28, 1987, letter from the NRC staff to I
SQUG.
Therefore, licensees participating in the SQUG program do not need to respond and provide separate responses to Generic Letter 87-02 until u e staff issues its safety evaluation report (SER).
All licensees participating in the SQUG implementation program should provide their schedules for plant-specific implementation no later than 60 days after receiving the generic
- SER, t
2.10 Topic III-7.B. Design Codes, Design criteria, Load Combinations, and Reactor Cavity Design Criteria (NUREG-0822, Section 4.12) l 10 CFR 50 (GDC 1, 2, and 4), as implemented by SRP Section 3.8, requires that Structures, systems, and components be designed for the loading that l
will bc imposed on them and that they conform to applicable codes and standards.
In Section 4.12 of the IPSAR, the staff concluded that areas of design code i
changes potentially applicable to the Oyster Creek plant for which the code in effect at that time required substantially greater safety margins than the earlier version of the code or for which no original code provision existed should be evaluated to ensure adequate margins of safety.
The licensee com-mitted to (1) review the NRC evaluation to determine applicability of the 2-14 i
i
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structural elements identified and (2) perform, on a sampling basis, an eval-uation of the code, load, and load combination changes in regard to existing as-built structures to assess the adequacy of the design.
-By letter dated June 4, 1984, the licensee submitted an evaluation of design codes, design criteria, and load combination changes for Oyster Creek as requested in Section 4.12 of the IPSAR.
In its safety evaluation dated October 29, 1986, the staff concluded that, on the basis of its review and that of its consultant, Franklin Research Center, the load and load combination issues were satisfactorily resolved.
With respect to the design code and criteria changes, 20 of the 23 issues were fully resolved.
For two of the design code changes (related to the reinforcement of openings), further information was requested.
For the remaining issue - concrete subject to high temperatures and thermal tran-sients - the licensee stated that further investigation of drywell thermal conditions was necessary.
The staff will review the information when it is provided by the licensee and report the results in a supplement to the IPSAR.
2.11 Topic III-10. A, Thermal-Overload Protection for Motors of Motor-0perated Valves (NUREG-0822, Section 4.14) 10 CFR 50.55a(h), as implemented by Institute of Electrical and Electronics Engineers Std. 279-1971 and 10 CFR 50 (GDC 13, 21, 22, 23 and 29), requires that protective actions be reliable and precise and that they satisfy the single-failure criterion using quality components.
2.11.1 Thermal-0verload Bypass (NUREG-0822, Section 4.14(1))
In Section 4.14(1) of the IPSAR, the staff concluded that the licensee had not demonstrated the adequacy of the setpoints for unbypassed thermal overloads on J
some safety-related valves.
The licensee agreed to evaluate the setpoints and propose any necessary corrective actions.
The licensee provided the methodology for establishing setpoints in a letter dated July 30, 1983.
In its SER dated August 20, 1984, the staff concluded that the licensee had developed a coherent methodology for establishing thermal-overload trip setpoints with all uncertainties resolved in favor of completing the safety-related-valve action.
The staff further concluded that the program, methods, and schedule proposed in the licensee's letter of July 30, 1984, pro-vide an acceptable resolution of the issue discussed in IPSAR Section 4.14(1).
This closes out this SEP issue.
2.12 Topic V-5, Reactor Coolant Pressure Boundary (RCPB) Leakage Detection (NUREG-0822, Section 4.16) 10 CFR 50 (GDC 30), as implemented by Regulatory Guide 1.45 and SRP Section 5.2.5, prescribes the types and sensitivity of systems and their seismic, indication, and testability criteria necessary to detect leakage of primary reactor coolant to the containment or to other interconnected systems.
Regulatory Guide 1.45 recommends that at least three separate leak detection systems be installed in 2-15 d
a nuclear power plant to detect unidentified leakage from the RCPB to the primary containment of 1 gpm within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
Leakage from identified sources must be isolated so that flow of this leakage may be monitored separately from uniden-tified leakage.
The detection systems should be capable of performing their functions after certain seismic events and of being checked in the control room.
Of the three separate detection methods recommended, two of the methods should be (1) sump level and flow monitoring and (2) airborne particulate radioactivity monitoring.
The third method may be either monitoring the condensate flow rate from air coolers or monitoring airborne gaseous radioactivity.
Other detection methods - such as monitoring humidity, temperature, or pressure - should be con-sidered to be indirect indications of leakage to the containment.
In addition, provisions should be made to monitor sy:,tems that interface with the RCPB for signs of intersystem leakage through methods such as monitoring radioactivity and water levels or flow.
2.12.1 Leakage Detection Systems (NUREG-0822, Section 4.16.1)
In Section 4.16.1 of the IPSAR, the staff stated that Oyster Creek had only one of the detection systems (sump level monitoring) recommended in Regulatory Guide 1.45.
The staff further stated that the plant had an airborne partic-ulate and gaseous radiation monitoring system (APGRMS) installed in the drywell.
This latter system is also recommended in the regulatory guide; however, the system had never been placed in operation at Oyster Creek because of problems.
The APGRMS would be w ed to detect RCPB leakage indirectly by measuring the radioactivity in the drywell atmosphere that had resulted from the reactor coolant water leakage into the drywell.
The licensee committed to (1) identify the system modifications necessary to make the airborne particulate and gaseous radioactivity monitors operational, (2) evaluate the reliability and sensitivity of the existing leakage detection systems, and (3) propose a schedule for any necessary system modifications or procedural changes.
In its letter dated July 29, 1985, requesting deferment of the installation and testing of the APGRMS to the Cycle 12R outage, the licensee stated that its evaluation of the APGRMS had revealed numerous problem areas requiring extensive redesign, modification, or replacement of the system.
The licensee was assessing various alternatives in order to arrive at a working system and stated that, considering the extent of the remaining design work and projected delivery times for equipment, it anticipated that the APGRMS would be installed and tested during the Cycle 12R outage.
For the above reasons, the licensee requested deferment of the installation and testing of an operating APGRMS to detect RCPB leakage to the Cycle 12R outage.
The staff granted this deferment in its letter dated October 6, 1986.
In its letter dated July 8, 1986, the licensee described the adequacy of its sump monitoring system to detect RCPB leakage.
The licensee concluded that this system's sensitivity was sufficient to allow safe shutdown before a crack would grow to an unstable length.
Limiting conditions for operation and surveil-lance requirements on this system were incorporated in the Technical Specifica-tions (TS) in Amendment 97 to the license dated January 6,1986.
Therefore, 2-16 4
-y v
the licensee has evaluated the reliability and sensitivity of the existing sump detection system as requested in IPSAR Section 4.16.1.
The licensee also stated in its letter dated July 8, 1986, that a new APGRMS would have to be designed, installed, and tested for Oyster Creek.
It con-cluded, however, that the APGRMS would be of little use in quantifying leakage rates to meet TS leakaca limits.
The APGRMS would measure the leakage indi-rectly through released radioactivity and could only be used as a trending indication of the leakage that must be confirmed and quantified by other means.
Therefore, the licensee concluded that the APGRMS is of limited value and other data such as drywell pressure, humidity, and temperature are available that can provide the information needed concerning RCP8 leakage.
The licensee has identified the system modifications needed to make the APGRMS operational and has committed to install the system during the Cycle 12 outage.
This completes the information requested from the licensee in IPSAR Section 4.16.1.
The licensee's request in its letter dated July 8, 1986, to cancel this commitment has been reviewed by the staff.
In its letter of March 12, 1987, the staff concluded that the licensee had not provided suffi-cient justification to cancel its commitment to install the APGRMS.
The licensee had not provided in detail the lack of sensitivity of the APGRMS, the cost of making the APGRMS operational, and the sensitivity of other data such as drywell pressure, humidity, and temperature to provide information concern-ing RCPB leakage.
Regulatory Guide 1.45 recommends that at least three sepa-rate detection methods should be used including sump detection and an APGRMS.
Therefore, the staff concludes that the APGRMS should be installed during the Cycle 12R outage.
Region I staff will verify that the APGRMS is installed dur-ing the Cycle 12R outage.
2.13 Topic V-11.A, Requirements for Isolation of High-and low-Pressure Systems (NUREG-0822, Section 4.19) 10 CFR 50.55a, as implemented by SPR Section 7.6 and Branch Technical Position ICSB 3 (NUREG-0800), requires that interlock systems important to safety be adequately designed to ensure their availability in the event of an accident.
i This includes those systems with direct interface with the reactor coolant system that have design pressure ratings lower than the reactor coolant system design pressure.
In Section 4.19 of the IPSAR, the staff concluded that the pressure interlocks on the reactor water cleanup system did not satisfy licensing criteria because they were not independent.
By letter dated August 4,1983, the licensee submitted further information on i
the design features of the reactor water cleanup system that would prevent low-pressure piping from being exposed to high pressure reactor coolant.
In an evaluation issued by letter dated September 20, 1983, the staff concluded that i
the interlock logic, the diversity of signals, and the relief valves provided reasonable assurance that the piping with design pressure ratings lower than the reactor coolant system design pressure will not be overpressurized.
There-fore this issue is considered resolved.
I 2-17 4
2.14 Topic VI-4, Containment Isolation System (NUREG-0822, Section 4.22) 2.14.1 Remote Manual Valves (NUREG-0822, Section 4.22.2) 10 CFR 50 (GOC 54 through 57) as implemented by SRP Section 6.2.4 and Regulatory Guides l.11 and 1.141, requires isolation provisions for the lines penetrating the primary containment to maintain an essentially leaktight barrier against the uncontrolled release of radioactivity to the environment.
In Section 4.22.2 of the IPSAR, the staff requested that the licensee evaluate the leakage detection provisions and operating stations of the remote manual valves in the core spray and containment spray systems.
The subject valves are those listed in Section 4.22.2 of the IPSAR and valve V-21-15.
By letter dated August 27, 1985, the licensee provided a response to the staff's request.
In accordance with Section 4.22 of the IPSAR, the licensee evaluated the isolation provisions for the core spray and containment spray suction lines.
The licensee's evaluation concluded that the operating station for all valves in question is located in the 460-volt switchgear room, which is accessible during postaccident conditions.
In addition, should a failure occur in any of these lines outside the containment that would require the system to be isclated, alarms and indications, which include pressure and flow indications and sump pump operation alarms, would alert the operator.
In its safety evaluation dated August 20, 1986, the staff stated that it con-sidered that the means to detect the need to isolate these lines were adequate and that the licensee had committed to revise plant procedures to include op-crator actions for line isolation before restart from the Cycle 11 refueling outage.
On this basis, the staff concluded that the proposed procedural revisions will ensure that the core spray and containment spray systems can be isolated when the need arises so as to provide containment 'qtegrity.
Implementation will be verified as part of routine inspections.
After the staff's safety evaluation was issued, Region I staff reported that it had verified that the procedures had been implemented and that this matter had been discussed in Region I Inspection Report 50-219/87-22.
This issue is considered fully resolved.
2.15 Topic VII-1.A, Isolation of Reactor Protection System From Non-Safety Systems, Including Qualification of Isolation Devices (NUREG-0822, Section 4.27) 10 CFR 50.55a(h) through Institute of Electrical and Electronics Engineers (IEEE) Std. 279-1971 requires that safety signals be isolated from non-safety signals.
i 2.15.1 Flux Monitoring Isolation (NUREG-0822, Section 4.27(1))
l In Section 4.27(1) of the IPSAR, the staff concluded that insufficient isolation capability had been demonstrated between the nuclear flux monitoring system (intermediate range monitors (IRMs) and average power range monitors (APRMs))
and non-safety devices (process recorders and plant computer).
The licensee 2-18 l
agreed to perform a failure mode and effects analysis to evaluate the potential for common-mode electrical fault propagation.
This analysis was submitted on August 3, 1984.
In a letter dated October 23, 1984, to the licensee, the staff stated that it had reviewed the licensee's submittal and concluded that there was insufficient information to support the licensee's conclusion that the lack of qualified isolation devices would not compromise the integrity of the reactor protection system (RPS).
Specifically, the following information or justification was not included in the licensee's submittal:
(1) The evaluation did not address the resistor isolation buffer circuitry between the RPS and the process computer.
(2) The evaluation concluded that the probability of maximum recorder input voltage being applied across the recorder input signal terminals (or R-18) was negligible.
However, no justification was presented to support this conclusion.
J (3) The evaluation did not describe any periodic testing for stray voltages j
and system capability to withstand maximum credible voltages, as required by IEEE Std. 279-1971 and IEEE Std. 379-1977.
In the absence of such testing, redundancy does not provide sufficient protection.
In letters dated July 8, 1985, and April 4, 1986, the licensee addressed the outstanding issues.
The staff reviewed this information and in a letter dated November 10, 1987, it advised the licensee that it required additional informa-tion regarding (1) the isolation amplifier between the nuclear instrumentation analog signals and the multiplexer cabinet for the process computer and (2) the R105 IRM/APRM process recorder.
The staff will review the information when it is received from the licensee and document the results in a supplement to the IPSAR, l
2.16 Topic VII-1.B. Trip Uncertainty and Setpoint Analysis Review of Operating l
Data Base (NUREG-0822, Section 4.28) 10 CFR 50.36c.1.ii(A) requires that where limiting safety-system settings are specified for a variable on which a safety limit has been based, the setting should be chosen so that the automatic corrective action will correct the most severe abnormal event anticipated before a safety limit is exceeded.
In Section 4.28 of the IPSAR, the staff stated that sensors RE02A, B, C, and D (core spray and isolation on low-low reactor water level) had setpoints at the extreme low end of their ranges and that these setpoints should be increased to a point where the margin to extreme range was at least equal to the instru-ment accuracy, or the sensors should be replaced with those having different ranges more suitable for the limiting safety system setting.
4 Even though the staff concluded that setpoint drift was of low importance to risk, the licensee committed to install the GE analog trip system (which had been previously reviewed and approved by the staff in conjunction with the review of GE Topical Report NEDO-21617) during the Cycle 11 outage.
2-19
i In Region I Inspection Report No. 50-219/87-08, the staff stated that the licensee had installed analog trip systems in place of sensors RE02A, B, C, and D.
Because of concerns regarding Static 0-ring switches (see Office of Inspection and Enforcement Bulletin 86-02), the licensee is evaluating the replacement of other sensors with analog trip systems.
The staff will review the licensee's evaluation when it is received and report the results in a supplement to the IPSAR.
2.17 Topic VIII-3.B, DC Power System Bus Voltage Monitoring and Annunciation (NUREG-0822, Section 4.32) 10 CFR 50.55a(h), through IEEE Std. 279-1971, and 10 CFR 50 (GOC 2, 4, 5, 17, 18 and 19), as implemented by SRP Section 8.3.2, Regulatory Guides 1.6,1.29, 1.32, 1.47, 1.75, and 1.118, and Branch Technical Position ICSB-21 (NUREG-0800), require that the control room operator be given timely indication of the status of batteries and their availability under accident conditions.
In Section 4.32 of the IPSAR, the staff stated that the licensee had committed to install alarms for the B and C battery breaker open, C battery charger open, and C battery ground in the control room.
The staff concluded that these alarms were acceptable and that with the other battery indications listed above, the plant dc power system bus voltage monitoring and annunciation would meet current criteria.
The licensee was to provide a schedule to complete these modifications.
In its letters dated November 16 and 29, 1982, the licensee stated that the necessary modifications would be completed by the end of the Cycle 11R refuel-ing outage and that, as an interim measure, there would be periodic inspections of the battery systems after the Cycle 10R outage.
In its safety evaluation dated June 22, 1983, the staff, however, was concerned about the ability of the licensee to monitor the battery charging current with sufficient accuracy to ensure that the battery has a low resistance connection to the bus.
The staff noted that a current shunt that would provide for easy measurement of charging current might ce too large for full-load operation.
Therefore, the staff re-quested a description of how the battery connection integrity will be monitored by the instrumentation that will be part of the final modifications.
The licensee responded to the staff's concern by letters dated June 7, 1985, and April 4, 1986, and during the meeting at the site on June 16 and 17, 1986, on the status of licensing actions (meeting summary is dated August 1,1986).
)
In its SER dated December 16, 1986, the staff stated that it had reviewed the information provided by the licensee and that on the basis of IPSAR Sec-J tion 4.32 and the staff's safety evaluation dated June 22, 1983, it concluded
)
that the battery status alarms that were to be installed during the Cycle 11R outage were sufficient to ensure that de power system bus voltage monitoring and annunciation met current criteria.
The staff also concluded that, on the basis of the procedures provided by the licensee and its proposed check of the resistance through the breakers, this concern was resolved.
Region I Inspection Report No. 50-219/87-08 indicated that, with respect to the installation alarms for 8 and C battery breaker open, C battery charger open, 1
2-20
. l
i and C battery ground, the inspector had verified that the functions identified are alarmed in the control room.
The alarm annunciators do not always have the same designation as the function; however, a review of the alarm response pro-cedures verified that the functions are included in the alarm.
(See also Sec-tion 4.12 of this supplement.)
j On the basis of the above, the staff considers this SEP issue closed.
2.18 Topic IX-5, Ventilation Systems (NUREG-0822, Section 4.34) 10 CFR 50 (GDC 4, 60 and 61), as implemented by SRP Sections 9.4.1, 9.4.2, i
9.4.3, 9.4.4, and 9.4.5, requires that the ventilation systems shall have the capability to provide a safety environment for plant personnel and for engi-neered safety features.
2.18.1 Core Spray and Containment Spray Pump Ventilation (NUREG-0822, Section 4.34(3))
The core spray and containment spray pumps are located in two corner rooms within the reactor building.
These rooms do not have. specific area ventilation systems.
Therefore, the staff concluded in Section 4.34(3) of the IPSAR that the licensee should demonstrate that these pump motors were qualified for the conditions that could be expected in these rooms following a loss-of-coolant
)
accident (LOCA) or make the appropriate plant modificatiuns.
In a letter dated September 1, 1983, the licensee stated that the core spray and containment spray pump motors are designed to function in environments with temperatures up to 185*F and 203 F, respectively.
In Amendment 42 to the Oyster Creek Unit 1 facility description and Safety Analysis Report, the i
licensee had calculated that the maximum post-LOCA temperature expected in the corner rooms without ventilation would be approximately 173'F.
In its safety evaluation dated April 26, 1984, the staff concluded that, on the i
basis of the thermal capability of the core spray and containment spray pump motors compared to maximum calculated room thermal conditions, provisions for ventilation are not necessary.
Therefore, this SEP issue is resolved.
2.18.2 Battery, Motor Generatar, and Switchgear Room Ventilation (NUREG-0822, Section 4.34(4))
t In Section 4.34(4) of the IPSAR, the staff found that both the B battery and i
motor generator room and the switchgear room ventilation systems were manually actuated from the control room by energizing a single relay.
Transfer of this single control relay (relay K) applies power to both the supply and exhaust fans in each room.
Thus, a failure of that relay to transfer or loss of power to that relay would preclude electrical power to the fans of each room.
The licensee agreed to evaluate the ventilation system design for the B battery and motor generator room and the consequences of a loss of ventilation in the switchgear room.
By letter dated August 21, 1984, the licensee provided the results of its eval-uation of the B battery and motor generator room and switchgear room ventila-tion systems.
The licensee also committed to instell redundant relays to en-sure adequate ventilation in these areas during the Cycle 11 refueling outage.
2-21
In its safety evaluation dated July 1, 1985, the staff concluded that, on the basis of its review of the licensee's evaluation and the resulting comitment to install redundant relays for these ventilation systems, this SEP issue is considered resolved.
Region I staff will verify that the redundant relays have been installed.
k b
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i 2-22 i
i
I Table 2.1 Intcgrated assessment summary l
SEP IPSAR Supplement Topic Section IPSAR Supplement Requirement >/
No.
No.
Titia Requirements Section No.
Status II-3.8, 4.1(1)
Condensate Water See IPSAR 2.1.1 Under review II-3.B.1, Pumps Item 4.6.4.
II-3.C 4.1(2)
Flooding Level None Procedures 4.1(3)
Canal Water Level Install water level 2.1.2 See IPSAR Instrumentation instrumentation in Section 4.1(5) intake canal.
(supplement Section 3.1.2).
4.1(4)
Isolation Condenser Demonstrate minimum 3.1.1 See IPSAR Flooding quantity of water 4.1.1 Section 4.1(1) maintained in con-(supplement densate storage tank Section 2.1.1).
m4 sufficient for long-term cooling and include w
minimum inventory in plant procedures.
4.1(5)
Low Water Level Shutdown None 3.1.2 Under review 4.1(6)
Hurricane Flooding of Revise emergeacy 4.1.2 Resolved Pumps procedures to identify alternate sater sources and flow paths should low elevation pumps be flooded.
4.1(7)
Protection During Internal Evaluate consca. maces of 2.1.3 Region I to Flooding offgas building i ading verify and confirm all (' cr entrance levels Wove 23.5 ft.
-- ~..
Table 2.1 (Continued)
SEP IPSAR Supplement Topic Section IPSAR Supplement Requirements /
No.
No.
Title Requirements Section No.
Status II-3.B, 4.1(8)
Groundwater Elevation See IPSAR Item 4.4(2).
2.4.1 Resolved II-3.B.1, i
II-3.C 4.1(9)
Roof Drains Install scuppers in the 4.1.3 Resolved reactor building and turbine b':ilding parapets.
III-1 4.2 Classification of Struc-Evaluate design of 2.2 Submit tures, Components, and specified components information Systems on a sampling basis, for staff upgrade if necessary, resiew.
and document classi-fication in FSAR update.
III-2 4.3.1 Reactor Building Steel Analyze and-identify 2.3.1 Submit "f
Structure Above the any needed upgrading of information "4
Operating Floor reactor building upper for staff steel structure for review.
wind loads.
4.3.2 Ventilation Stack Analyze and identify 2.3.2 Resolved any needed upgrading of ventilation stack for wind loads.
4.3.3 Effects of Failure of Analyze turbine building 2.3.3 Resolved Nonseismic Category I capacity for wind loads, Structures evaluate consequences of failure and identify any needed upgrading.
4.3.4 Components Not Enclosed None in Qualified Structures
Table 2.1 (continued)
SEP iPSAR Supplement Topic Section IPSAR Supplement Requirements /.
No.
No.
Title Requirements Section No. Status III-2 4.3.5 Exterior Masonry Walls None 4.3.6 Roof Decks Provide analysis of 2.3.4 See IPSAR reactor building roof.
Section 4.3.1 (supplement Section 2.3.1).
Analyze capacity of turbine bui! ding roof to withstand wind loads.
4.3.7 Intake Structure, Oil Analyze capacity to 2.3.5 Under review Tanks, and Diesel Gener-withstand wind and ator Building tornado loads and upgrade, if necessary.
U 4.3.8 Load Combinations See IPSAR Item 4.12.
2.3.6 See IPSAR Section 4.12 (supplement Section 2.10).
4.3.9 Soil and Foundation None Capacities 4.3 Control Room /
- 2. 3.7/
Submit Architectural 2.3.8 evaluation /
Components submit evaluation.
III-3.A 4.4(1)
Hydrostatic Loads None (Combination) 1
=.-
Tcble 2.1 (Continued)
SEP IPSAR Supplement Topic Section IPSAR Supplement Requirements /
No.
No.
Title Requirements Section No.
Status
'II.3.A 4.4(2)
Hydrostatic Loads Evaluate short-2.4.1 Resolved (Short-Duration) duration hydrostatic loads on and flota-tion potential of structures essential to safe shutdown in conjunction with flooding emergency procedures (IPSAR Item 4.1(6)).
4.4(3)
Below-Grade Penetration None Flooding III-3.C 4.5.1 Intake and Discharge None s,
g, Canals cn 4.5.2 Intake Structure Trash Formalize existing 4.2.1 Resolved Racks and Intake Screens inspection practice as part of shift turnover or inservice inspection (ISI) procedures until water level modifica-tion is complete (IPSAR Item 4.1(3)).
4.5.3 Roof Drains See IPSAR Item 4.1(2).
None (Resolved) 4.5.4 Inspection Program Develop and implement 4.2.2 Resolved a formal inspection program for water control structures.
Table 2.1 (Continued)
SEP IPSAR Supplement Topic Section IPSAR Supplement Requirements /
No.
No.
Title Requirements Section No.
Status III-4.A 4.6.1 Emergency Diesel Analyze potential for 2.5.1 See IPSAR Generators and Fuel and consequences of Section 4.3.7 Oil Day Tank tornado-missile damage (supplement of the diesel generator Section 2.3.5).
building.
4.6.2 Mechanical Equipment Evaluate the potential 2.5.2 Under review Access Area for and consequences of tornado-missile impact in the reactor building access door region and identify any necessary corrective actions.
4.6.3 Control Room, Reactor None m4 Building, and Turbine
.i Building Heating, Ventilating, and Air Conditioning (tNAC)
Systems 4.6.4 Condensate Storage Tank, Provide protection for 2.5.3 See IPSAR Torus Water Storage Tr% sufficient systems and Section 4.1(1) and Service Water and components to ensure a (supplement Emergency Service safe shutdown in the Section 2.1.1).
Water Pumps event of damage from tornado missiles.
Table 2.1 (Continued)
SEP IPSAR Supplement Topic Section IPSAR Supplement Requirements /
No.
No.
Title Requirements Section No.
Status III-4.B 4.7 Turbine Missiles Inspect turbine and 2.6 Resolved propose inspection frequency based on results.
Justify monitoring 2.6 Resolved program for main steam and reheat control valves.
111-4.0 4.8.1 Truck Explosion None 4.8.2 Aircraft Hazards Evaluate potential for 2.7.1 Resolved or consequences of aircraft impact.
m E
III-5.A 4.9(1)
Cascading Pipe Breaks See IPSAR Item 4.16.
See IPSAR Section 4.16 (supplement Section 2.12).
4.9(2)
Jet Impingement Effects None 4.9(3)
Drywell Penetration None III-5.B 4.10(1)
LOCA Outside Containment None 4.10(2)
Emergency Condenser Evaluate and identify 2.8.1 Submit Isolation any necessary modifica-information tions to provide. leakage for staff detection to ensure review.
that flaws would be detected before pipe break occurs.
Table 2.1 (Continued) i SEP IPSAR Supplement Topic Section IPSAR Supplement Requirements /
i No.
No.
Title Requirements Section No.
Status III-6 4.11(1)
Piping Systems Analyze on a sampling.
2.9.1 Under review basis and verify adequacy of support designs for the seismic resistance of specified piping systems.
4.11(2)
Mechanical Equipment Demonstrate that the 2.9.2 Under review control rod drive system and vessel internals have suffi-cient capacity to resist a safe shutdown earthquake or take-corrective action.
m 4.11(3)
Electrical Equipment Reevaluate 4160-V 2.9.3 Under review-switchgear panel anchorage and demon-strate, on a sampling basis, adequacy of electrical panel supports.
4.11(4)
Ability of Safety-None Related Electrical Equipment To Function 4.11(S)
Qualification of Cable Provide plan to imple-2.9.4 Submit Trays cant results of SEP information Owners Group Program' for staff on a plant-specific review.
basis.
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u Table 2.1 (Continued)
SEP IPSAR Supplement Topic Section IPSAR Supplement Requirements /
No.
No.
Title Requirements Section No.
Status III-7.B 4.12 Design Codes. Design Evaluate adequacy of ori-2.10 Submit Criteria, Load Combina-ginal design criteria on a information tions, and Reactor Cavity sampling basis for speci-for staff Design Criteria fied structural elements.
review.
III-8.A 4.13 Loose-Parts Monitoring None and Core Barrel Vibration Monitoring III-10.A 4.14(1)
Thermal-Overload Bypass Evaluate thermal-overload 2.11.1 Resolved bypasses for engineered safety features (ESF) valves.
4.14(2)
Magnetic Trip Breakers None 0
IV-2 4.15 Reactivity Control None Systems, Including Functional Design and Protection Against Single Failures V-5 4.16.1 Leakage Detection Systems Evaluate reliability of 2.12.1 Install air-leakage detection borne particu-systems and evaluate late and gas-sensitivity in conjunc-eous radiation tion with Topic III-5.A monitoring sys-analysis.
tem; Region I to verify.
4.16.2 Operability Requirements Identify action for loss 3.2 Resolved.
of leakage detection 4.3.1 Region I to in Technical Specifica-verify.
tions and include testing in procedures.
Table 2.1 (Continued)
SEP IPSAR Supplement Topic Section IPSAR Supplement Requirements /
No.
No.
Title Requirements Section No.
Status V-5 4.16.3 Intersystem Leakage None 4.16.4 Reactor Coolant None Inventory Balances V-6 4.17 Reactor Vessel Integrity Submit a plan for the 3.3 Resolved material surveillance capsules.
V-10.8 4.18 Residual Heat Removal Review and upgrade, if 4.4 See IPSAR Sec-System Reliability necessary, shutdown tions 4.1(1),
procedures to specify 4.1(4), 4.6.4, alternate sources of and 4.30.1 I
water for primary and (supplement secondary makeup, with Sections 2.1.1, J,
particular attention to 3.1.1, 4.1.1, external events.
2.5.3, and 4.10).
V-11.A 4.19 Requirements for Isola-Demonstrate relief 2.13 Resolved tion of High-and Low-capacity and accept-Pressure Systems able consequences, or identify corrective action to protect reactor water cleanup system.
V-12.A 4.20 Water Purity of BWR Implement proposed 3.4 Resolved Primary Coolant procedure and modify 4.5 Technical Specifica-tions to be consistent.
VI-1 4.21.1 Organic Materials Inspect and repair, if 4.6.1 Resolved necessary, drywell coatings and recoat the torus.
Table 2.1 (Continued)
SEP IPSAR Supplement Topic Section IPSAR Supplement Requirements /
No.
No.
Title Requirements Section No.
Status VI-1 4.21.2 Postaccident Chemistry None VI-4 4.22.1 Locked-Closed Valves Provide physical locking 4.7.1 Resolved devices to ensure valves are not inadvertently opened.
4.22.2 Remote Manual Valves Evaluate leakage 2.14.1 Resolved detection provisions and, if necessary, relocate the operating station for isolation valves in the containment spray and core spray systems.
n, 4.22.3 Valve Location None 4.22.4 Instrument Lines None 4.22.5 Valve Location and Type None 4.22.6 Administrative Controls None VI-7.A.3 4.23 Emergency Core Cooling Include emergency 3.5 Resolved System Actuation System condenser logic testing in the Technical Specifications.
VI-7.A.4 4.24 Core Spray Nozzle None Effectiveness
Table 2.1 (Continued)
SEP IPSAR Supplement Topic Section IPSAR Supplement Requirements /
No.
No.
Title Requirements Section No.
Status VI-7.C.1 4.25(1)
AC Automatic Bus Evaluate the existing 4.8.1 Region I to Transfers automatic bus transfers verify.
and identify corrective actions to ensure faulted loads would not be transferred.
4.2S(2)
DC Automatic Bus None Transfers VI-10.A 4.26.1 Responte-Time Testing None 4.26.2 Instrumentation for Verify all safety logic 3.6.1 Resolved Reactor Trip System channels tied to the (RTS) Testing reactor mode switch are n,
J, tested by procedure.
w Include logic channel 3.6.1 Resolved testing in Technical Specifications.
4.26.3 Dual-Channel Testing None VII-1.A 4.27(1)
Flux Monitoring Isolation Perform failure mode and 2.15.1 Submit effects analysis to information determine whether isola-for staff tion devices are required review.
and identify any needed upgrading.
4.27(2)
Reactor Protection Install Class IE 4.9.1 Resolved System (RPS) Protective protection at the RPS Trip power supply and RPS interface.
_ _ =
Table 2.1 (Continuea)
SEP IPSAR Supplement Topic Section IPSAR Supplement Requirements /
No.
No.
Title Requirements Section No.
Status VII-1.8 4.28 Trip Uncertainty and Install analog trip 2.16 Submit Setpoint Analysis system.
information Review of Operating for staff Data Base review.
VII-2 4.29 Engineered Safety See IPSAR Item 4.14(1).
2.11.1 Resolved Features System Control Logic and Design VII-3 4.30 Systems Required for Provide minimum inventory 4.10 See IPSAR Sec-Safe Shutdown for condensate storage tions 4.1(4),
tank as a water source for 4.1(6), and flooding events (IPSAR 4.18 i
Item 4.1(4)) and identify (supplement non-ESF equipment in Sections 3.1.1, 4
cooldown procedures 4.1.1, 4.1.2, (IPSAR Item 4.18).
and 4.4).
VIII-2 4.31(1)
Diesel Generator Modify annunciators to 4.11.1 Resolved Annunciators conform to IEEE Std. 279-1971.
~!
4.31(2)
Diesel Generator Trip Evaluate bypass of two 4.11.2 Region I to Bypass trips (voltage-ampere verify reactive and reverse power) during accident conditions.
r VIII-3.B 4.32 DC Power System Bus Schedule installation of 2.17 Resolved Voltage Monitoring and specified battery status 4.12 Annunciation alarms.
VIII-4 4.33 Electrical Penetrations None of Reactor Containment
- =-
Tt:ble 2.1 (Continued)
SEP IPSAR Supplement Topic Section IPSAR Supplement Requirements /.
No.
No.
Title Requirements Section No.
Status IX-5 4.34(1)
Restoration of Evaluate and revise, if 3.7.1 Submit Ventilation necessary, the loss-of-4.13.1 information 2
offsite power procedures for staff to ensure that restora-review.
tion of ventilation I
systems will not over -
load the diesels.
4.34(2)
Reactor Building None Ventilation
)
4.34(3)
Core Spray and Contain-Demonstrate subject 2.18.1 Resolved ment Spray Pump pumps can operate with Ventilation a loss of ventilation, or identify corrective J,
action, as necessary.
w 4.34(4)
Battery, Motor Generator, Evaluate effects of loss 2.18.2 Region I to and Switchgear Room of ventilation to the verify Ventilation subject rooms and identify any needed upgrading.
a XV-1 4.35 Decrease in Feedwater.
None Temperature, Increase in Feedwater Flow, and Increase in Steam Flow and Inadvertent Opening of a Steam Ger=->'nr Relief or Safety valve 4
.______________._.____.___..______s,
..,. ~_.,v.
.,_,.,_.,,__.,,.,n
.,,..v.,_
y
Table 2.1 (Continued)
SEP IPSAR Supplement Topic Section IPSAR Supplement Requirements /
No.
No.
Title Requirements Section No.
Status XV-16 4.36 Radiological Consequences Implement BWR Standard 3.8 Submit of Failure of Small Technical Specifica-information Lines Carrying Primary tion limits for primary for staff Coolant Outside coolant activity.
review.
Containment XV-18 4.37 Radiological Con-See IPSAR Item 4.36.
See IPSAR secuences of a Main Section 4.36 Steam Line Failure (supplement Outside Containment Section 3.8).
XV-19 4.38 Loss-of-Coolant Accidents Develop and implement a 3.9 Resolved Resulting From Spectrum preventive maintenance of Postulated Pipe program for the main Breaks Within the steam isolation valves, ma Reactor Coolant or justify existing Pressure Boundary maintenance based on operating experience.
Submit results of 3.9 Resolved evaluation including testing experience.
l
,.e-c-
-4.---,r w
,&a-,-
e-.-
-.m
-..e
3 TOPICS RESOLVED BY CHANGES TO PLANT TECHNICAL SPECIFICATIONS OR PROCEDURES During the integrated assessment for Oyster Creek, a number of issues were resolved by commitments from the licensee to perform evaluations in order to determine whether modifications to plant Technical Specifications are warranted.
l The following sections describe the actions taken regarding the resolution of IPSAR issues involving Technical Specifications or procedural changes.
3.1 Topic II-3.B, Flooding Potential and Protection Requirements; Topic II-3.B.1, Capability of Operating Plants To Cope With Design-Basis Flooding Conditions; Topic II-3.C, Safety-Relatea Water Supply (Ultimate Heat Sink (UHS)) (NUREG-0822, Section 4.1)__
10 CFR 50 (GDC 2), as implemented by SRP Sections 2.4.2, 2.4.5, 2.4.10, and 2.4.11 and Regulatory Guides 1 59 and 1.27, requires that structures, systems, and components important to safety be designed to withstand the effects of nat-ural phenomena such as flooding.
The safety objective of these topics (II-3.B.
II-3.B.1, and II-3.C) is to verify adequate operating procedures and/or system design provided to cope with the design-basis flood.
The site grade elevation is 23 feet mean sea level (MSL).
During the staff's review of the hydrology related topics, the following flooding elevations were identified, as defined by current licensing criteria:
probable maximum hurricane (PMH) - 22 feet MSL probable maximum precipitation (PMP) - 23.5 feet MSL As a result of these flooding levels, the staff identified the issues discussed in the following sections.
3.1.1 Isolation Condenser Flooding (NUREG-0822, Section 4.1(4))
In Section 4.1(4) of the IPSAR, the staff stated that because the makeup sources for the isolation condensers were susceptible to a single failure and flooding, the plant did not have a reliable means for maintaining a safe shutdown.
In addition, the makeup water sources to the isolation condenser should be identi-fied to ensure that the best quality of water is available.
The staff also stated that the licewea would evaluate the need for redundancy for the condensate transfer pump power supply.
In addition, the plant emergency procedure for flooding would be revised to include the fire water storage tank as a redundant source of water supply to the emergency condenser.
(The eleva-tion of the fire water storage tank is above the PMH flood level.) The licensee agreed to demonstrate that the minimum quantity of water maintained in the con-densate storage tank and in the fire water rtotage tank was sufficient for long-term cooling, using either tank.
The fire water storage tank is a backup to the condensate storage tank.
3-1
1 J
l The licensee also proposed to include minimum inventory of the water to be maintained in the condensate storage tank in the operating procedures and i
j designate this as the primary source.
In its letters dated July 26, 1985, and April 21, 1986, the licensee stated l
that makeup water to the isolation condensers is provided by the condensate j
storage tank and the fire water storage tank.
These tanks can provide a vol-ume of water of nearly 1 million gallons, which should be sufficient to main-l l
tain the reactor in hot shutdown using the isolation condensers for 10 days.
i This time is sufficient to take corrective actions to restore submerged compo-i
{
nents.
The tanks and the pumps are above the probable maximum hurricane flood-l ing level of 22 feet MSL.
l J
The licensee explained in the meeting of June 16 and 17, 1986 (meeting summary i
i dated August 1,1986), that its procedures require a minimum of 20 feet or
)
250,000 gallons in the condensate storage tank (CST).
The intake tour sheet requires a minimum of 350,000 gallons in the fire water storage tank.
The high i
wind conditions for Emergency Procedure 2000-ABN-3200.31 are the following:
l (1) tornado watch or warning, (2) hurricane watch or warning, (3) tornado funnel cloud in the area, and (4) sustained wind speeds greater than 74 mph.
This pro-cedure requires the CST to be filled to 43 feet or 537,500 gallons and the iso-i lation condensers to be filled (50,000 gallons).
In its safety evaluation dated I
l November 28, 1986, the staff stated that the licensee had stated that it could, if needed, bring in a fire truck and pump water into the isolation condensers i
]
using an alternate connection from the fire water main.
i y
l This completes the staff review of this SEP issue.
The fact that the condensate transfer pumps are located outside the plant and are susceptible to damage from i
1 tornado missiles is discussed in Section 2.
The procedures are verified by Region I staff as indicated in Section 4.1.1.
i l
3.1.2 Low Water Level Shutdown (NUREG-0822, Section 4.1(5))
{
l In Section 4.1(5) of the IPSAR, the staff stated that a Technical Specification
)
change was under review that would allow the operator to stop operation of the i
dilution pump when the level was low and thus raise the water level in the in-i take canal.
The dilution pumps were required to continue running under certain
}
conditions to maintain canal water temperature within limits.
This Technical Specification change along with a water level gage (with remote readout in the j
control room) (Section 4.1(3)) in the intake canal would enable the operator 1
j to respond in a timely manner to the low water level in the canal.
I In Amendment 66, issued on March 24, 1983, the Technical Specifications related l
1 j
to water quality were removed from the Oyster Creek license; therefore, the j
intent of the above change was met.
However, in a letter dated April 21, 1986, j
the licensee requested to cancel its commitment to install an automatic water 1
level gage in the intake canal with a remote readout in the control room.
As stated in Section 2.1.2 of this supplement, the licensee proposed revising 2
Station Procedure 2000-ABN-3200.31 "High Winds," to require a plant shutdown when the water level at the intake structure cannot be verified to be less
[
4
]
than elevation 4.5 feet M5L.
l 3
j 3-2
?
..r
_ _. _ = _ _ _. _
_ _ _, __ i
In a letter dated November 28, 1986, the staff advised the licensee that its submittal of April 21, 1986, did not address the use of this instrumentation for measuring the water level at the intake structure to determine if it is near or below the service water pump suction elevation.
The staff also stated that this was discussed in Section 4.1(5) of the IPSAR and requested that the licensee provide a discussion of the administrative controls to monitor the canal water for low water level near or below the service water pump suction elevation and the actions to be taken by the control room operators.
In a letter dated November 6, 1987, the licensee provided the information requested.
The staff is evaluating this information.
When it completes its review, the staff will report the results in a supplement to the IPSAR.
3.2 To3ic V-5, Reactor Coolant Pressure Boundary (RCPB) Leakage Detection (NJREG-0822, Section 4.16) 3.2.1 Operability Requirements (NUREG-0822, Section 4.16.2)
In Section 4.16.2 of the IPSAR, the staff stated that the Oyster Creek Technical Specifications did not contain limiting conditions for operation or surveillance requirements regarding the leakage detection systems, as recommended by Regulatory Guide 1.45 and the Boiling-Water Reactor (BWR) Standard Technical Specifications (NUREG-0123).
The staff also stated that, in conjunction with the procedural changes described in Section 4.16.1, the licensee had committed to provide the appropriate action requirements in the Technical Specifications for inoperable leakage detection systems (i.e., an inability to measure leakage) and any necessary procedural changes to provide surveillance and testing commensurate with the required sensitivity.
By letter dated August 23, 1985, which superseded its letter dated October 22, 1984, the licensee requested an amendment to the Oyster Creek Technical Speci-fications that would revise the limiting conditions for operation and add sur-veillance requirements for the reactor coolant system leakage.
In its letter of January 6, 1986, the staff issued an amendment to the Oyster Creek Provisional Operating License that authorized the following changes to the Technical Specifications: the revision of the limiting conditions for oper-ation and the addition of surveillance requirements for reactor coolant system leakage.
In its letter the staff found that the proposed specifications regard-ing the reactor coolant leakage and leakage detection systems were more restric-tive than the current specifications and, with one e m eption, were consistent with the BWR Standard Technical Specifications.
The exception was that the pro-posed specifications did not include a limiting condition for cperation (LCO) in regard to pressure boundary as do the BWR Standard Technical Specifications.
The staff recommended that the licensee submit an additional license amendment application that included an LCO in regard to pressure boundary leakage in Technical Specification 3.3.0.
The staff also advised the licensee that its letter closed out the staff's actions in regard to the issue identified in Sec-tion 4.16.2 of the IPSAR.
3-3
Office of Inspection and Enforcement Inspection Report No. 50-219/87-08 identi-fled inadequate procedural implementation of the Technical Specification.
Reso-lution of this aspect of the item is discussed in Section 4.3.1.
On March 17, 1987, the licensee submitted a Technical Specification change re-quest that would limit the unidentified leakage for the reactor coolant system to a maximum leak rate increase of 2 gpm within any 24-hour period during oper-ation at steady-state power.
The staff is reviewing the licensee's request l
independently of the SEP.
3.3 Topic V-6, Reactor Vessel Integrity (NUREG-0822, Section 4.17)
Appendices G and H to 10 CFR 50 and 10 CFR 50-55(g), as implemented through 4
Regulatory Guide 1.99, require that reactor vessel integrity be ensured by re-i view of aspects such as fracture toughness, surveillance programs, and neutron irradiation.
In Section 4.17 of the IPSAR, the staff determined that the licensee should submit a plan for the capsule exposure schedule and how the test results will be used to modify operation (e.g,, setting nil ductility temperature limits).
By letter dated March 10, 1983, the licensee provided information regarding the reactor vessel material surveillance program at Oyster Creek.
That letter indicated that the No. 2 material capsule would be removed during the Cycle 10 refueling outage and testing and analysis results would be provided to the NRC staff.
In addition, the licensee provided the schedule for removal of the next
]
capsule.
In a letter dated April 28, 1983, the staff determined that the licensee's letter of March 10, 1983, constituted an acceptable response to the issue identified in Section 4.17 of the IPSAR.
The staff also concluded that staff review of the capsule test results will be conducted as a routine operating reactor action independently of the SEP.
The licensee's response is considered sufficient to close out this SEP issue.
3.4 Topic V-12.A, Water Purity of BWR Primary Coolant (NUREG-0822, Section 4.20) 10 CFR 50 (GDC 14), as implemented by Regulatory Guide 1.56, requires that the reactor coolant boundary (RCPB) have minimal probability of propagating failure.
This includer corrosion-induced failures from impurities in the reactor cociant i
system.
In Section 4.20 of the IPSAR, the staff concluded that the licensee should pro-vide Technical Specification changes to incorporate pH level conductivity and chloride limits and implement a time-related conductivity limit in operating procedures, The licensee agreed to complete these actions before startup from the Cycle 10 outage.
As a result, the licensee, in a letter dated September 18, 1984, proposed to revise the Technical Specifications regarding chlorides and conductivity to ensure consistency with Regulatory Guide 1.56, i
In its safety evaluation dated November 21, 1985, the staff concluded that the proposed Technical Specification changes regarding reactor water conductivity 3-4
and chloride concentration limits met the limits and appropriate corrective actions established in Regulatory Guide 1.56 and were therefore acceptable.
On this basis, the staff issued Amendment 93 to Provisional Operating License No. OPR-16 for Oyster Creek on November 21, 1985.
This amendment authorized the following changes to the Oyster Creek Technical Specifications: incorpo-ration of the additional restrictions on conductivity and chloride limits in Sec-tion 3.3.E, "Reactor Coolant Quality," and revision of its Basis.
However, in its safety evaluation, the staff also concluded that the application to amend the Technical Specifications did not address the guideline in footnote "a" of Table 1 of Regulatory Guide 1.56, whNh states that the total time for all incidents during which the acceptable reactor water chemistry limits in Table 1 are exceeded should not exceed 2 weeks per year.
The staff considers this re-striction on plant operation to be a necessary part of the method (described in Regulatory Guide 1.56 and acceptable to the staff) for implementing the cri-teria in General Design Criterion 14 with regard to minimizing the probability of corrosion-induced failure of the reactor coolant boundary in BWRs.
This restriction is in the BWR Standard Technical Specifications (NUREG-0123).
On this basis, the staff requested that the licensee propose the incorporation of such a restriction in the Technical Specification, or justify why such a Tech-nical Specification was not needed.
At a meeting on February 6, 1987, the licensee stated that the controls on the reactor coolant quality in Specification No. 1302-28-001, Revision 2, provided the additional requirement requested by the staff in its letter issuing Amend-ment 93.
This specification restricts Oyster Creek operation when the limit is exceeded 2 weeks in any consecutive 12-month period.
In Inspection Report No. 50-219/87-08, the staff stated that the NRC inspector had verified that the time-related conductivity limit, the chloride concentra-tion limit, and the pH limit for reactor coolant specified in Specification No. SP-1302-28-001 had been incorporated into Station Procedure 827.1, "Primary System Analysis; Reactor Water." Also, Amendment 93 to the license incorpo-rated into the Technical Specifications the chloride and conductivity limits established in Regulatory Guide 1.56.
The implementation of this Technical Specification has been verified by NRC inspection (Impection Report No.
50-219/87-08) as discussed in Section 4.5.
Therefore, this SEP issue is fully resolved.
3.5 Toaic VI-7.A.3, Emergency Core Cooling System Actuation System (NJREG-0822. Section 4.23) 10 CFR 50.55a(h), as implemented by Institute of Electrical and Electronics Engineers Std. 279-1971, and 10 CFR 50 (GDC 37), as implemented by Regulatory Guide 1.22, require that equipment important to safety be tested periodically at power.
In Section 4.23 of the IPSAR, the staff concluded that testing of the logic trains and associated components of the emergency condenser should be included in the Technical Specifications.
By letter dated June 4, 1984, the licensee provided the results of its review of the Oyster Creek Technical Specifications.
The licensee concluded that modification of the existing Technical Specifications concerning the testing of the emergency condenser logic trains was not warranted.
3-5
1 l
The plant parameters that actuatu the emergency condenser are reactor high pressure and low-low reactor water level; the testing and calibration of these instrument channels is included in Technical Specification Table 4.1.1.
Simic larly, the instrument channels that detect and isolate an emergency condenser line break (high-ficw-differential pressure) are also included in that table.
Section 4.8 and Table 4.1.2 of the Technical Specifications require a test of the emergency condenser actuation and isolation trip system at each refueling i
outage.
This frequency i. consistent with that required for other protective instrumentation and trip systems at Oyster Creek.
In addition, the licensee is implementing a reference index that will cross-reference the Technical Specification surveillance requirements.
The above is acceptable to the staff and was documented in the staff's safety evaluation dated July 1, 1985.
This closes out this SEP issue.
3.6 Topic VI-10.A, Testing of Reactor Trip System and Engineered Safety Features, Including Response-Time Testing (NUREG-0822, Section 4.26) 10 CFR 50 (GDC 21) requires that the reactor protection system be designed to permit periodic testing of its functioning, including a capability to test channels independently.
3.6.1 Instrumentation for Reactor Trip System (RTS) Testing (NUREG-0822, Section 4.26.2)
In Section 4.26.2 of the IPSAR, the staff found that the reactor mode switch and some instruments were not specifically identified for testing in the Technical Specifications.
The licensee stated that plant procedures require testing of all redundant instrumentation required for safety.
However, the licensee agreed to review plant surveillance procedures to ensure that all safety logic channels tied to the reactor mode switch were surveyed.
In additinn, changes to the Technical Specifications would be made to incorpo-3 j
rate the required testing before startup from the Cycle 10 outage.
By letter dated May 31, 1984, the licensee provided the results of its review of protective instrumentation testing required by either plant procedure or i
Technical Specifications.
The licensee noted that although an explicit test of the reactor mode switch was not specified in the Technical Specifications, the switch was tested in various positions when other logic channels were tested. of the May 31, 1984, submittal showed for each contact in the switch, tests that sere performed, associated Technical Specification requirements, and which position (run, shutdown, etc.) the switch was in.
In the staff's safety evaluation
]
dated July 15, 1985, and Inspection Report No. 50/219/87-08 related to this matter, the staf f concluded that the testing being performed was sufficient to test the functioning of the mode switch, j
The staff also stated that the licensee +-
w loped a cross-reference indexing r
system between the Technical Specificatt int h rveilla c procedures.
i t
This index will show the correlation between the testing of instrumentation and logic channels and trip systems using the surveillance procedures and the Technical Specification requirements.
On this basis, the staff concluded that the existing TS requirements, as supplemented by the cross-reference indexing system, should be adequate to ensure needed testing is performed.
Hnwever, the staff requested that the licensee submit the index to the staff.
On October 22, 1985, the cross-reference index was submitted.
In a letter dated January 17, 1986, the staff advised the licensee that it had reviewed the index and noted that it contained tables listing the Technical Specification require-ments for surveillance, the actual frequency as specified in the procedure, the procedure number, and the schedule for the test.
In its letter, the staff stated that, on the basis of its review of the submittal dated October 22, 1985, it concludes that the indexing system is sufficiently detailed and complete to sat-isfy the issue raised by the staff in the IPSAR review.
This closes out this SEP item.
3.7 Topic IX-5, Ventilation Systems (NUREG-0822, Section 4.34) 10 CFR 50 (GDC 4, 60, and 61), as implemented by SRP Sections 9.4.1, 9.4.2, 9.4.3, 9.4.4, and 9.4.5, requires that the ventilation systems sha.ll have the capability to provide a safe environment for plant personnel and for engineered safety features.
3.7.1 Restoration of Ventilation (NUREG-0822, Section 4.34(1))
The licensee will provide a submittal as discussed in IPSAR Section 4.13, 3.8 Topic XV-16, Radiological Consequences of Failure of Small Lines Carrying Primary Coolant Outside Containment (NUREG-0822, Section 4.36) 10 CFR 100, as implemented by SRP Section 15.6.2, requires that the radiological consequences of f ailure of small lines carrying primary coolant outside con-tainment be limited to small fractions of the exposure guidelines of 10 CFR 100.
In Section 4.36 of the IPSAR, the staff concluded that the Technical Specifi-cations should be modified to include the BWR Standard Technical Specification (NUREG-0123) reactor coolant activity limits, sampling frequencies, and action requirements.
By letter dated October 22, 1984, the licensee proposed a revised Technical Specification for Oyster Creek for primary coolant activity.
The staff re-viewed the licensee's submittal and requested additional information concern-ing (1) the definition for dose equivalent iodine-131, (2) limits for non-iodine radioactivity in the reactor coolant as shown in the BWR Standard Technical Specifications (NUREG-0123), (3) the limiting condition for operation as shown in NUREG-0123, and (4) the annual reporting requirement for radioiodine spiking as shown in NRC Generic Letter 85-19, "Reporting Requirements on Primary Coolant Iodine Spikes."
3-7 l
In a letter dated October 23, 1986, the licensee provided the requested infor-mation and proposed a Technical Specification change.
The staff reviewed this i
information and in a meeting on June 30, 1987, advised the licensee to revise its proposed Technical Specification in regard to the wording of definitions i
and sampling frequencies.
The licensee is revising the proposed Technical Specification in accordance with staff requirements. When the licensee's submittal is received, the staff 1
will review the information and do:ument the results in a supplement to the IPSAR.
3.9 Topic XV-19, Loss-of-Coolant Accidents Resulting From Spectrum of i
Postulated Pipe Breaks Within the Reactor Coolant Pressure Boundary (NUREG-0822, Section 4.38) 10 CFR 100, as implemented by SRP Section 15.6.5, requires that the radiologi-cal consequences of a design-basis loss-of-coolant accident be limited to the exposure guidelines for both the 0- to 2-hour exclusion area boundary and the j
30-day low population zone (LPZ) boundary.
i In Section 4.38 of the IPSAR, the staff found that the major contribution to 1
the 30-day LPZ exposure was from main steam line isolation valve (MSIV) leak-j age.
Therefore, the staff required that the licensee develop and implement a preventive maintenance program to limit MSIV leakage or justify the existing i
program on the basis of testing results for the valves.
i The licensee proposed to review existing maintenance practices and those of i
other BWR facilities, identify any necessary corrective actions, and upgrade the maintenance program, if necessary, before startup from the Cycle 11 refuel-t i
ing outage.
The licensee responded to IPSAR Section 4.38 in its submittals j
dated May 18, 1984, and September 12, 1985.
The staff has evaluated the infor-mation provided by the licensee and documented its findings in its safety eval-J uation dated May 22, 1986.
During NRC Inspection 50-219/86-04, the Region I ll staff reviewed the Oyster Creek MSIV leakage test results and maintenance his-tory to determine the extent of leakage at Oyster Creek and the effectiveness 1
1 of the licensee's maintenance program.
During this inspection, the regional staff determined that MSIV leakage had not been excessive generally less than 100 standard cubic feet per hour (scfh).
Available test data showed that on i
only one occasion, in 1982, did one valve leak in excess of 100 scfh.
Leakage i
data from five outages were reviewed for the period 1977 through the 1983-84 outage and showed that on only two occasions, in 1978 and 1982, did two valves in series require corrective maintenance as determined by the leakage tests.
The maximum leakage through any single penetration would have been 14.13 and 22.9 scfh for these cases.
1 l
Also, on the basis of the review of MSIV maintenance data, the Region I staff determined that the preventive and corrective maintenance being performed has been effective in maintaining the valves' performance.
The licensee s main-tenance program for these valves includes input from both General Electric and i
the valve manufacturers, Atwood and Morrill.
The licensee's routine preventive maintenance program for MSIVs calls for two valves to be rebuilt each refueling outage and the other two valves to be repacked.
The licensee is continuing discussions with General Electric and the valve manufacturers to ensure that 4
repair methods are kept up to date.
r 3-8 j
In its safety evaluation dated May 22, 1986, the staff concluded that the licensee has developed and implemented and is keeping up to date a maintenance program adequate to maintain the MSIVs in an acceptable condition.
Therefore, the staff concludes that the issue in IPSAR Section 4.38 is satisfactorily resolved.
3-9
4 IPSAR TOPIC RESOLUTIONS CONFIRMED BY NRC REGION I 0FFICE During the integrated assessment for Oyster Creek, a number of issues were re-solved by commitments made by the licensee for specific plant modifications or procedural changes.
After the IPSAR for Oyster Creek was issued, the Region I l
office was asked through Task Interface Agreement 83 to verify that plant modi-fications had been implemented and to review changes to plant operating proced-ures made by the licensee.
Table 2.1 includes a list of IPSAR actions for which confirmation by the Region I office was requested.
Region I personnel conducted onsite inspections for each item identified below.
The inspections consisted of examinations of installed equipment as well as a review of supporting procedures and other documentation.
Inspection findings and the results of the review are documented in inspection reports as noted in the following sections.
4.1 Topics II-3.B, Flooding Potential and Protection Requirements; II-3.B.1, Capability of Operating Plants To Cope With Design-Basis Flooding Conditions; II-3.C, Safety-Related Water Supply (Ultimate Heat Sink (UHS)) (NUREG-0822, Section 4.1) 4.1.1 Isolation Condenser Flooding (NUREG-0822, Section 4.1(4))
In Section 4.1(4) of the IPSAR, the staff required the licensee to make procedural revisions to include the fire water storage tank as a edundant source of water supply to the emergency condenser and to include in operating procedures a minimum inventory of water to be maintained in the condensate storage tank.
The following licensee procedures specify actions associated with emergency condenser water supplies:
Procedure 307, "Isolation Condenser System," states, in relation to fill-ing the isolation condenser, "In emergency situations fire protection shall be used if condensate transfer is not available." The procedure also provides instructions for providing makeup water to the isolation condenser from the fire protection system.
Procedure 316, "Condensate System," specifies that 20 feet (250,000 gal-lons) of water should be maintained in the condensate storage system.
Procedure 333, "Plant Fire Protection System," specifies that 310,000 gal-lons of water or more should be maintained in the fire water storage tank.
Procedure 2000-ABN-3200.31, "High Winds," specifies certain actions to be taken at specific sea water levels.
Among these actions are filling the isolation condenser to the high level alarm (7.7 feet) and filling the condensate storage tank to the high level alarm (43 feet).
Inspection Report No. 50-219/86-38 reports on the above.
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These provisions are acceptable; however, full resolution of this issue depends on the resolution of the related issue in IPSAR Section 4.1(1) (Section 2.1.1 of this supplement).
See related discussions in Sections 2.1.1 and 2.5.3 of this supplement.
4.1.2 Hurricane Flooding of Pumps (NUREG-0822, Section 4.1(6))
In Section 4.1(6) of the IPSAR, the staf f indicated that the licensee hid pro-posed to update emergency procedures, to identify the alternate water sources and flow paths if the intake structure became flooded, and to identify the priority of water sources and flow paths to be used to ensure a safe shutdown.
The licensee has provided procedural instructiens in Station Procedure 2000-ABN-3200.31, "High Winds," for the actions to be taken in the event of high water level in the intake structure.
The instructions include actions to be taken in shutting down the circulating water pumps and the service water In addition, Station Procedure 307, "Isolation Condenser System," pro-pumps.
vides instructions for providing makeup water to the isolation condenser using the fire protection system should the preferred condensate transfer system not be available.
Inspection Report No. 50-219/87-04 reports on the above.
This issue is considered to be fully resolved.
4.1.3 Roof Drains (NUREG-0822, Section 4.1(9))
In Section 4.1(9) of the IPSAR the staff stated that the licensee had committed to drill holes in the parapets and install scuppers to preclude the potential for buildup of rainwater on the roof of either the reactor building or the turbine building.
The NRC resident inspector verified that holes have been provided and scuppers have been installed in the reactor and turbine building parapets.
Inspection Report No. 50-219/87-08 reports on the above.
This issue is considered to be fully resolved.
4.2 Topic III-3.C, Inservice Inspection of Water Control Structures (NUREG-0822, Section 4.5) 4.2.1 Intake Structure Trash Racks and Intake Screens (NUREG-0822,
)
Section 4.5.2)
In Section 4.5.2 of the IPSAR, the staff stated that the licensee had agreed to i
formalize as part of shift turnover procedures the shift inspection of the intake structure and to modify the screen wash system to prevent buildup of sea lettuce.
The licensee has in place an intake area tour sheet that has to be completed l
each shift.
Also, licensee internal report Number 500-0C-533 Div II (Budget Activity 402188) describes the modifications performed on the screen wash system during the Cycle 10 refueling outage to prevent buildup of sea lettuce.
4-2
Inspection Report No. 50-219/87-04 reports un the above.
This issue is considered fully resolved.
4.2.2 Inspection Program (NUREG-0822, Section 4.5.4)
In Section 4.5.4 of the IPSAR, the staff stated that the licensee had agreed to provide an inspection program that included review by qualified engineering personnel of water control structures and to establish the inspection and docu-mentation of water control structures following extreme events.
The licensee has in place four procedures that deal with the inspection of water control structures.
Procedure 9410-SUR-4512.09, "0CNGS Non-Radiological Environ-mental Surveillance," provides for the inspection of intake and discharge canal banks monthly or after severe storms.
Procedure 9410-SUR-4570.01, "0yster Creek /
Forked River Hydrographic Surveying," provides for an annual hydrographic survey of the Oyster Creek and Forked River waterways, which serve as discharge and in-take waterways for the plant.
Procedure 9430-SUR-4550.01, "0yster Creek / Forked River Environmental Engineering Survey," provides for an annual environmental engineering surveillance of the Oyster Creek intake and discharge waterways east of Route 9.
Procedure 2000 ABN-3200.31, "High Winds," provides for the inspection of the intake structure following high wind conditions.
These pro-cedures appear to satisfy the water control structure inspection requirenents.
Inspection Report No. 50-219/87-08 reports on the above.
This issue is considered to be fully resolved.
4.3 Topic V-5, Reactor Coolant pressure Boundary (RCPB) Leakage Detection (NUREG-0822, Section 4.16) 4.3.1 Operability Requirements (NUREG-0822, Section 4.16.2)
In Section 4.16.2 of the IPSAR, the staff required the licensee to provide the appropriate action requirements in the Technical Specifications for inoperable leakage detection systems and any necessary procedural changes to provide sur-veillance and testing commensurate with the required sensitivity.
The licensee submitted the necessary Technical Specification change request, which resulted in the issuance of License Amendment 97 on January 6, 1986.
This amendment provided the limiting conditions for operation and added sur-veillance requirements for the reactor coolant leakage detection system.
The licensee normally provides instructions for performing Technical Specifi-cation-required surveillance tests of this type in 600-series procedures.
These procedures satisfy the requirements of the Technical Specifications and Regulatory Guide 1.33, which require implementing procedures for each surveil-lance test listed in the Technical Sptcifications.
During the review of this item, the staff determined that no 600-series procedure had been prepared to perform the surveillance listed in Technical Specification 4.3.H for channel calibration of the primary containment sump flow integrator and the primary containment equipme.+t drain tank flow integrator.
This failure to provide a surveillance test implementing procedure is contrary to the requirements of 4-3
Technical Specification 6.8.1 and Regulatory Guide 1.33, which require imple-menting procedures for each surveillance calibration listed in the Technical Specifications, and is considered to be a violation (219/87-08-01).
Records show calibrations had been performed on these instruments in July 1985 and July 1986.
These calibrations were performed in accordance with a Techni-cal Specification supporting installed instrumentation list (TSSIIL) procedure.
Calibrations performed in accordance with this TSSIIL do not have a detailed implementing procedure, nor do they have the same documentation and review requirements as do 600-series surveillance procedures.
For the 1985 and 1986 tests, TSSIIL calibration data sheets were available.
However, without the benefit of an implementing procedure, it took plant engineers several hours, including talking to the technician who performed the test, to determine how the calibration was conducted.
During the July 1986 calibration, the licensee found that the drywell sump leak rate counter was defective, A maintenance and construction short form was initiated, and the counter was replaced and tested satisfactorily.
Following the identification of the failed counter on July 3, 1986, Deviation Report 86-287 was prepared on the same day.
A plant engineering work request was initiated on July 17, 1986, to review the effect af the drywell sump flow counter error on leak rate calculations.
Plant Engineering Task Assignment (PETA)86-141 was prepared on July 29, 1986, to perform this evaluation.
A responsible technical review of the deviation report was performed on August 1, 1986, and the PETA was completed on October 9, 1986.
This review determined that the Technical Specification limit of < 5.0 gpm unidentUied 'aak rate was not exceeded because of the as-found-counter error.
The actions associated with the failure to prepare the required surveillance test procedure were also reviewed.
This review determined that following the preparation of the Technical Specification change request associated with the leak rate instrumentation, licensing Action Item (LAI) 84179.01 was written on January 7, 1985.
This LAI assigned plant engineering personnel with the responsibility of preparing the necessary administrative controls, surveil-lance, etc.
In response to this LAI, plant engineering personnsi identified the actions that had been taken.
These actions included the assignment of PETA 85-244 to the instrumentation and control group to write a calibration procedure for the drywell equipment drain tank flow integrator and for the drywell sump flow integrator.
This PETA was written on January 23, 1985, and specifkally identified the task scope as writing a 600-series procedure for calibration of the drywell sump flow integrator and the drywell equipment drain tank flow integrator.
Another LAI, 84179.03, was written on January 22, 1986, following the issuance of Technical Specification Amendment 97 to ensure procedural compliance with the amendment.
PETA 85-244, 2 years af ter its preparation, was still open.
The NRC resioent inspector discussed the failure to complete this PETA in a timely manner with the licensee.
The inspector believed that the failure to provide a required 600-series surveillance procedure resulted from an improper prioritization of the PETA.
Licensee representatives stated that at the time the PETA was prioritized, some personnel considered that the performance of the surveillance testing in accordance with the TSS!!L program met the Technical Specification requirements.
They felt that the failure to prepare a required surveillance l
procedure was not due to improper prioritization but was the fault of the 4-4
TSS!!L procedure, which did not clearly establish that the TSS!!L program is not to be used for the performance of surveillance tests listed je the Technical Specifications.
The licensee further agreed that proper prioritization of PETAs is important and that after this incident had been identified, it had reviewed the prioritiza-tion of PETAs.
L 1
l Inspection Report No. 50-219/87-08 reports on the above.
In a letter dated June 1, 1987, responding to Inspection Report No. 50-219/
87-08 and the deficiency identified therein related to this topic, the licensee committed to issue the needed procedure before the next scheduled surveillance.
Region I staff will verify that the licensee has issued the procedure in order to resolve this item.
4.4 Topic V-10.B, Residual Heat Removal System Reliability (NUREG-0822.
Section 4.18)
In Section 4.18 of the IPSAR, the staff indicated that the licensee had agreed to implement generic guidelines for emergency procedures.
The licensee has replaced the previously existing emergency procedures with l
general emergency operating procedures developed in conjunction with the BWR Owners Group and the TMI Action Plan requirements.
Inspection Report No. 50-219/87-04 reports on the above.
This provision is considered to be acceptable; however, the procedural resolu-tion could be affected by the resolution of the issues discussed in IPSAR Sections 4.1(1), 4.1(4), 4.6.4, and 4.30 (Sections 3.1.1, 4.1.1, 2.5.3, and 4.10 of this supplement, respectively).
4.5 Topic V-12.A, Water Purity of BWR Primary Coolant (NUREG-0822, Section 4.20)
In Section 4.20 of the IPSAR, the staff indicated that the licensee was to incorporate into plant procedures the time-related conductivity limit, the chloride concentration limit, and the pH limit foc reactor coolant referenced in "BWR Water Quality Specification" (Specification No. SP-1302-28-001).
The licensee also was to incorporate the conductivity and chloride limits established in Regulatory Guide 1.56 into the facility Technical Specifications.
The NRC inspector verified that the time related conductivity limit, the chloride concentration limit, and the pH limit for reactor coolant specified in Specification No. SP-1302-28-001 have been incorporated into Station Procedure 827.1, "Primary System Analysis; Reactor Water." Also, License Amendment 93 l
incorporated into the Technical Specifications the chloride and conductivity limits establisheo in Regulatory Guide 1.56.
Inspection Report No. 50-219/87-08 reports on the above, s
This issue is considered to be fully resolved.
\\
4-5
4.6 Topic VI-1, Organic Materials and Postaccident Chemisty (NUREG-0822, Section 4.21) 4.9.1 Organic Materials (NUREG-0822, Section 4.21.1)
In Section 4.21.1 of the IPsAR, the staff stated that the licensee was to ascer-tain the chemical composition of the existing drywell coatings.
If these coat-ings contained hydrocarbons, they were to be removed or the licensee was to sub-mit an evaluation to justify the continued use of these coatings.
By letter dated February 10, 1984, the licensee provided a description and re-sults of the drywell inspection conducted during the Cycle 10 refueling outage.
i The licensee concluded that the chemical composition was satisfactory.
The torus interior was coated during the Cycle 10 refueling outage.
Inspection Report No. 50-219/87-08 reports on the above.
This issue is considered to be fully resolved.
4.7 Topic VI-4, Containment Isolation System (NUREG-0822, Section 4.22) 4.7.1 Locked-Closed Valves (NUREG-0822, Section 4.22.1)
In Section 4.22.1 of the IPSAR, the staff identified 31 valves that are either test, vent, drain, er sample line manual isolation valves that connect to piping penetrating the containment.
The licensee was to provide administrative procedures to ensure these valves are locked closed.
Two of the valves on the list, V-14-21 and V+14-39, should have been V-14-20 and V-14-40, respectively.
Also, V-17-51 has been replaced with V-17-51 because of the addition of the postaccident sampling system.
The NRC inspector verified that all valves listed have been included in the containment system integrity valve checkoff list of Procedure 312, "Reactor Containment Integrity and Atmosphere Control," or on the valve checkoff list for Procedure 305, "Shutdown Cooling System Operation." All valves identified in IPSAR Section 4.22.1 are required by these lists to be 1ccked closed.
Tne inspector verified that valve statuses in the containment system integrity valve checkoff list had been last confirmed in November 1986 before the startup following the last refueling.
Inspection Report N;. 50-219/87-08 reports on the above.
I This issue is considered to be fully resolved.
4.8 Topic VI-7.C.1, Appendix K - Electrical Instrumentation and Control Re-Reviews (NUREG-0822, Section 4.25) 10 CFR 50 (GDC 17), as implemer.ted by Regulatory Guide 1.6 and Institute of Elec-trical and Electronics Engineers Std. 308-1974, requires that onsite electrical l
power supplies and their onsite distribution systems have sufficient independence l
to perform this safety function assuming a single failure.
1 4-6 I
4.8.1 AC Automatic Bus Transfers (NUREG-0922, Section 4.25(1))
In Section 4.25(1) of the IPSAR, the staff noted that there were seven automatic transfers of load or load groups between redundant sources in the ac system.
The licensee agreed to perform a coordinated load and circuit breaker analysis to establish any corrective actions necessary to preclude automatic transfer of faults between redundant power sources.
By letter dated September 1, 1983, the licensee submitted an analysis of the coordination of protective devices for fault current interruption.
In some cases, a fault could be transferred to a redundant safety bus.
Therefore, by letter cated July 30, 1984, the licensee proposed to replace the timing devices in these breakers so that the fault interrupt currents were so coordinated that the fault could not be transferred.
In the staff's safety evaluation dated November 16, 1984, the staff concluded l
that the relay coordination (i.e., time response characteristics) provided sufficient independence between redundant electrical divisions to preclude the automatic trasfer of faulted loads between those divisions.
Therefore, the staff found that the licensee's proposal to replace the two relays with over-lapping time response characteristics was acceptable.
The staff also agreed with the emended schedule discussed in the licensee's letter of October 25, 1984, to provide for the installation of the new relays at the first outage of 5 or more days after the necessary parts were received.
Region I staff will verify that the proper relays have been installed.
Verification of the instal-lation of the relays will close out this SEP issue.
4.9 Topic VII-1.A, Isolation of Reactor Protection System From Non-Safety Systems, Including Qualification of Isolation Devices (NUREG-0822, Section 4.27) 4.9.1 Reactor Protection System (RPS) Protective Trip (NUREG-0822, Sec-tion 4.27(2))
In Section 4.27(2) of the IPSAR, the staff indicated that the licensee was to install Class 1E protection at the interface between the reactor protection system power supply and the reactor protection system.
A completed licensing action item documents the installation of six electrical protection assemblies, qualified to Class 1E requirements, between reactor protection system motor generator sets 1-1 and 1-2 and auxiliary transformer and protection system panels 1 and 2.
This modification was completed during Cycle 11R under Budget Activity 402032.
Inspection Report No. 50-219/87-08 reports on the above.
This issue is considered to be fully resolved.
4.10 Topic VII-3, Systems Required for Safe Shutdown (NUREG-0822, Section 4.30)
IPSAR Section 4.30 references Sections 4.1(4), 4.1(6), and 4.18.
The issues in Sections 4.1(6) and 4.18 are resolved, as discussed previously.
Full resolu-tion of these issues depends on the resolution of the,ssue in Section 4.1(1),
4-7
which is indirectly referenced through Section 4.1(4).
See the related dis-cussions in Sections 2.1.1 2.5.3, 3.1.1, 4.1.1, and 4.4 of this supplement.
4.11 Topic VIII-2, Onsite Emergency Power Systems (Diesel Generator)
(NUREG-0822, Section 4.31) 4.11.1 Diesel Generator Annunciators (NUREG-0822, Section 4.31(1))
In Section 4.31(1) of the IPSAR, the staff statt.d that the licensee had agreed to make certain modifications to the diesel generator annunciators.
Licensee documentation shows that the following modifications to the diesel generator annunciators were oerformed under Modication E.T. 312-78 to satisfy the commitments made to the NRC staff:
(1) removing existing nondisabling alarms from ' 9 resent diesel generator trouble alarm (2) providing a new annunciator for the manual mode switch not in automatic (3) redesigning the working of the annunciator windows to reflect the condi-tions more clearly (4) providing a low battery voltage sensor with an alarm function indicating diesel generator dc failure Inspection Report No. 50-219/87-08 reports on the above.
This issue is considered to be fully resolved.
4.11.2 Diesel Generator Trip Bypass (NUREG-0822, Section 4.31(2))
By letter dated November 16, 1982, the licensee committed to modify diesel gen-erator trips (bypass of lead voltage-ampere reactive and reverse power trips) by the end of the Cycle 11 outage.
The staff concluded that the modifications would be acceptable when they were completed and verified.
Verification by Region I staff of the implementation of these modifications will close out this SEP issue.
4.12 Topic VIII-3.B, DC Power System Bus Voltage Monitoring and Annunciation (NUREG-0822, Section 4.32)
In Section 4.32 of the IPSAR, the staff stated that the licensee had committed to install alarms for B and C battery breaker open, C battery charger open, and C battery ground.
The NRC inspector verified that the functions identified are alarmed in the control room.
The alarm annunciators do not always have the same designation as the function; however, a review of the alarm response procedures verified that the functions are included in the alarm.
Inspection Report No. 50-219/87-08 reports on the above.
This issue is considered to be fully resolved.
4-8
4.13 Topic IX-5, Ventilation Systems (NUREG-0822, Section 4.34) 4.13.1 Restoration of Ventilation (NUREG-0822, Section 4.34(1))
In Section 4.34(1) of the IPSAR, the staff stated that the licensee was to re-view and modify as required the loss-of offsite power procedure to ansure that operation of ventilation systems was adequately addressed and would not overload the diesel generators.
The result of this evaluation was to be submitted by March 1983.
The licensee does not have a loss-of-offsite power procedure, but provides the necessary instructions for restoring emergency buses to service, if lost, in Station Procedure 341, "Emergency Diesel Generator Operation." This procedure provides guidance on diesel generator load limitations and load sequencing.
Also, Region I Inspection Report 50-219/86-37 documents a review conducted to ascertain that the present configuration of the plant's offsite and onsite elec-tric power systems is capable of sustaining and/or switching loads as required to support the safe operation of the plant.
On the basis of discussions with the NRC staff, the licensee has agreed to submit the required evaluation.
Excect as noted above, no violations were identified.
4-9
5 REFERENCES General Electric Company, Topical Report NED0-21617 "Analog Transmitter / Trip Unit System for Engineered Safeguard Sensor Trip Input," December 1978.
Institute of Electrical and Electronics Engineers (IEEE) Standards:
' 9-1971, "Criteria for "rotection System for Nuclear Power Generating Stations."
308-1974, "Criteria for Class 1E Power Systems for Nuclear Power Generating Stations."
379-1977, "Guide for the Application of the Single Failure Criterion to Nuclear Power Generating Station Protection System."
5-1
APPENDIX NRC STAFF CONTRIBUTORS AND CONSULTANTS This supplement is a product of the NRC staff and its consultants.
The prin-cipal contributors to this report were:
NRC Staff W. Baunack C. Ferrell P. Chen C. Jamerson T. Cheng E. McKenna J. Donohew, Jr.
T. Michaels A. Oromeric F. Orr R. Fell M. Srinivasan Consultants Franklin Research Center MPR Associates 1
Appendix
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Docket No. 50-219 "ffiE"6'.f"$MEar Regulatory Commission (NRC) has prepared Supplement 1 to the final Integrated Plant Safety Assessment Report (IPSAR) (NUREG-0822), under the scope of the Systematic Evaluation Program (SEP), for the Oyster Creek Nuclear Generating Station, located in Ocean County, New Jersey and operated by GPU Nuclear Corporation and Jersey Centeral Power and Light Company (colicensees).
The SEP was initiated by the NRC to review the design of older operae,g nuclear power plants to reconfirm and document their safety.
This report doca.aents the review completed under SEP for those issues that required refined engineering evaluations or the continuation of ongoing evaluations subsequent to issuing the Final IPSAR for the Oyster Creek plant.
The review has provided for (1) an assessment of the significance of differences between current technical positions on selected safety issues and those that existed when the Oyster Creek plant was licensed, (2) a basis for deciding how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety. The final IPSAR and its supplement will form part of the bases for considering the conversion of the existing provisional operating license to a full-term operating license.
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