ML20009B589
| ML20009B589 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 07/09/1981 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Finfrock I JERSEY CENTRAL POWER & LIGHT CO. |
| References | |
| TASK-03-01, TASK-05-11.B, TASK-07-03, TASK-3-1, TASK-5-11.B, TASK-RR LSO5-81-07-030, LSO5-81-7-30, NUDOCS 8107160353 | |
| Download: ML20009B589 (2) | |
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July 9,1981 Docket No. 50-219 N
LS05-81-07-030
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.2 Mr. I. R. Finfrock, Jr.
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Vice President
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Jersey Central Power & Light Company Post Office Box 388 Forked River, New Jersey 08731 b
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Dear Mr. Finfrock:
SUBJECT:
SEP TOPICS III-1, V-11.B. & VII-3 FOR OYSTER CREEK is a draft technical evaluation report that has been prepared by our support contractor. provides 1) a list of the signifi-cant electrical instrumentation and control equipment that is required to,-
protect public health and safety in the Oyster Creek Design, 2) the tech-nical basis for Enclosure 2; and 3) the technical basis for Et ::losure 3. is the staff's draff safety evaluation report (SER) for SEP Topic V-11.B.
This SER is'.thei: basis for concluding that the RHR system suction and discharge interlocks satisfy current licensing criteria. is the staff's draft safety evaluation report for SEP Topic VII-3. This evaluation proposes modifications to the indication systems used for safe shutdown.
Your coments on Enclosures 1, 2, and 3 are requested within 30 days.
L The need to actually implement the changes will be detennined during the integrated plant safety assessment. These topic assessments may be revised in the future if your facility design is changed or if NRC criteria relat-ing to this topic are modified before the integrated assessment is completed.
Sincerely, Dennis M. Crutchfield, Chief Operating Reactors Branch No. 5 Division of Licensing ftab'h I
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July 9,1981 Docket NoLS05-81-07 50-219 030 Mr. I, R. Finfrock, Jr.
Vice President Jersey Central Power & Light Company Post Office Box 388 Forked River, New Jersey 08731
Dear Mr. Finfrock:
SUBJECT:
SEP TOPICS III-1, V-ll.B. & VII-3 FOR OYSTER CREEK is.a draft technical evaluation report that has been prepared Dy our support contractor. provides 1) a list of the signifi-cant electrical instrumentation and control equipment that is required to protect public health and safety in the Oyster Creek Design, 2) the tech-nical basis for Enclosure 2; and 3) the technical basis for Enclosure 3.
. is the staff's draft safety evaluation report (SER) for SEP Topic V-11.B.
This SER is the basis for concluding that the RHR system suction and discharge interlocks satisfy current licensing criteria. is the staff's draft safety evaluation report for SED Topic VII-3. This evaluation proposes modifications to the indication systems used for safe shutdown.
3 Your comments on Enclosures 1, 2, and 3 are requested within 30 days.
The need to actually implement the changes will be determined during the integrated plant safety assessment. These topic assessments may be revised l
in the fature if your facility design is changed or if NRC criteria relat-ing to this topic are modified before the integrated assessment is completed.
Sincerely, bUGh$v A..a/ e.
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y Dennis M. Crutchfield, Chief Operating Reactors Branch No. 5 Division of Licensing j
Enclosures:
-As stated cc w/ enclosures:
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Mr. I. R. Finfrock, J r.
ec G. F. Troubridge, Esquire Gene Fisher Shaw, Pittman, Potts and Trowbridge Bureau Chief 1800 M Street, N. W.
Bureau of Radiation Protection Washington, D. C.
20036 380 Scotts Road Trenton, New Jersey 08628 J. B. Lieberaun, Esquire Berlack, Israels & Lieberman Comnissioner 25 Broadway New Jersey Departnent of Energy New York, New York 10004 101 Commerce Street Newark, New Jersey 07102 Natural Resources Defense Council 91715th Street, N. W.
Licensing Supervisor Washington, D. C.
20006 Oyster Creek Nuclear Generating Station J. Knubel P. O. Box 388 BWR Licensing Manager Forked River, New Jersey 08731 Jersey Central Power & Light Coapany Madison Avenue at Punch Bowl Road Resident Inspector Morristown, New Jersey 07960 c/o U. S. NRC P. O. Box 445 Joseph W. Ferraro, Jr., Esquire Forked River, New Jersey 08731 Deputy Attorney General State of New Jersey Departnent of Law and Public Safety 1100 Raymond Boulevard Newark, New Jersey 07012 Oceari County Library Brick Township branch 401 Chambers Bridge Road Brick Town. New Jersey 08723 Mayor Lacey Township P. O. Box 475 Forked River, New Jersey 08731 Commissioner Department of Public Utilities State of New Jersey
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101 Commerce Street Newark, New Jersey 07102 U. S. Environmental Protection Agency Region II Office ATTN: EIS C0ORDINATOR 26 Federal Plaza New York, New York 10007
Ob92J SEP TECHNICAL EVALUATION TOPIC VII-3 ELECTRICAL. INSTRUMENTATION AND CONTROL FEATURES OF SYSTEMS REQUIRED FOR SAFE SHUTDOWN OYSTER CREEK NUCLEAR STATION Jersey Central Power and Light Company l
June 1981 D. A. Weber l
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CONTENTS
1.0 INTRODUCTION
I 2.0 REVIEW CRITERIA 2
3.0 RELATED SAFETY TOPICS AND INTERFACES.............
2 4.0 REVIEW GUIDELINES 4
5.0 DISCUSSION AND EVALUATION 5
5.1 Instrumentation.....................
5 5.1.1 Evaluation....................
6 5.2 Safe Shutdown Systems..................
7 5.2.1 Onsite Power Unavailable.............
9 5.2.1.1 Evaluation 10 5.2.2 Offsite Power Unavailable 10 5.2.2.1 Evaluation 10 5.3 Shutdown and Cooldown Capability Outside the Control Room 11 5.3.1 Evaluation....................
11 5.4 RHR Reliability and Interlocks 11 5.4.1 Evaluation....................
12 6.0
SUMMARY
12 7.0 SAFE SHUTDOWN EI&C FEATURES FOR CONSIDERATION BY SEP TOPIC 111-1 13
8.0 REFERENCES
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SEP TECHNICAL EVALUATION TOPIC VII-3 ELECTRICAL, INSTRUMENTATION AND CONTROL FEATURES OF SYSTEMS REQUIRED FOR SAFE SHUTDOWN OYSTER CREEK NUCLEAR STATION
1.0 INTRODUCTION
This report is part of the Systematic Evaluation Program (SEP) review of Topic VII-3, " Systems Required for Safe Shutdown." The objective of-this review is to determine whether the electrical, instrumentation, and control (EI&C) features of the systems required for safe shutdown, including support systems, meet current licensing requirements.
The systems required for safe shutdown have been identified by the NRC SEP. The systems were reviewed to ensure the following safety objectives are met:
(1) Assure the design adequacy of the safe shutdown system to automatically initiate operation of appropriate systems, including reactivity control systems, such that fuel design limits are not exceeded as a result of operational occurrences and postulated accidents, and to automatically initiate systems required to bring the' plant to a safe shutdown (2) Assure that required systems, equipment, and con-trol to maintain the unit in a safe condition dur-ing hot shutdown are appropriately located outside the control room, and have the capability for sul-sequent cold shutdown of the reactor using suitable procedures (3) Assure only safety grade equipment is required to bring primary coolant systems from a high pressure to low pressure cooling condition.
The scope of this review specifically includes an evaluation of the electrical, instrumentation, and control features necessary for operation of the identified safe shutdown systems.
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e The review evaluates the systems for operability with and without offsite power and the ability to operate with any single f ailure. The EI&C review of safe shutdown systems only includes those features not covered under other SEP Topics. Specific items which will be ccvered under other SEP reports are identified in Section 4.0, Review Guide-lines.
2.0 REVIEW CRITERIA Current licensing criteria for safe shutdown is contained in the following:
(1)
IEEE Standard 279-1971, " Criteria for Protection Systems for Nuclear Power Generating Stations" (2) GDC-5, " Sharing of Structures, Systems, and Com-ponents" (3) GDC-13, " Instrumentation and Control" (4) GDC-17, " Electric Power Systems" (5) GDC-19, " Control Room" (6) GDC-26, " Reactivity Control System Redundancy and Capability" (7) GDC-34, " Residual Heat Removal" (8) GDC-35, " Emergency Core Cooling" (9) GDC-44, " Cooling Water."
3.0 RELATED SAFETY TOPICS AND INTERFACES The following list of SEP topics are related to.the safe shutdown topic with respect to El&C features, but are not being specifically reviewed under this topic:
(1) SEP III-10.A, " Thermal Overload Protection for
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Motors of Motor-Operated Valves" 2
(2) SEP IV-2, " Reactivity Control Systems Including Functional Design and Protection Against Single Failures" (3) SEP VI-7A3, "ECCS. Actuation System" (4) SEP VI-Cl, Appendix K, "EI&C Re-reviews" (5) SEP VI-10A, " Testing of RTS and ESF Including Response Time Testing" (6) SEP VI-108, " Shared ESF, Onsite Emergency Power, and Service Systems for Multiple Unit Facilities" (7) SEP VII-1, " Reactor Trip System" (8) SEP VII-2, "ESF Control Logic and Design" (9) SEP VIII-2, "Onsite Emergency Power Systems--Diesel Generators" (10) SEP VIII-3, " Emergency DC Power Systems" (11) SEP IX-3, " Station Service and Cooling Water Systems" (12) SEP IX-6, " Fire Protection."
Where safe shutdown system EI&C response is affected by the above-mentioned topics, that particular SEP review has been consulted for determination of overall safe shutdown system performance. Where the SEP topic review is not available, the effect on safe shutdown system performance has been identified as being based on an assumed operating condition of the effecting system. The safe shutdown review will be considered pre!iminary until resolution of the effecting topic is com-pleted and found to be in accordance with assumptions made in this review.
The completion of this review impacts upon the'following SEP top-ics, since capabilities relating to safe shutdown is required in the topic:
(1) SEP VII'.-1A, " Potential Equipment Failures Associ-ated with a Degraded Grid Voltage" 3
(2) SEP VIII-2, "Onsite Emergency Power Systems--Diesel Generators."
4.0 REVIEW GUIDELINES The capability to attain a safe shutoown has been reviewed by evaluating the systems used for normal shutdown (onsite power not avail-able) and emergency shutdown (cffsite power not available). CRP 7.4 was applied to each system to ensure the following guidelines were met:
(1) They have the required redundancy (SRP 7)
(2) They meet the single failure criterion (RG 1.53, ICSB BTP 18)
(3) They have the required capacity and reliability to perform intended safety functions on demand (SRP 7).
Additionally, SRP 5.4 requirenents contained in BTP RSB 5-1 were reviewed to determine if the systems required for shutdown cooling (residual heat removal) met the following criteria:
(1) The systems are capable of being operated from the control room with only offsite or only onsite power available (2) The systems are capable of bringing the reactor to cold shutdown with only offsite or only onsite power available within a reasonable period, assuming the most limiting single failure The electrical equipment environmental qualification and physical separation are being reviewed under other topics, as is the seismic equipment qualification, and are not reviewed in this Yeport. Sec-tion 7.0 consists of a list of safety related EI&C eg'uipment necessary for safe shutdown to be used in resolving SEP Topic III-1, "Classifica-tion of Structures, Components, and Systems."
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5.0 DISCUSSION AND EVALUATION 5.1 Instrumentation The NRC SEP Review of Safe Shutdown Systems identified the instru-mentation available in the control room necessary to bring the reactor from the hot shutdown to cold shutdown condition. This review also evaluated the nucis r instrumentation, since this parameter must be (mnitnred to ensure the reactor achieves and maintains shutdown con-
~ditiont, Various system parameters, such as pump running or valve position Indications, are not included in the list of safe shutdown instruments of 'iehle 4.2 of the SEP Review of Safe Shutdown Systems.
This is due to the fdct that indication is provided by the control /
operate circuitry. Avai'lehility of control / operate circuitry to run the system also means availab Mity of the required indication.
Similarly, if the control / operate (tecuitry is unavailable such that system operation is not possible, then system indication is not inandatory.
The nuclear instrume 1tation consists of two 1adependent +24 V DC buses (Panels A & B) and'two incapendent 125 V AC bu:6c (Protection System Panels hc. I and 2) providing redundant indication of each range of power level. The 1.24 V DC buses are powered by battery / charger arrangements. The 120 V AC buses are supplied primary power from independent MG sets (powered from independent buses) and backup power from Vital MCC Buses lA2 or 182 with a manual transfer switch. There l
are no single failures which would result in the loss of both 1,24 V DC l
systems or both 120 V AC systems.
l The reactor parameter indicators (level, pressure, and temperature) are powered from the Protection System Panel No. 1, the Continuous Instrument Panel No. 3, or the Instrument Panel 4C. Since the reactor operates in a saturated steam / water environment, knowledge of either temperature or pressure implies knowledge of the other (use of steam tables or suitable chart required). No electrical single failure.vould result in failure of pressure, temperature, and level indicating 5
in:truments. Therefore, operators would have th9 necessary informat'i~ n o
available for determining reactor temperature, ievel, and pressure.
Mechanical / hydraulic (Yarway) indicators of reactor level are available in the reactor building such that, with appropriate communications, control room personnel would have reactor level indication.
The instruments in the control room providing indication of reactor pressure, temperature, and level, as well as those providing indication nf various systems, flow, temperature, pressure, valve position, etc.,
are independent of the instruments used to initiate Reactor' Protection System (RPS) and Engineered Safety Features (ESF) actions. Failure of any of the instruments providing indication in the control room has no effect on the operation of the RPS and ESF actuation systems.,
The indications for power to the various AC and DC buses are sup-plied by lights, meters, or alarms powered from the bus being monitored.
Loss of power to the bus would be indicated in the control room, and no single failures of indications would effect the ability to monitor any other bus.
Indication af Service Water, Containment Ccoling, Core Spray, l'alation Condenter, and Shutdown Cooling sy3 tem parameters such as tiow, temperature, level, and pressure available in the contrei room, 4 rawered by the AC instrument bus. While loss of the AC instrument L.s would cause a loss of this indication, each of these systems has dire " reading in dcators available at its local control station.
Stat.- of fisw for some systems such as Core Spray can be inferred from the pump running / valve open indicators (not powered by.the AC instrument bus) and by reactor parameters of level, pressure, and temperature.
S.1.1 Evaluation.
The instrumentation necessary for reaching and maintainino cold shutdown at Oyster Creek does not meet current licens-ing criteria since potential single failures could render vital indica-tions neccssary for maintaining plant control inoperable. Suitable direct reading local indications are available and could be used if 6
operatorswerestationedatthelocalindicatorsandhadadequatecoN-munications with the control room. Such action would have to be justi-fied by the licensee under the topic of limited operator action outside the control room.
5.2 Safe Shutdown Systems The SEP review of Safe Shutdown Systems identified the systems i
required for short-term cooling (immediately after reactor shutdown) and long-term cooling (when the re
.or is cooled to the SCS design
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temperature limit of 350 F) with only offsite and only onsite power available.
Normal short-term cooling is provided by dump ~ing steam from the reactor to the main condenser via the turbine bypass valves. Circulat-ing Water System (CWS) removes heat by condensing the steam. The feed-water system then returns the water the the reactor. This cooling method is only available when offsite power is available. Failure of
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the feedwater control system, turbine hydraulic control system, AC essential aus, or loss of CWS flow to the condenser can render this method of cocling inoperative. The systems in this method are not class 1E but are being considered as an available means to remove decay Oce;.
' merger ^y or alternate short-term cooling involves operation of the Iscistion (emergency) Condenser System (ICS), Automatic Depressuri-zatior System (ADS), or the main steam safety valves.
The isolation condenser consists vf two independent condensers, each with its own steam inlet line from the reactor and return line to the recirculating loops.
The inlet line centains two normally open Motor Operated Valves (MOVs) in series which Eaintain the isolation condenser and piping at reactor pressure. The return lines also con-tain two MOVs in series; the first is normally closed and the second is normally open. The normally open. valves in the return lines are the only ones that are inside containment. No single failures have been 7
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identified which would completely disable this system, However, a false high flow signal (which would shut all MOVs) coupled with a loss of Motor Control Center (MCC) LAB 2 would prevent the use of this system as this MCC supplies both inaccessible MOVs that are inside containment.
Even under these conditions s'hort term or emergency cooling can still be achieved with the Automatic Depressurization System (ADS).
The ADS consist of five DC operated electromatic relief valves (EMRVs) which are located on the main steam lines and discharge into the tower. These valves are set to operate automatically at pressure below the sixteen main steam safety valves or may be operated manually, if plant condition required the need to immediately decrease pressure and cool the system. The EMRVs are also automatically activated by signals indicating low-low reactor water level, high drywell pressure, and core spray booster pump discharge pressure. Two of the EMRV's are operated from one DC source and the other three from a secor.d independ-ent DC source. No single failures have been identified which would completely disable the entire system.
The main steam safety valves are mechanical valves with no electri.-
cal input and no manual operating capability. They operate to prevent overpressurization of the reactor by dumping steam directly into the drywell.
The Shutdown Cooling System (SCS) can be used for long-term cool-l ing (below 350 F).
It consists of a single suction line, three par-l allel pump and heat exchanger loops, and a common discharge line. The SCS heat exchangers are cooled by the Reactor Building Closed Cooling Water (RBCCW) system flow and the RBCCW heat exchangers are in turn l
cooled by Service Water (SW) flow. There are multipl'e single failures which can render the SCS inoperable including loss of Motor Control Center (MCC) DC-1 bus which provides power to the normally closed SCS.
l pump suction and discharge valves and loss of MCC 1AB2 bus which powers the normally closed system suction and discharge valves.
In addition, the loss of the RBCCW system or the SW system can render the SCS system l
inoperable. Since t'ne SCS and its auxiliaries were not designed and i
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constructed with the quality of the plant safety system, the ADS, Core Spray System, and Containment Cooling System are relied upon for long-term cooling. The Core Spray (CS) System is a low pressure system and would require the use of the ADS to reduce the reactor pressure to about 285 psig. This system, in conjunction with the EMRV's of the ADS, could function as a closed loop by filling the vessel with Core Spray, and overflowing hot water back to the torus through the relief valves. The systerr consists of two independent loops, each with paralled core spray and core spray booster pumps that pump water from the torus to the reactor. Single failures exists which can cause the loss of one of the two loops of the CS system, but only one loop is needed to maintain reactor level and provide cooling water.
As the Core Spray system does not have any heat exchangers to cool the water in the torus, the Containment Spray System and the Emergency Service Water System could be used to perform this function. The Emer-gency Service Water System pumps supply cooling water to the Containment Spray heat exchangers. The Containment Spray pumps take suction from the torus and discharge the water through the heat exchangers to the containment header inside the drywell. The torus water being sprayed inside the drywell will eventually overflow from the drywell back into the torus. These system each contains two independent loops with par-allel emergency service water pumps, containment spray pumps and heat exchangers. Single failures exist which could cause the loss of one of the two trains but the flow through one train is sufficent to provide the necessary heat remotal capabilily.
5.2.1 Onsite Power Unavailable. During power' operation, Oyster Creek normally supplies all of the plant loads from the main generator via the two secondary windings of the Auxilary Transformer. These second:sey windings supply two independent power trains A and B.
Loss of the main generator power during operation will result in a reactor scram and turbine trip. The buses normally powered by the generator will transfer to the start-up transformers SA and SB which are powered from independent 230 KV switchgear sources. Only single failures 9
involving buses, switchgear, etc. downstream of the transformer feed lines to the distribution system are considered.
Single failures of El&C features, such as loss of the feedwater control system, exist which could disable the normal short-term cool-down methods. However, no El&C single failure rendering the the normal short-term cooldown methods inoperable can also cause failure of the isolation condenser and the ADS. There are multiple single failures previously mentioned which can render the SDCS inoperable. However, since only one core spray loop is needed to maintain core level and provide cooling water, there are no single failures which can disable both the SDCS and Core Spray systems. Therefore, the short-term and long-term cooling capability meet the current licensing criteria with only offsite power available. Excluding the non-class lE methods, the required short-term and long-term cooling is still provided.
5.2.1.1 Evaluation. The systems required for short-term and long-term cooling at Oyster Creek are capable of providing the required cooling assuming no onsite power is available and a single failure.
5.2.2 Offsite Power Unavailable. During normal operation, a loss of offsite power will result in a reactor scram, turbine trip, and momentary loss of power to the AC distribution system. Subsequently, diesel generater 1 and 2 (DGl and DG2) will be automatically started to supply power at buses 1C and ID, respectively (Train A and B).
Assuming both diesel generators are available, there are no single f ailures which would disable the ICS, ADS, and core spray system. As before, there are no single failures which would disable both long-term cooling systems (SDCS and core spray) if AC power from one or both diesel generators is available.
5.2.2.1 Evaluation. The systems required 'or reaching and maintaining cold shutdown conditions at Oyster Creek are capable of providing the required cooling assuming only onsite power is available and a single EI&C failure.
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5.3 Shutdown and Cooldown Capability Outside the Control Room The capability to maintain the plant in hot shutdown from outside the control room exists at Oyster Creek. Reactor parameters such as level, pressure, and temperature can be monitored at local stations outside the control room. Reactor temperature (therefore, pressure) can be determined at the ICS system local indication or by local indications such as system discharge pressure for core spray or SDCS.
Local control stations exist for the pumps and valves of the systems required for safe shutdown described in Section 5.2.
Additionally, many of the valves are also capable of being manually operated (such as the ICS return isolation valve and the SDCS isolation valves). However, no procedures for taking.he plant to cold shutdown from outside the control room exist.
5.3.1 Evaluation. Adequate capability exists to maintain the red: tor at hot shutdown from outside the control room. No procedures exist for taking the reactor to cold shutdown from outside the control room.
5.4 RHR System Reliability and Interlocks The SDCS at Oyster Creek is designed to withstand RCS design pres-sure. Therefore, the isolation valve interlocks required by BTP RSB 5-1 are not applicable. The isolation valves have interlocks to prevent opening and to automatically close when RCS temperature in any of the 0
five coolant loops exceeds the 350 F design temperature of the SDCS.
A bypass line provides a flow path from eac:. SDCS pump discharge to its suction to provide the necessary flow to prevent pump overheating due to a discharge isolation valve being closed. A low suction pressure trip will stop the pump, if a suction valve has closed during operation, to prevent pump damage due to cavitation.
There are no requirements in the Oyster Creek Technical Specifica-tions for testing the SDCS interlocks and isolation circuitry during i
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SDCS operation. The electrical circuitry is not designed to permit'-
testing while the system is operating without c momentary interruption in system operation. Although licensed prior to the issuance of RG 1.68, (concerning preope, rational and startup testing) Oyster Creek conaucted such tests and has demonstrated SDCS operability on several occasions as noted by the SEP Review of Safe Shutdown Systems, Sec-tion 4.5.
5.4.1 Evaluation. The SDCS meets the current licensing criteria of BTP RSB 5-1 in accordance with SEP Topics V-10.B and V-ll.B.
6.0
SUMMARY
The systems required to tak6 the reactor from hot shutdown to cold shutdown, assuming only offsite p0wer is available or only onsite power is available and a single failure, are capable of automatic initiation to bring the plani to a safe shutdown and are in compliance with current licensing guidelines and the si.fety objectives of SEP Topic VII-3.
Single failures of El&C equipment cannot render all short-term cooling system inoperable.
The reactor inst:umentation available to control room operators to reach and maintain the reactor iq cold shutdown conditions meets cur-rent licensing criteria since reactor parameter indications (level, temperature, and pressure), are redundant and/or supplied from different sources. Although indirect means exist to measure important parameters such as C.S. flow, Service Water flow, etc., loss of the 120 V AC instrument bus renders riormal flow, temperature, pressure, and level indicators in the control room for Containment Spray,. Service Water, SDCS, RBCCW, Core Spray, and ICS inoperable.
The capability to maintain the reactor in hot shutdown from outside' the control room exists and is in compliance with the safety objectives of SEP Topic VII-3. No procedures exist to take the plant from hot to cold shutdown from outside the control room to satisfy the safety objec-tives of SEP Topic VII-3.
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7.0 SAFE SHUTDOWN EI&C FEATURES FOR CONSIDERATION BY SEP TOPIC III-l ELECTRICAL DISTRIBUTION (including support structure, but not individual loads) 1.
ALL AC SAFETY RELATED SWITCH GEAR, LOAD CONTROL, AND MOTOR CONTROL CENTER BUSES--including all feeders, incoming or outgoing, control circuits, indicating circuits, bus work and support structures 2.
ALL DC BUSES--including 125 V, 250 V, 24 V batter-ies, chargers, breakers, bus work, and support structures 3.
DIESEL GENERATORS 1 and 2--including control and indicating circuitry, and control and indication of vital DG auxiliaries such as lube oil, fuel, and cooling INSTRUMENTATION (including support structures) 1.
REAciOR LEVEL 2.
REACTOR PRESSURE 3.
REACTOR TEMPERATURE 4.
NEUTRON MONITORING (including in-core monitoring) 6.
AREA AND SYSTEM RADIATION MONITORING SYSTEMS (includes pumps, valves, control, indication, and support structures) 1.
SHUTDOWN COOLING SYSTEM 2.
REACTOR BUILDING CLOSED COOLING WATER 3.
EMERGENCY SERVICE WATER SYSTEM 4.
CONTAINMENT COOLING 5.
CORE SPRAY 13
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ADS 7.
ISOLATION CONDENSER 8.
TORUS (suppression. pool) 5.
CONTROL R0D DRIVE SYSTEM (scram function only)
8.0 REFERENCES
1.
Final Safety Analysis Report, Dresden.Nyc]ggp powgp $tation, Units 2 and 3.
2.
Code of Federal Regulations, 10 CFR 50, Appendix A, " General Design Criteria for Nucic,7r Power Plants."
3.
IEEE Standard 279-1971, " Criteria for Protection Systems for Nuclear Power Generating Stations."
4.
NUREG 75/087, Nuclear Re7Jlatory Commission Standard Review Plan 7.4, " Systems Required for Safe Shutdown" and 5.4, " Residual Heat Removal."
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SYSTEMATIC EVALUATION PROGRAM TOPIC V-ll.B 0YSTER CREEK TOPIC V-ll.B RHR INTERLOCK REQUIREMENTS I.
INTRODUCTION The RHR System is normally located outside of primary containment.
It is an intermediate pressure system (usually 600 psia) and has motor operated valve (MOV) isolation valves connecting it to the RCS.
If the RHR is at pressure, a LOCA could result with a loss of all capability of core re-flooding since the coolaht inventory could be lost outside of antainment.
73 prevent inadvertent opening of the M0'l's while the RCS is at pressure, an "0 PEN FERMISSIVE" interlock should be provided.
II. REVIEW CRITERIA review criteria are presented in Section 2 of EG&G Report 0392J, Tha
' Jtr' cal, Instrumentation and Control Features of Systems Required for a.
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III. RELATED SAFETY TOPICS AND INTERFACES There are no review areas outside the scope of this topic and there are no other safety topics that are dependent on the present topic information for coupletion.
IV.
REVIEW GUIDELINES The review guidelines are presented in Section 4 of EG&G Report 0392J.
V.
EVALUA110N As noted in Section 5.4 of EG&G Report 0392J the-shutdown cooling system (RHR) at Oyster Creek is designed for reactor coolant system press.ure.
VI.
CONCLUSION RHR pressure interlocks are not required.
SYSTEMATIC EVALUATION PROGRAM TOPIC VII-3 0YSTER CREEK TOPIC VII-3 SYSTEMS REQUIRED FOR SAFE SHUTDOWN I.
INTRODUCTION The systems aspects of the review of Systems Required for Safe Shutdown was conducted as part of Topic V-10.B (RHR Reliability). This safety evaluation is limited to the electrical instrumentation and control sys-tems identified as being required for safe shutdown.
Plant systems that are needed to achieve and maintain a safe shutdown condition of the plant, including the capability for prompt hot shutdown of the reactor from out-side the control room were reviewed.
Included also, was a review of the design capability and method of bringing the plant from a high pressure condition to low pressure cooling assuming the use of only safety grade equipment. The objectives of the review were to assure:
(1) The design adequacy of the safe shutdown system to (a) initiate automatically the operation of appropriate systems, including the reactivity control systems, such that specified acceptable fuel design limits are not exceeded as a result of anticipated operat-(
ional occurrences or pcstulated accidents and (t) initiate the operation of systems and components required to bring the plant to a safe shutdown.
(2) That the required systems and equipment, including necessary in-strumentation and controls to maintain the unit in a safe condition during hot shutdown, are located at appropriate places outside the control room and have'a potential capability for subsequent cold shutdown of the reactor through the suitable procedures.
l (3) That only safety grade equipment is required to bring the reactor coolant system from a high pressure condition to a low pressure
~
cooling condition.
II.
Review Criteria The review criteria are presented in Section 2 of EG&G Re) ort 0392J,~
" Electrical, Instrumentation, and Control Features of Systems Required 5
for Safe Shutdown."
l III.
Related Safety. Topics and Interfaces Review areas outside the scope of this topic and safety topics that are l
dependent on the present topic information for completion are identified in Section 3 of EG&G Report 0392J.
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l
IV. Review Guidelines The review guidelines are presented in Section 4 of EG&G Report 0392J.
V.
Evaluation As noted in EG&G Report 0392J, the systems required to take Oyster Creek from hot shutdown to cold shutdown, assuming only offsite power is avail-able or only onsite power is available and a single failure, are capable of initiation to bring the plant to safe shutdown and are in compliance with current licensing criteria and the safety objectives of SEP Topic VII-3, except that long-term cooling (RHR) is susceptible to single fail-ures that render vital indications necessary for maintaining plant control inoperable in the control room.
VI.
Conclusions The instrumentation systems that provide vital indications for maintaining control of reactor shutdown should be modified to satisfy the single fail-ure criterion.
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