ML13329A100
| ML13329A100 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 10/31/1979 |
| From: | Southern California Edison Co |
| To: | |
| Shared Package | |
| ML13329A101 | List: |
| References | |
| RTR-NUREG-0578, RTR-NUREG-578 TAC-44652, NUDOCS 8006110013 | |
| Download: ML13329A100 (66) | |
Text
RESPOWES.TO.NRC REQUIREMENTS ESTABLISHED TO DATE
-FOLLOWING THE THREE MILE ISLAND ACCIDENT SAN ONOFRE NUCLEAR GENERATING STATION UNIT 1 OCTOBER, 1979 8006110
TABL OFCONTENTS
- 1. Introduction and Summary
- 2. Response to NUREG-0578 Recommendations NUREG-05 7 8 Section 2.1 Design and Analysis SupplyoRequipemaned rthe 2
2.1.1 mergenlcy Power S p oequirered fRe2 Pressurizer Heaters,poeOertdRie valves and Block Valves and Pressurizer LevelIndicators in PWR's 2.1.2 Performance Testing for BWR and PWR Relief and Safety Valves 2.1.3*a Direct andication of Poweroperated Relief.
Valve and Safety ValvePotiffrPR 1
and BWR's 2.1.3.b Instrumentation for Detection of Inadequate 8
Core Cooling in pWR's and BWR's 1
2 4
Containment Isolation Provisions for 2.1.4 PCRnsandmenR's PWR s a d B R'sforExternal Recombiners 13 2.1.5.a Dedicated Penetrations Systems or post-Accident Purge Sysem 2.1.5.b Inerting BWR Containments 15 2.1.5.c Capability tInstalehydroen Recombiner at Each Light Water Nuclea Power Pl nt 16 of Systems outside Containment 2.1.6.a nLkelrity ontain Radioactive Materials (Engineered Safety Systems and Auxi Systems) for PWR's and BWR1 iew Of Plant Shielding and 2.1.6-b Design Reviwolyiaino Equipment Environmental Q ualification ofEqimt for Spaces/Systems Which May Be Used in post-Accident Operations 2.1.7.a Automatic Initiation of the Auxiliary Feedwater System for PWR'S Auxiliary FeedWater low Indication to 2.1.7*
Steam Generators for PWR's EGU o
0000 FILE COP
TABLE OF CONTENTS (Cont'd)
Page No.
2.1.8.a Improved Post-Accident Sampling Capability 22 2.1.8.b Increased Range of Radiation Monitors 24 2.1.8.c Improved In-Plant Iodine Instrumentation 26 2.1.9 Analysis of Design and Off-Normal Transients 27 and Accidents 2.2 Operations 2.2.1.a Shift Supervisor's Responsibilities 30 2.2.1.b Shift Technical Advisor 32 2.2.1.c Shift and Relief Turnover Procedures 33 2.2.2.a. Control Room Access 34 2.2.2.b Onsite Technical Support Center 35 2.2.2.c Onsite Operational Support Center 37 2.2.3 Revised Limiting Conditions for Opera-38 tion of Nuclear Power Plants Based Upon Safety System Availability
- 3.
Response to Additional Followup Action Recommendations Identified in the September 13, 1979 Letter from D. G.
Eisenhut to All Operating Nuclear Power Plants 3.1 Instrumentation to Monitor Containment Conditions During the Course of an Accident 3.1.1 Containment Pressure Indication 39 3.1.2 Containment Hydrogen Monitor 40 3.1.3 Containment Water Level Indication 41 3.2 Reactor Coolant System High Point Vents 42 3.3 Emergency Preparedness Improvements 3.3.1 Emergency Plan Conformance to Regulatory Guide 1.101
TABLE OF CONTENTS (Cont'd)
Page No.
3.3.2 Incorporation of Plant Instrumentation 45 into Emergency Plan Action Level Criteria 3.3.3 Emergency Operations Center Requirements 46 3.3.4 Improved Offsite Monitoring Capabilities 47 3.3.5 Compatibility of Federal, State, Local, 48 and Utility Emergency Plans 3.3.6 Test Exercises of Approved Emergency Plans
- 49.
Appendix 1:
Summary of Design Criteria Appendix 2:
Summary of Plant Modifications Schedule
1A
(-1 Purpose This report provides Southern California Edison Company's response for San Onofre Unit 1 to the September 13, 1979 letter from D. G. Eisenhut of the Nuclear Regulatory Commission to all operating nuclear power plants.
ScoDe Section 2 of this report responds to all TMI-2 Lessons Learned Task Force Short Term Recommendations as contained in NUREG 0578 as modified by the September 13, 1979 letter identified above.
The organization of this section corresponds to the enumeration of recommendations in NUREG 0578.
Section 3 of this report responds to additional recommendations for contain ment instrumentation, reactor coolant system vents, and emergency plan improvements as contained in Enclosures 3, 14 and 7 to the September 13, 1979 letter identified above. The organization of this section is consistent with the position requirements of Enclosures 3, 4 and 7.
Appendices 1 and 2 'provide summaries of the design criteria and implementation schedules, resp ctively for stationmodifications which are being implemented as discussed in Sections 2 and 3.
Southern California Edison Company was represented atthe September 26, 1979 08 Las Vegas, Nevada Regional Meeting concerning TMI Short-Term Implementation Action. In addition, the Company was represented at a series of Topical Meetings concerning further clarification of the Onsite Technical Support Center, Shift Technical Advisor, Relief and Safety Valve Performance Testing, and Reactor Coolant System Venting held in Bethesda, Maryland October 10-12, 1979.
The information contained in this report reflect consideration of the NRC guidance provided in those meetings.
RECOMMENDATIONS
-2 Section 2.1.1 -
Emergency Power Supply Requirements for the Pressurizera Heaters, Power-Operated Relief Valves and Block Valves, and Pressurizer Level Indicators in PWRs A. Positions on Pressurizer Heater Power Supply Position 1:
The pressurizer heater power supply design shall provide the capability to supply, from either the offsite power source or the emergency power source (when offsite power is not available), a predetermined number of pressurizer heaters and associated controls necessary to establish-and maintain natural circulationat hot standby conditlons," the required heaters and their controls shall be connected to the emergency buses in a manner that will provide redundant power supply capability. (Schedule:
Complete implementation by January 1, 1980.)
Response
The current station design complies with the stated position requirement as discussed below. All pressurizer heaters and associated controls can be supplied from either offsite power or the emergency diesel generators. In order to determine the minimum heater capacity and the time frame when the heaters must be available to maintain natural circulation, a study was performed by Westinghouse as authorized by the Westinghouse Owner's Group.
The results of this study conservatively indicate that for the 1300 ft3 pressurizer at San Onofre Unit 1 a heater capacity of 125 Iw applied within four hours is adequate to maintain reactor coolant system pressure, thus keeping the primary coolant subcooled and providing core cooling via natural circulation.
At San Onofre Unit 1, each emergency diesel generator can separately and independently energize a normal pressurizer heater group having a capacity of 117 KW and a backup pressurizer heater group having a capacity of 482 KW.
Accordingly, the power supply configuration at San Onofre Unit 1 assures that a predetermined number of pressurizer heaters and associated controls necessary to establish and maintain natural circulation cah be-connected in a manner that provides redundant power supply capabilty.
Position 2:
Procedures and training shall be established to make the operator aware of when and how the required pressurizer heaters shall be connected to the emergency buses.
If required, the procedures shall identify under what conditions selected emergency loads can be shed from the emergency power source to provide sufficient capacity for the connection of the pressurizer heaters.
(Schedule Complete implementation by January 1, 1980.)
-3 esponse:
s oPsto Based on the results of the study discussed in responsenoePostionmk above, operating procedures and training will be implemented which makes the operator aware of when and how to energize the pressurizer heaters. A review of emergency power loads indicates that sufficient diesel generator capacity exists such that load shedding is not required.
The operating procedures and training will be implemented by January 1, 1980.
Position 3:
The time required to accomplish the connection of the preselected pressurizer heaters to the emergency buses shall be consistent with the timely initiation and maintenance of natural circulation conditions.
(Schedule: Complete implementation by January 1, 1980.)
Response
The current station design complies with the stated position requirement as discussed below.
All pressurizer heaters may be energized from the control room by manually operating the pressurizer heater control switches. As discussed in the response to Position 1 above, the required pressurizer sheaters must be energized within four hours to prevent loss of subcooling.
To provide additional conservatism for maintenance of natural circulation conditioi, the procedure revisions discussed in response to Position 2 above will include a requirement to be able to energize the required pressurizer heaters within one hour.
One hour provides sufficient time to perform the manual operation of energizing the pressurizer heaters.
Position 4:
Pressurizer heater motive and control power interfaces with the emergency buses shall be accomplished through devices that have been qualified in accordance with safetygrade requirements.
(Schedule: Complete implementation by January 1, 1980.)
Response
The current station design complies with the stated position requirement as discussed below. The design, construction and operational characteristics of the pressurizer heater interfaces are the same as those utilized for safety-related interfaces with the electrical buses.
-4 B. positions on power SunplY for Pressurizer Relief and Block Valves and Pressurizer Level Indicators Position 1:
Motive and control components of the power-operated relief valves (PORVs) shall be capable of being supplied from either the offsite power source or the emergency power source when the offsite power is not available.
(Schedule: Complete implementation by January 1, 1980.)
Resoonse:
The current station design complies with the stated position requirement as discussed below. The Pressurizer power operated relief valves (PORV's) (2) are spring loaded, nqrmally closed type and are normally supplied with station instrument air to open. The station instrument air system can be powered from either offsite power or the emergency diesel generators. In addition, the PORV's have a backup pneumatic supply from the station nitrogen system to ensure their functioning on loss of station instrument air. The backup pneumatic nitrogen supply is from a pressurized source which doesbnot relyon components depending on electrical power.
For each PORV, the pneumatic supply is controlled by a solenoid valve which is energized-to-open. The control circuits for these valves are supplied from independent vital buses.
The vital buses can be supplied from either offsite poweror the emergency diesel generators.
Position 2:
Motive and control components associated with the PORV block valves shall be capable of being supplied from either the offsite power source or the emergency power source when the offsite power is not available.
(Schedule:
Complete implementation by January 1, 1980.)
Response
The current station design complies with the stated position requirement as discussed below. The POR block valves are spring loaded, normally open and require station instrument air to close. The station instrument air system can be powered from either offsite power or the emergency diesel generators.
Although it is not required to meet the stated position requirement, a backup pneumatic supply from the station nitrogen system will be installed for each PORV block valve similar to that discussed in response to Position 1 above for the PORV's. This backup pneumatic supply will ensure the functioning of the valves following loss of station instrument air.
For each PORV block.valve, the pneumatic supply is controlled by a solenoid valve which is energizedtoopen. The control circuits for these val.ves are supplied froi independent vital buses. The vital buses can be supplied from either offsite power or the emergency diesel generators.
-5 The design criteria to be utilized for the modifications associated with the back up pneumatic supply are included in Appendix 1. The implementa tion schedule for the odificatio5 is included in Appendix 2. Based on the implementation schedule, engineering and procurement efforts will be completed by January 1, 1980.
In addition, completion of the construction efforts is expected
,o require approximately one month of which the last two weeks require a station outage. Accordingly, construction which does not require a station outage is scheduled to commence on January 1, 1980; construction will be completed during the first outage of sufficient duration, or during the next refueling outage which is now scneduled for March-April, 1980.
Position 3:
Motive and control power connections to the emergency buses for the P0RVs and their associated block valves shall be through devices that have been qualified in accordance with safety-grade requirements.
(Schedule:
Complete implementation by January 1, 1980.)
Response
The current station design complies with the stated position requirement as discussed below. The design, construction and operational characteristics of the PORV'c and their associated block valves interfaces are the same as those utilized for safety-related interfaces with the electrical buses.
Control power to the PORV's and PORV block valves is provided such that one PORV and its associated block valve is fed from the same vital bus. The other PRV and its associated block valve is fed from a separate vital bus.
Each of these vital buses can be supplied from offsite or emergency onsite power.
This strain" alignment configuration, in conjunction with the valve fail-safe positions (i.e., PORV's fail closed and block valves fail open),
provides for both single failure protection and redundancy.
Position 4:
The pressurizer level indication instrument channels shall be powered from the vital instrument buses. These buses shall have the capability of being supplied from either the offsite power source or the emergency power source when offsite power is not available. (Schedule: Complete implementation by January 1, 1980.)
Resoonse:
The current station design complies with the stated position requirement as discussed below. All pressurizer level indication instrument channels are currently powered from the vital buses.
These buses have the capability of being energized from either offsite power or the emergency diesel generators.
-o Section 2.1.2 -
Performance Testing for BWR and PWR Relief and Safety Valves position:
Pressurized water reactor and boiling water reactor licensees and applicants shall conduct testing to qualify the reactor-coolant system relief and safety valves under expected operating conditions for design basis transients and accidents. The licensees and applicants shall determine the expected valve operating conditions through the use of analyses of accidents and anticipated operational occurrences refereslcedain Regulatory Guide 1.70, Revision 2. The single failures applied to these analyses shall be chosen so that dynamic forces o the safety and relief valves are maximized. Test pressures shall be the highest predicted by conventional safety analysis procedures. Reactor coolant system relief and safety valve qualification shall include qualification of associated control circuitry, piping and supports as well as the valves themselves.
(Schedule: Submit program description and schedule by January 1, 1980 and complete test program by July, 1981.)
Resoon se:
A full scale prototype qualification testing prograd of relief and safety valves under expected operating conditions for design basis transients and accidents will be undertaken on an ndustrywide basis rather than by individual licensees. The Electric Power Research Institute (EPRI) is currently developing such a program. Southern California Edison Company will assist EPRI by providing financial support and technical assistance as requested through the Westinghouse Owner's Group. A program description and schedule for completion of the testing will be submitted prior to January 1, 1980.
-7 Section 2.1-3.-
Direct Indication of Power-Operated Relief Valve and SafetySValve Position for PWRs and BWRs Position:
Reactor system relief and safety valves shall be provided with a positive indication in the control room derived from a reliable valve position detection device or a reliable indication of flow in the discharge pipe.
(Schedule:
Complete implementation by January 1, 1980.)
Response
Both PORV's and their associated block valves have stem mounted limit switches which provide positive indications of the valve positions in the control room.
These switches are not currently safety related and do not have alarm functions. The pressurizer safety valves do not presently have direct position indications.
Currently, two alternative methods for providing safety related positive control room indications and alarms are under evaluation.
One method would utilize nonredundant safety related stem mounted limit switches on each of the valves with position indications in the control room.
In addition, an alarm will be provided in the control room which would indicate if any of the PORV's or safety valves are not fully closed.
No alarm function would be provided for the PORY block valves since they are normally open.
The other method would utilize a nonredundant safetyerelated acoustic device on the piping downstream of the PORV's and the safety valves with indication and alarm functions in the control room.
Based on the results of the evaluation, one of the above methods will be utilized.
The design criteria to be utilized for the modifications are included in Appendix 1. The implementation schedule for the modifications is included in Appendix 2. Based ol the implementation schedule, engineering and procurement efforts will be completed by January t, 1980.
In addition, completion of the construction efforts is expected to require approximately one month of which the last two weeks require a stationpae outage. Accordingly, construction which does not require a s-tation outage is scheduled to commence on January 1, 1980; zonstruction will be completed during the first outage of sufficient duration, or during the next refueling outage which is now scheduled fot March-April, 1980.
IIIII
-8 Sedtion 2.1.3.b -
Instrumentation for Detection of Inadequate Core Cooling In PWRs and BWRs Position 1:
Licensees shall develop procedures to be used by the operator to recognize inadequate core cooling with currently available instrumentation. The licensee shall provide a description of the existing instrumentation for the operators to use to recognize these conditions. A detailed description of the analyses needed to form the basis. for operator training and procedure development shall be provided pursuant to another short-term requirement, "Analysis of Off-Normal Conditions; Including Natural Circulation" (see Section 2.1.9 of this appendix).
In addition, each PWR shall install a primary coolant saturation meter to provide on-line indication of coolant saturation conditions. Operator instruction as to use of this meter shall include consideration that it is not to be used exclusive of other related plant parameters.
(Schedule:
Develop procedures and describe existing instrumentation and install a subcooling meter by January 1, 1980.)
Resoonse:
The Westinghouse Owners Group is currently in the process of developing generic procedure guidelines to enable the operator to recognize inadequate core cooling with existing instrumentation. OAll guidelines o developed will be appropriately incorporated into San Onofre Unit 1 procedures. The San Onofre Unit 1 procedure revisions are scheduled for completion by January 1, 1980, based on the current Westinghouse Owner's Group schedule for completion of the generic guidelines by October 31, 1979.
As described in our June 25, 1979 letter in Docket No. 50-206, station operators have already been instructed regarding instrumentation available to detect core voiding and verify natural circulation.
A controls grade primary coolant saturation recorder is currently scheduled for installation prior to January 1, 1980.
This recorder will receive input from a pressurizer pressure transmitter and a hot leg RTD from any reactor coolant loop.
The recorder will have switching capability to choose any one of thethree hot leg loop temperature signals. All pressure and temperature input signals are from safety related instrumentation which is not used for any other control or indication functions.
The recorder will be powered from a vital bus which has the capability of being energized from either offsite power or the emergency diesel generators.
The recorder will display hot leg temperature, the saturation temperature corresponding to measured pressurizer pressure and the margin to saturation, all in OF. The recorder will also provie an alarm signal for margin to saturation of less than 500 F.
-9 Additionally, procedures have been implemented requiring operator use of a saturation temperature/pressure curve, including instructions for its use and significance, indicating-the saturation pressure which would correspond to the hot leg temperature. Operator use of the saturation temperature/
pressure curve includes use of safety related instrumentation independent from that which will be used for the saturation recorder. This curve and its instructions will be utilized as a back-up to the primary coolant saturation recorder.
Operating procedures will be developed or revised, as appropriate, to include instructions as to use of the saturation recorder and its associated back-up curve.
The procedures will specify that the saturation recorder is not to be used exclusive of other related station parameters.
These procedures will be implemented coincident with placing the recorder in service.
Position 2:
Licensees shall provide a description of any additional instrumentation or controls (primary or backup) proposed for the plant to supplement those devices cited in the preceding section giving an unambiguous, easy-to interpret indication of inadequate core cooling. A description of the functional design requirements for the system shall also be included. A description of the procedures to be used with the proposed equipment, the analysis used in developing these procedures, and a schedule for installing the equipment shall be provided.
(Schedule: Submit new level instrument design by January 1, 1980, and complete installation by January 1, 1981.)
Response
The Westinghouse Owner's Group is currently performing analytical work in the area of inadequate core cooling to determine if additional instrumen tation or controls are necessary. The analytical work is scheduled to be completed by October 31, 1979.
Any instrumentation or controls determined to be necessary will be appropriately incorporated into the San Onofre Unit 1 design.
The functional design of any such additional instrumentation or controls will be submitted by January 1, 1980, along with a description of the procedures to be used with the equipment, and the analysis used in developing these procedures. However, the installation of any instru mentation will be deferred pending completion of the integrated assessment of potential modifications identified by review of station design and operation in connection with the Systematic Evaluation Program (SEP).
SEP Review Topics which are expected to have a direct bearing on implementation of the stated position requirement include:
Topics III-4.C, III-5.A and 111-6.
For example, Topic III-5.A, Effects of Pipe Breaks on Structures, Systems and Components Inside Containment, will evaluate the effect of pipe breaks inside containment on the ability to safety shutdown and to mitigate the consequences of the pipe break. Based on the evaluation, pipe rerouting, additional supports and relocation may be required. Since installation of instrumentation is dependent on the layout inside contain ment and the potential requirement for pipe whip or jet impingement protec tion, this review topic must be completed prior to initiating engineering work. Similarly, completion of Topic III-4.C, Internally Generated Missiles; and III-6, Seismic Design Considerations will also likely result
-10 in new requirements and/or design criteria which affect the installation of instrumentation.
The assumptions and requirements to be utilized in the analytical work are the subject of continuing discussions between the Owner's Group and the NRC.
Accordingly, the completion dates discussed above are also subject to.
change based on the outcome of the discussions between the Owner's Group and the NRC.
Section 2.1.4 -
Containment Is'.lation Provisions for PWRs and BWRs Position 1:
All containment isolation system designs shall comply with the recommenda tions of SRP 6.2.4; i.e., that -here be diversity in the parameters sensed for the initiation of containment isolation.
(Schedule: -Complete implementation by January 1, 1980.)
Response
The Containment Isolation Signal (CIS) at San Onofre Unit 1 will be modified to include diversity in the parameters sensed for initiation as recommended by SRP 6.2.4.
CIS currently takes place on containment pressure above 2 psig. The Safety Injection Actuation Signal (SIAS) will be incorported as the diverse initiation parameter.
The design criteria to be utilized for the modifications are included in Appendix 1. The implementation schedule for the modificati ons is included in Appendix 2. Based on the implementation schedule, engineering and procurement efforts will be completed by February 15, 1980. In addition, completion of the construction efforts is expected to require approxi mately three months of which the last seven weeks require a station outage. Accordingly, construction which does not -equire a station outage is scheduled to commence February 15, 1980? ionistruction will oecompleted during the first outage of sufficient dur-ion, or during the next refueling outage which is now scheduled for March-April, 1980.
Position 2:
All plants shall give careful reconsideration to the definition of essential and nonessential systems, shall identify each system determined to be essential, shall identify each system determied to be nonessential, shall describe the basis for selection of each essential system, shall modify their containment isolation designs accordingly, and shall report the results of the reevaluation to the NRC. (Schedule: Complete imple mentation by January 1, 1980.)
Response
The current station design complies with the stated position requirement as discussed below. A thorough review of the San Onofre Unit 1 contain ment isolation design was completed as.part of the Sphere Enclosure Project and assessment of compliance with 10CFR50 Appendix J. The results of this evaluation were provided in Attachment 2 of the enclosure to our April 21, 1976 letter in Docket No. 50-206 to the NRC which identified (1) containment penetrations by number, (2) isolation valves provided for such penetrations (including valve designation and location, and (3) whether the isolation valves are subject to type C leakage rate testing. In addition, tabular information was provided comparing each line/pentration with 10 CER 50, Appendix J requirements.
-12 An additional review of the containment design was completed in response to I.E. Bulletin No.79-06A and Revision 1 thereto. Our letter dated June 25, 1979 in Docket No. 50-206 to the NRC provided additional information to supplement our responses to I.E. Bulletin No.79-06A and Revision 1 thereto. This information identified the containment penetrations which are not isolated by a CIS and provided the basis for not isolating these.
Position 3:
All nonessential systems shall be automatically isolated by the containment isolation signal. (Schedule: Complete implementation by January 1, 1980.)
Response
The current station design complies with the stated position requirement as discussed below. All containment penetrations which are required to be isolated are isolated by a CIS. Where isolation is not warranted, the basis for nonisolation is as described in the referenced letters discussed in the response to Position 2, above.
Position 4:
The design of control systems for automatic containment isolation valves shall be such that resetting the isolation signal will not result in the automatic reopening of containment isolation valves. Reopening of containment isolation valves shall require deliberate operator action.
(Schedule: Complete implementation by January 1, 1980.)
Response
The design of control systems for automatic containment isolation valves will be modified to prevent automatic reopening of these valves upon reset of the CIS.
The design criteria to be utilized for the modifications are included in Appendix 1. The implementation schedule for the modifications is included in Appendix 2. Based on the implementation schedule, engineering and procurement efforts will be completed by February 15, 1980. In addition, completion of the construction efforts is expected to require approxi mately three months of which the last seven weeks require a station outage. Accordingly, construction which does not require.a station outage is scheduled to commence February 15, 1980; construction will be completed during the first outage of sufficient duration, or during the next refueling outage.which is now scheduled for March-April, 1980.
-13 Section 2.1.5.a -
Dedicated Penetrations for External Recombiners or Post-Accident Purge Systems Position:
Plants using external recombiners or purge systems for post-accident combustible gas control of the containment atmosphere should provide containment isolation systems for external recombiner or purge systems' that are dedicated to that service only, that meet the redundancy and single failure requirements of General Design Criteria 54 and 56 of Appendix A to 10 CFR Part 50, and that are sized to satisfy the flow requirements of the recombiner or purge system. (Schedule: Provide a system description and implementation schedule by January 1, 1980 and complete installation by January 1, 1981.)
Response
San Onofre Unit 1 uses a purge system for post-accident combustible gas control of the containment atmosphere. Based on the concerns surrounding the generation of hydrogen identified in I.E. Bulletin 79-06A and Revision 1 thereto, Westinghouse was requested to determine the amount of combustible gas which may be generated based on station specific design (e.g., use of stainless steel fuel cladding and hydrazine containment spray additive).
This evaluation is scheduled to be completed by November 1, 1979.
Based on the results of this evaluation, a system will be provided to adequately control post-accident combustible gas which may be generated inside containment. Two alternative methods which are being evaluated include (1) upgrading the existing purge system, and 2) installing hydrogen recombiners inside or outside containment.
A description of the method to control post-accident combustible gas will be provided by Janiary 1. 1980.
However, implementation of the modifi cations will be deferred pending completion of the integrated assessment of potential modifications identified by review of station design and operation in connection with the Systematic Evaluation Program (SEP).
SEP Review Topics which are expected to have a direct bearing on implemen tation of the stated position requirement include: Topics 111-2, III-4.A, III-4.C, III-5.A, III-5.B, 111-6, VI-5, VI-2.D and VI-3. For example, Topic 111-6, Seismic Design Considerations, will include the specification of a new seismic spectra for the purpose of seismic reevaluation. Since modifications will require seismic analyses which utilize the new response spectra, this review topic must be completed prior to initiating engi neering work. Similarly, completion of Topic 111-2, Wind and Tornado Loadings; III-4.A, Tornado Missiles; III-4.C, Internally Generated Missiles; III-5.A, Effects of Pipe Breakdown Structures, Systems and Components Inside Containment; III-5.B, Pipe Break Outside Containment; VI-5, Combustible Gas Control; VI-2.D, Mass and Energy Release for Postulated Pipe Breaks Inside Containment; and VI-3, Containment Pressure and Heat Removal Capability will also likely result in new requirements and/or design criteria which affect the post-accident combustible gas control system.
-1 4 Section 2.1.5.b -
Inerting BWR Containments San Onofre Unit 1 is a Westinghouse PWR; accordingly this item is not applicable.
-15 Section 2.1.5.0 -
Capability to Install Hydrogen Recombiner at each Light Water Nuclear power Flant--
No action is required on this position requirement at this time as discussed in the NRC letter dated September 13, 1979 to all operating nuclear power plants.
-16 Section 2.1.6.a Integrity of Systems Outside Containment Likely to Contain Radioactive Materials (Engineered Safety Systems and Auxilary Systems) for PWRs and WRs position:
'Applicants and licensees shall immediately implement a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as low-aspractical levels.
This program shall include the following:
- 1.
Immediate Leak Reduction
- a. Implement all practical leak reduction measures for all systems that could carry radioactive fluid outside of containment.
- b. Measure actual leakage rates with system in operation and report them to NRC.
- 2.
Continuing Leak Reduction Establish and implement a program of preventive maintenance to reduce leakage to aslowapractical levels. This program shall include periodic integrated leak tests at a frequency not to exceed refueling cycle intervals.
(Schedule:
Complete implementation by January 1, 1980.)
Response
The Recirculation System (RS); portions of the Containment Spray System (CSS); portions of the Chemical and Volume Control System (CVCS) including Letdown and Makeup; the Primary Coolant and Containment Atmosphere Sampling Systems and the Gaseous Radioactive Waste Systems have been identified as systems which process primary coolant, and could contain high level radioactive materials.
Programs for these systems will be implemented as discussed below. The Residual Heat Removal System is located entirely within the containment and is not included among the systems covered by these programs.
Procedures outlining the leak reduction measures will be implemented by January 1,- 1980.
In addition, procedures will be implemented by January 1, 1980 describing preventive maintenance to reduce leakage to asnlow-aS-practical levels, including periodic integrated leak tests at a frequency not to exceed refueling cycle intervals.
All practical leak reduction measures will be implemented for the above systems. Procedures describing the leak rate testing will be prepared and implemented.
However, the leak rate testing of the above systems can only be performed during a shutdown. Therefore, the initial measurements of leakage rates for the above systems will be performed during the first outage of sufficient duration, or during the next refueling outage which is now scheduled for March-April, 1980.
The actual leak rate measured will be reported to the NRC within 30 days following completion of the
-leak rate testing.
-17 Design Review of Plant Shielding and Environmental SectionQualificatio of Equipment for Spaces/Systems Which May BeUJsed in_?0 st-ACCidentODeratiOfls ititea sof a post-accident release of radioactivity equivalent With the assumption la o y Gtdsh3 n
b e as acivaletY to that described in Regulatory Guides 1.3 and 1. (i.e.,rfhe equiaent of 50,0 of the core radioiodine and 100-%
of the core noble gas activity contained in the primary coolant), each licensee shall perfo t
a radiation and ~iedin deignreview of the spaces around systems that may, as a and shielding desient, contain highly radioactive materials. The design review should identify the location of vital areas a we p p s,
thew contould ro t radwaste control stations, emergency power suppliesa motor control centers, and instrument areas, in which personne byuthe may be unduly limited or safety equipment may be unduly degraded by the radiation field during postaaccident operations of these systems.
radiatione fiedcuriss to vital areas and Each licensee shall provide for adequa ace ss c increased pema tor protectionl of safety equipment by design changes, icesdpraeto temporary shielding, or post-.accident procedural controls. The design ecti~~cton nosaYspdedrehneeded for review shall determine which types of corrective aConslae neded for vital areas troughout the facility.
(Schedule: CJaplete th1,eif review by January 1. 1980.
Implement plant modifications bJ 1981.)
Resoonse:
A radiation and shielding design review of the spaces around systems that could contain highly radioactive materials is being performed assuming.e that the equivalent of 50% of the core radioiodine and 100% of the core noble gas is contained in the primary coolant. The design review also assum es radiation levels are limited to less than 15 mr/hr for areas norequiri contain e 1yae less than 100 mr/hr for areas requiring ae be frequnt aosvcc pand less than 10 CER Part 20d fo oerpeasnt Based on the results of this review, installation dure revisions and/or or temporary hielding an/or apo ac tocdreaaevai e operational station modifications will be accomplished to provide adequ access to vital areas and protection of appropriate safety equipment.
Thedesgn evew will be completed and any station modifications willb ihdetiie toe N
b January 1, 1980.
However implementation of any ideifiato willNE b deered pending completion of the integrated assess en oftnilmodifications idebedferntified by review of station design and operationl in connection with the Systematic Evlaio rgrm(E?). SEP Review Topics which are expected to have a direct earing on implementA tion of the stated position requirement include: Topic 111-2, 1 IjA tionofC th-5.B and 111-6.
For example, Topic 111-2, Wind and Tornado Loading, will evaluate the design basis tornado winds nd e pes 6 d oa su or equipment in accordance with eeulatory Guides 1.76 and 1.117.
It is likely that some tornado critera ilbesbihe
(e.g., Regulatory Guide 1.76) since such criteria was not a design basis at Sn Oofr Unt 1 Sine te eecton f sheldig will require co nide ron i 1f trado loadings, this review topic must be completedi prior to initiating engineering workiInadd tion Toi a11-ew Seismic Design Considerations, will include the speefication-Siane sei resignpons etratiorsthe purpose of seismic reevaluationh Since the rectons ofetr edfor will require seismic analyses which utilize the new response speialdthis reie topic must be completed prior to initiating response spectra, Similarly, completion of Topic aII.
ipe Break Missiles; i1-4.C, Internally Geerated Miss lesew requirements and/or Outside Containment will also likely resut n of shielding.
design criteria which affect the erection
-19 Section 2.1.7.a -
Automatic Initiation of the Auxiliary Feedwater System for PWRs Position:
Consistent with satisfying the requirement of General Design Criterion 20 of Appendix A to 10 CFE Part 50 with respect to the timely initiation of the auxiliary feedwater system, the following requirements shall be implemented in the short term:
- 1. The design shall provide for the automatic initiation of the auxiliary feedwater system.
- 2.
The automatic initiation signals and circuits shall be designed so that a single failure will not result in the loss of auxiliary feedwater system function.
- 3.
Testability of the initiating signals and circuits shall be a feature of the design.
- 4.
The initiating signals and circuits shall be powered from the emergency buses.
- 5.
Manual capability to initiate the auxiliary feedwater system from the control room shall be retained and shall be implemented so that a single failure in the manual circuits will not result in the loss of system function.
- 6.
The a-c motor-driven pumps and valves in the auxiliary feedwater system shall be included in the automatic actuation (simultaneous and/or sequential) of the loads to the emergency buses.
- 7.
The automatic initiating signals and circuits shall be designed so that their failure will not result in the loss of manual capability to initiate the AFWS from the control room.
In the long term, the automatic initiation signals and circuits shall be upgraded in accordance with safety-grade requirements. (Schedule:
Complete implementation of control grade by January 1, 1980.
Complete implementation of safety grade by January 1, 1981.)
Response
Conceptual engineering efforts have been initiated to provide automatic actuation of the auxiliary feedwater system consistent with the above position requirements. However, final engineering, procurement and construction efforts required to fully implement the above position requirements will be deferred until (1) the NRC Bulletins and Orders Group completes their review of the auxiliary feedwater system, and (2) the completion of the integrated assessment of potential modifications
-20 nstation esig and operation in connection with t
h e S y tex a t T o p i c I I i mr os o.
S E P R e v i e w T o p i c s w h i c h a r e e x p e c t e d t o h a v e a d
m fi n o n o w h c o n a t i l z e t h e s t a e d p s i n a lA requirement include: Topics 111 b2 e completed prto k Outsid Cona6n sr e x am, Topi c 11-6 Seis ic esig Con.sid er ati ons t aill include the peniain of a n e m response spectra for the purpoh e of seismic tera i o n e in e m ification s
at tomate the auxiliary feedwater system will require seismic analyses which utilize feeie n re as in specra.
his evie topc mut be completed prior to jititn ia nginleering work. In addition, Topic oI5~
ip a
utside Contain-ayi ment, will reevalu ate the prev ious pipe vre a
d/otside e a t ana si l i acodance with current criteria. This will include the evlaino othecr hg enrylnsad econsideration of the acceptability of aug mented inservice inspection. This latte itemr patclrfafcsteed conairilentpeetrtin aeawhere the feedwater Pipes and auxilay ed woar en p pe ntrconc.
Pp eoation and/Or pipe break restraints may be pecesary. Since automation of the auxiliar area, this review mayle nmodifications in the containment penetration ae ai work.
involve m dbe completed prior to initiating final engioering wor.4.,
Simila ly com le io of To ic 111-2 Wind and Tornado Ladings ; SystemsA Tornado Missiles; 111-.C, nternally Generlated Misies, vi-3 Systemsl Required for Safe Shutdownd t
wich a ect also likely result in new requirements and/or eign criteriawih fec the auxiliary feedwater system.
-21 Section 2.1.7.b -
Auxiliary Feedwater Flow Ierators positn uirements set forth in GDC 13 to provide twth satisfying the requit ascertain the actual hefollowieo Consie a i
haot oo m t s intended function,th folwn the capability in the control ro t aseai th aculprrmne f
the AFWS when it is called to perform its itne ultOl h olwn requirements shall be implemented:
flow to each steam
- 1.
5f~tygrae inicaton f auxiliary feedwater fo oec ta
- 1..safety-grade indicat ided xi aontrol room.d generator shall be provi es shall be powere
- 2.
heauxliryfeewaerflow intrumet chanfln l e poeredc from lar th feelc abues consistent with stisfying the emre0
- 2.
1ry feedwater system set forth in Auxiliary Sysection 10.4.9*
Standard Review plan, Se control grade by anuary 1980 chedule Complete implementation of by January 1,
1980 cmp ulee ilentation Of safety grade b aur
,18.
auae:tliary feedwater system described in response to S e a toti o n d o f a e d 0 7 t h c o t oi s r e p o r t w i l l i n c l u d e a d d i t io o f conoI)oo f ndicatiof eh steam generator. The auxiliary feedter flow indication wilb osstent with the above exceptl ro mi l b o s ai u e c i e i n single req~ir~ets9 the tated ingle fiueciein fairemns criterio wilhe Set by av ifg separate and indepedetFo fidiciteo n
each stem geeaor5fce only one of three steaml g~~~~enerators srqie ormv deca hetfo the reactor coolant inystm.
t eai teach sta geeato has water lee indication displayed and alarmed in th cotlrom
-22 Secton.1..a ~~rovd ?0t~Acidnt amnin~ CaabilitY Section positf.
view of the reactor coolan and containment a
n w~ ota f l s th an r ho r) a sa p e nder ac idety atmospher s e i s a l b pe formed to deter ine t ne ac idenit A esgnand operat is alee shlspromhu) ample d d
a xcesitions oce mnt o a r a e r eve to ana individual in rem sp e am tl ofai (iss thsta n ed s m eeo teuri er a o udonnel to prold shol0ietpo vi xtremitie r e tures of 3 and
-3/ Re s to the wholeto p
erol A dpe i gne and Accde nt eaon itions should assume a egsiat or t
ex es ve Y. A ci en ro ncts f t e r v e additional design aeset tis1ioth nd rhus ce-hioue s pe sart olniamn hfe
- 1. release of fail re l o in sd he r eew h indica te t at e gla could not promptlY aind safelyb o idte samplesysi if eat e cl ad din g so u ld tb e p oi to h oul d te spc eri a.
e i n
p t h r e oi e w o f t h e i lo s. c a l p c ru r e a n h ly S t o on i r i ret oe m ine the capability t ouroc tl r onadoe dt ee etpe 1ha r e
sedic fA c res o shall b S perf d u e d i f i
by quantify (less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) crai radi~
oo e t
,h ar indicat or ofg t h e d e g r e e o f c o r e pam a g er S o h s a n u acres i a rey n o b l e g a e s ( w h ic h
indicate ladding failure) tio nle s (which
~ ~~erturs) and non vol il i sot os (whic itdcat fuet Oftfl teinitiarea ctor coolant spectrum w should actor cond to aReg ora ai t mo re ea m e. T e revies of diec raito fromese p igan coponents in the au x(iliary bild'n n
Gu5i e 1.
rn n ~
i n f o i b e effluents - I direc raditiO d dirctor equip o f prsontamination and die c an iy e ive l m iation o
f the review indicates that the nalse thued cno b efrmdi prompt manner with oisting equi
- tis, t e e ne c ite i a t i e n t p o c e w en t s h a l l b e u n d e r t a ke t o8 0.e t
ho ver a
a r y
m e n t p r c u r e o el o g i c a l a n a l y s e s
, c e r t a i n c h e m i c a s n a l l b e r o e d In addcheoical analyseso assuming a highly hn addition to Lthetra re modr c os e yo e t iSecsi to eror bron and chloride hm r oiv initial sample (Regulatory Guide
- e. od r o urcey boo sml aajiSwtina ou n theclrd sample analysis withi anayssit (S hedule: olete the deign review, preparereid and describel propoe procedures antd modifications by January 1, 1980 Implement plan mo if c ti n byonnuarinm98n.
doperational review ofe oolaterand co A d e i g n n s a m l i n 'sy t e m s w i l l b i e l y m a n n e r d u r i n g a c c i d e n t / 4 R S atmosphere
,apln st le s perforime d to eersofin the 3
ea to obtain s m l s i ineefduadal~1 Revi ins of personn el withou an epA y ro edrioft i
c o n d i t i o n s w i h u a nyO r e i i s m dc i e s a y a e u t t t o to the whole body or exrente d owpctielY A thpoedr rei os modi fications deterie to 80 He necesar as a e mp rr wrsaill be provided byJaur I
p mr review januaryar sta io 1980; howe of the i mlemra e nt t o of ny eewil b deered endngcompletifin breview Of station modifincatilloendef rredtior fie the r, -4a ated P o ra s h i e d i n o f p o t e t i a
) n w i t h t h e S y s t e m a t i c -
e u t o S e c t o r n asesmet nconnecti on d in the response oSct n
assertiossas isue design and op r t o in e r l Is e o t The basis f or thIs deerli(sdsus 216bOf NUREG -578 in this reot
-23 With respect to radiological spectrum and chemical analyses. the Westing house Owner's Group is currently studying methods to perform these analyses. The study will include development of guidelines for sample preparation, evaluation and recommendations regarding application of automatic or in-line analyses, review of alternative manual analysis procedures (including specification of equipment and shielding requirements), and specification of minimum capability for gamma spectroscopy equipment.
The tentative Westinghouse Owner's Group schedule for completing the study is March, 1980.
Following completion of the study, the results will be reviewed for application at San Onofre Unit 1 and methods will be developed to perform the radiological spectrum and chemical analyses.
These methods may involve procedure revisions and/or station modifications. Any procedure revisions and/or station modifications determined necessary to perform the analyses will be provided within 90 days from receipt of the Westinghouse Owner's Group's study. However, implementation of any necessary modifica tions (i.e., permanent or temporary shielding) will be deferred as discussed above.
-24 Section 2.1.8.b -
Increased Range of Radiation Monitors Position 1:
Noble gas effluent monitors shall be installed with an extended range designed to function during accident conditions as well as during normal operating conditions; multiple monitors are considered to be necessary to cover the ranges of interest.
- a.
Noble gas effluent monitors with an upper range capacity of 105 microcuries/cc (Xe-133) are considered to be practical and should be installed in all operating plants.
- b.
Noble gas effluent monitoring shall be provided for the total range of concentration extending from normal condition (ALARA) concentra tions to a maximum of 105 microcuries (Xe-133). Multiple monitors are considered to be necessary to cover the ranges of interest. The range capacity of individual monitors should overlap by a factor of ten.
(Schedule: Complete procedures by January 1, 1980; complete installation by January 1, 1981.)
Response
A noble gas effluent monitor will be provided such that capability to monitor the total range of concentration extending from normal condition (ALARA) concentrations to.a maximum of 105 microcuries/cc (Xe-133). The design criteria to be utilized for the monitor are included in Appendix 1.
The implementation schedule for the monitor is included in Appendix 2.
Based on the implementation schedule, engineering and procurement efforts will be completed by December 1, 1980.
In addition, completion of the construction efforts is expected to require approximately four.months of which the last two months require a station outage. Accordingly, con struction which does not require a station outage is scheduled to commence December 1, 1980; construction will be completed during the first outage of sufficient duration, or during the refueling outage which is now scheduled for September-October, 1981.
In addition, procedures will be revised or developed as necessary for estimating the noble gas release rates if the existing effluent instrumen tation goes off scale.
All release sources, such as the main steam safety valves, will be considered.
The procedures will be implemented by January 1, 1980.
Position 2:
Since iodine gaseous effluent monitors for the accident condition are not considered to be practical at this time, capability for effluent moni toring of radioiodines for the accident condition shall be provided with sampling conducted by adsorption on charcoal or other media, followed by onsite laboratory analysis.
(Schedule: Complete procedures by January 1, 1980; complete onsite laboratory analysis capability by January 1, 1981.)
-25 Re2ponse:
Stack effluent radiojodine samples are currently obtained by adsorption on Stac efluet raioidinds mes ae ulate cartridge sampcepare also charcoal crtridges. In additionl, particult sirtie. saT eae aaablsoy obtained. These samples are currently analyzed onsi ng acident condi for onsite laboratory analysis of these cartridges durins acc lton tions is currently being evaluated, and the results of this evaluation, including a proposed method for meeting the intent of-the stated position reurment, will be provided by Januar y 1, 1980. However, the incluig airpooiddb auayi 9
ns (i.e*, permanent or implementation of any necessary station modification oe perintgr temporary hielding) will be deferred pending completion of the sntegrated assessmentof potential modifications identified bm review of station desgn nd pertio inconection with the Systematic Evaluation Program.
aesig an operatiodei ecann tin chseSterspseto Section The basis for this deferral is as discussed in the response t eto 2.1.6.b of NUREG-05 7 8 in this report.
In addition, procedures will be revised or developed as necessary by Jnadito, p uestimating the radiciodine release rates if the existing effluent instrumentation goes of sal.
llreles oucs such as the main steam safety valves, will be considered.
position 3:
Incontainment radiation level monitors with a maxie range of 108 rad/hr shal b intaled. A minimum of two such monitors that are phscal separated shall be provided. Monitors shall be Complte installation by function in an accident environment.
(Schedule:
Complet inaltonb January 1, 1981.)
Resp2nse:
Two radiation level monitors with a maximum range of 108 rad/hr will be installed in contaiment. The design criteria to be utilized for the monitors are included in Appendix 1. The implementation for the monitors is inlueda e addin AppndBased on the implementation schedule, is included in Append i
- t effort wil b e c m lec tm e to require engineering and procurement efforts will be comprts s e mer a 1980.
In addition, completion, of the construction efforts is expected to-require Inaddroite four months of which the last two months require a station ut cr in shu i
which does not require a station outage is scheduled to commence December g t9h costuciniluemleted during the first outage of sufficient duration, or during h e~ln outage which is now scheduled for September-October, 1981.
-26 Section 2.1..
ITproved In-Plant Iodine Instrumentation position :
Erovide equipment and associated training and Each licensee shall Proideum e
on oiecnet iai procedures for accurately determining the airborne iodine concenaione areas within the facility where plant personnel may be present during an accident.
(Schedule:
Complete implementation by January 1, 1980.)
Response:.
Equipment for gamma energy spectrum analysis to determine the airborne iodine concentrations currently exit tSnOor nt1 prtours for~tji~ig tisequipment to determine airborne iodine concentain will be reviewed and revised, as required, and the associated training completed by January 1, 1980.
-27 Section 2.1.9
-Analsis of Design and ff-Normal Transienlts and Accidents position:
the following are required:
Analyses, procedures, and training addressina
- 1. Small break loss-of-coolant accidents;
- 2. Inadequate core cooling, and
- 3.
Transients and accidents.
Some analysis requirements for small breaks have already been specified by the Bulletins and Orders Task Force.
These should be completed. In addition, pretest calculations of some of the Loss of Fluid Test (LOFT) small break tests (scheduled to start in September 1979) shall be performed as means to verify the analyses performed in support of the small break emergency procedures and in support of an eventual long term verification of compliance with Appendix K of 10 CFR Part 50.
In the anaysi of comdequate core cooling, the following conditions shall be analyzed using realistic (best-estimate) methods:
- 1. Low reactor coolant system inventory (two examples will be required -
LOCA with forced flow, LOCA without forced flow).
- 2. Loss of natural circulation (due to loss Of heat sink).
These calculations shall include the period of time during which inadequate core cooling is approached as well as the period of time durinG which adequate core cooling exists. The calculations shall be carried out in real time far enough that all important phenomena and instrument indications are included. Each case should then be repeated taking credit for correct operator action. These additional cases will provide the basis for developig appropriate emergency procedures. These calculations should also provide the analytical basis for the design of any additional instrumentation needed to provide operators with an unambiguous indication of vessel water level and core cooling adequacy (see Section 2.1.3.b in Thand accidents shall include the design basis Te anls oftrnseeet codnthaal inludela events specified in Section 15 of each ESAR.
The analysesshlinuda single active failure for each system called upon to function for a particular event.
Consequential failures shall alsoton behonideed Failures of the operators to perform required control manacti be ive cosidraton orpermutations of the analyses. Operator actions al ude cfse ten for permutof function of a safety system shall bea given cnserahe complete loss of. fuci need not address passive also be considered.
At present, these analyses need In the recent failures or multiple system failures in the short term.
anaysi ofsmal beakLOCAs, complete loss of auxiliary feedwater was considered.
The complete loss of auxiliary feedwater may be added tO h
0 falure beng cnsiered in the analysis of transients and accidents if fiuris conde thatmor is needed in operator training beyond the shortterm actions to upgrade auxilirfewarsytm~lbiiy
-28 Similarly, in the long term, multiple failures and passive failures may be considered depending in part on staff review of the results of the short-term analyses.
The transient and accident analyses shall include event tree analyses, which are supplemented by computer calculations for those cases in which the system response to operator actions is unclear or these calculations could be used to provide important quantitative information not available from an event tree.
For example, failure to initiate high-pressure injection could lead to core uncovery for some transients, and a computer calculation could provide information on the amount of time available for corrective action. Reactor simulators may provide some information in defining the event trees and would be useful in studying the information available to the operators. The transient and accident analyses are to be performed for the purpose of identifying appropriate and inappropriate operator actions relating to important safety considerations such as natural circulation, prevention of core uncovery, and prevention of more serious accidents.
The information derived from the preceding analyses shall be included in the plant emerge ncy procedures and operator training. it is expected that analyses performed by the NSSS vendors will be put in the form of emer gency procedure guidelines and that the changes in the procedures will be implemented by each licensee or applicant.
0 In addition to the analyses performed by the reactor vendors, analyses of selected transients should be performed by the NRC Office of Research, using the best available computer codes, to provide the basis for compara isons with the analytical methods being used by tChe reactor vendors.
These comparisons together with comparisons to data, including LOFT small_
break test data, will consititute the short-term verification effort to assure the adequacy of the analytical methods being used to generate emer gency procedures.
(Schedule: Analyses, procedural changes and operator training shall be provided following the schedule in Table B-2 of NtREG-0578.)
Response
Southern California Edison Company is participating as a member of the Westinghouse Owner's Group in the review of areas described in the above stated position.
The Owner's Group is performing generic analyses, and developing procedure guidelines to address these areas as discussed below:
- 1. The small break LOCA generic analyses and prepration of emergency pro cedure guidelines have been completed and a report (WCAP-9600) was submitted to the NRC on June 29, 1979.
Additional information regarding WCAP-9600 was submitted to the NRC by Owner's Group's letters dated September 11, 1979 and September 28, 1979.
Station specific small break LOCA analysis is being performied for San Onofre Unite 1 to conform the applicability of the generic results and is scheduled for completion by October 31, 1979. Station procedure revisions and operator training based on the applicable generic 0
procedure guidelines are scheduled for completion by January 1, 1980.
-29 0
- 2. Inadequate core cooling analyses are being performed and procedure guidelines developed by the Owner's Group. The results are scheduled to be submitted by October 31, 1979. Following completion of the generic analyses, an evaluation will be performed to determine the need for any station specific analysis.
Station procedure revisions will be completed by January 1, 1980 based on the applicable generic guidelines.
- 3. Analysis of transients and accidents will be performed and procedure guidelines developed by the Owner's Group. A short term generic program has been developed with results scheduled to be submitted by January 1, 1980. Short term station specific program results will be completed by April 1, 1980.
Revised station procedures and operator training will be completed by July 1, 1980.
- 4.
LOFT pretest calculations are being performed by the Owner's Group and will be submitted by November 15, 1979.
The completion dates for the inadequate core cooling analysis and tran sients and accidents analysis discussed in Items 2 and 3 above are subject to continuing discussions between the Owner's Group and the NRC in order to develop specific requirements. Accordingly, the completion dates for the related station specific review, procedure guidelines and operator training discussed in Items 2 and 3 above are also subject to change based on the outcome of the discussions between the Owner's Group and the NRC.
0
-30 Section 2.2.1.a -
Shift Supervisor's Responsibilities Position 1:
The highest level of corporate management of each licensee shall issue and periodically reissue a mariagemfernt directive that emphasizes the primary management responsibility of the shift supervisor for safe operation of the plant under all conditions o his shift and that clearly establishes his command duties.
(Schedule: Complete implementation by January 1, 1980.)
Response
The Vice President of Power Supply will issue a management directive by January 1, 1980 and annually thereafter to meet the above stated position requirement.
Position 2:
Plant procedures shall be reviewed to assure that the duties, responsi bilities, and authority of the shift supervisor and control room operators are properly defined to effect the establishment of a definite line of command and clear delineation of the command decision authority of the mshift supervisor in the control room relative to other plant management personnel. particular emphasis shall be placed on the following:
- a. The responsibility and authority of the shift supervisor shall be to maintain the broadest perspective of operational conditions affecting the safety of the plant as a matter of highest priority at all times when on duty in the control room.
The idea shall be reinforced that the shift supervisor should not become totally involved in any single operation in times of emergency when multiple operations are required in the control room.
- b. The shift supervisor, until properly relieved, shall remain in the control room at all times during accident situations to direct the activities of control room operators.
Persons authorized to relieve the shift supervisor shall be specified.
- c. If the shift supervisor is temporarily absent from the control room during routine operations, a lead control room operator shall be designated to assume the control room-command function. These temporary duties, responsibilities, and authority shall be clearly specified.
(Schedule: Complete implementation by January 1, 1980.)
Response
Administrative procedures will be reviewed and revised, as required, to assure that the duties, responsibilities, and authority of the shift supervisor and control room operators are properly defined to meet the stated position requirements. These procedures will be implemented by January 1, 1980.
-31 Position 3:
Training programs for shift supervisors shall emphasize and reinforce the responsibility for safe operation and the management function the shift supervisor is to provide for assuring safety.
(Schedule:
Complete implementation by January 1, 1980.)
Response
Training programs will be developed for the shift supervisors to meet the stated position requirement by January 1, 1980.
Position 4:
The administrative duties of the shift supervisor shall be reviewed by the senior officer of each utility responsible for plant operations.
Administrative functions that detract from or are subordinate to the management responsibility for assuring the safe operation of the plant shall be delegated to other operations personnel not on duty in the control room. (Schedule: Complete implementation by January 1, 1980.)
Response
The Vice President of Power Supply will review the administrative duties of the shift supervisor by January 1, 1980. Any administrative functions that are determined to detract from or are subordinate.to management responsibilities for assuring safe plant operation will be delegated to other operations personnel not on duty in the control room.
-32 Section 2.2.1.b -
Shift Technical Advisor Position:
Each licensee shall provide an on-shift technical advisor to the shift supervisor. The shift technical advisor may serve more than one unit at a multi-unit site if qualified to perform the advisor function for the various units.
The shift technical advisor shall have a bachelor's degree or equivalent in a scientific or engineering discipline and have received specific training in the response and analysis of the plant for transients and accidents. The shift technical advisor shall also receive training in plant design and layout, including the capabilities of instrumentation and controls in the control room. The licensee shall assign normal duties to the shift technical advisors that pertain to the engineering aspects of assuring safe operations of the plant, including the review and evaluation of operating experience.
Based on a reassessment of the stated position requirement by the NRC, entitled, "Alternatives to Shift Technical Advisors" (Enclosure 2, to the NRC letter dated September 13, 1979 to all operating plants) was also provided as additional guidance to meeting the intent of the stated position requirement. (Schedule:
Shift technical advisor on duty by January 1, 1980 and completely trained by January 1, 1981.)
Response
An on-shift technical advisor will be provided to meet the intent of the stated position requirement as clarified by the "Alternatives to Shift.
Technical Advisors" contained in the NRC letter dated September 13, 1979.
The advisor will be placed on shift by January 1, 1980 and completely trained by January 1, 1981.
-33 Section 2.2.1.c -
Shift and Relief Turnover Procedures Position:
The licensees shall review and revise as necessary the plant procedure for shift and relief turnover to assure the following:
- 1. A checklist shall be provided for the oncoming and offgoing control room operators and the oncoming shift supervisor to complete and sign. The following items, as a minimum, shall be included in the checklist:
- a.
Assurance that critical plant parameters are within allowable limits (parameters and allowable limits shall be listed on the checklist).
- b.
Assurance of the availability and proper alignment of all systems essential to the prevention and mitigation of operational transients and accidents by a check of the control console (what to check and criteria for acceptable status~shall be included on the checklist);
- c.
Identification of systems and components that are in a degraded mode of operation permitted by the Technical Specifications. For such systems and components, the length of time in the degraded mode shall be compared with the Technical Specifications action statement(this shall be recorded as a separate entry on the checklist).
- 2. Checklists or logs shall be provided for completion by the.
offgoing and oncoming auxiliary operators and technicians.
Such checklists or logs shall include any equipment under maintenance or test that by themselves could degrade a system critical to the prevention and mitigation of operational transients and accidents or initiate an operational transient (what to check and criteria for acceptable status shall be included on the checklist); and
- 3. A system shall be established to evaluate the effectiveness of the shift and relief turnover procedure (for example, periodic independent verification of system alignments).
(Schedule:
Complete implementation by January 1, 1980.)
Response
Procedures governing shift and relief turnover will be implemented to meet the stated position requirements by January 1, 1980.
In addition, a system will be established to evaluate the effectiveness of the turnover procedures by January 1, 1980.
Section 2.2.2.a -
Control Room Access Position:
The licensee shall make provisions for limiting access to the control room to those individuals responsible for the direct operation of the nuclear power plant (e.g., operations supervisor, shift supervisor, and control room operators), to technical advisors who may be requested or required to support the operation, and to predesignated NRC personnel. Provisions shall include the following:
- 1. Develop and implement an administrative procedure that establishes the authority and responsibility of the person in charge of the control room to limit access.
- 2. Develop and implement procedures that establish a clear line of authority and responsibility in the control room in the event of an emergency. The line of succession for the person in charge of the control room shall be established and limited to persons possessing a current senior reactor operator's license. The plan shall clearly define the lines of communication and authority for plant management personnel not in direct command of operations, including those who report to stations outside of the control room.
(Schedule: Complete implementation by January 1, 1980.)
Resoonse:
Administrative procedures will be developed and implemented to meet the provisions of the stated position requirements by January 1, 1980.
04
-35 Section 2.2.2.b -
Onsite Technical Support Center Position:
Each operating nuclear power plant shall maintain an onsite technical support center separate from and in close proximity to the control room that has the capability to display and transmit plant status to those individuals who are knowledgeable of and responsible for engineering and management support of reactor operations in the event of an accident. The center shall be.habitable to the same degree as the control room for postulated accident conditions. The licensee shall revise his emergency plans as necessary to incorporate the role and location of the technical support center.
Records that pertain to the as-built conditions and layout of structures, systems and components shall be stored and filed at the site and acces sible to the technical support center under emergency conditions. Exam ples of such records include system descriptions, general arrangement drawings, piping and instrument diagrams, piping system isometrics, electrical schematics, wire and cable lists, and single line electrical diagrams. It is not the intent that all records described in ANSI N45.2.9-1974 be stored and filed at the site and accessible to the tech nical support center under emergency conditions; however, as stated in that standard, storage systems shall provide for accurate retrieval of all pertinent information without undue delay. (Schedule: Establish center by January 1, 1980 and upgrade to meet all requirements by January 1, 1981.)
Response
An onsite technical support center (OTSC) is currently established in the Visitor's Viewing Area adjacent to the Control Room and is large enough to hold at least 25 individuals. The OTSC has communication links with the control room, operational support center (as described in response to Section 2.2.2.c of NUREG-0578 of this report), emergency operations center (as described in response to Requirement 3 of the Near Term Requirements for Improving Emergency Preparedness of this report) and the NRC (as described in the May 3, 1979 letter in Docket No. 50-206 submitted in response to I.E. Bulletin No.79-06A and Revision 1 thereto).
The OTSC is habitable to the same degree as the Control Room for postulated accident conditions.
Station records, including but not limited to, systems descriptions, general arrangement drawings, piping and instrument diagrams, piping system isometrics, and electrical drawings are currently stored onsite (in the Engineering Drawing Management (EDM) Center) in another building.
These documents and records are readily accessible to the OTSC under emer gency conditions.
Currently, the OTSC does not have the capability to display vital station technical data; however, a message control box (pass through type) is available which would allow the OTSC to gain access (without entering the
-36 sc anri of v ita l station technical data Whi~ miht e ipe bhar regineeings mf gmet personnel in support of ic rea t er tiolS in the event of an accident. Toa amp coe n the apa oiif o su p 1 reactor operati r to ns in the event of a n ccid n ite O a up oa t i of ato of a Technical Data Display ad uli S T be d g rn at v ing evalu tsed include (1) ue e
d r nepp nt n
ltehnaie lebe compl eeiso 2 ar ietio r sem s c a n a n d Z 0 0 m, c o l o r, c l o s e d c i r c i t e re v i s i o n s, c am e a i to o r display consoles having video tape plabc sro aiion 2 ehar wre instrumentation and recordersn, (3) a combination of 1 and 2.
h ehia Dt ipa n T h d uig e rsia to be utilized for e p Tec iec a e a tai e sp e nd e T he s n i se System are included in Appeie rd fo h ehia aaDslyad7~nfit System is included in Appendix
. qae o e imp emee ale, e gineerin and procurment efor pte ei c ol t e D ecembern9, 1980 In addition, completion fo t construconlefforts is expected to require a.pOiaeY fu ontsofhc the lat w months require a a station utagi
~ c o f l. c n t u t o w h i c h d o esa s t n o t r e q u i r e a t i o n wi l l b e c o pi mtsch d l d t o m n e D c m e 9 9 0 construction will be coml e e din dur oing ly her firs 198ag of s f'c u a i n or during th 0utge hic is nw s ceue fodupe bra ton, 191 In addition, sched led o c0 --
-suvisedn to e r ibe81 the existence du gthe a f re t outge f L and will be t e r - c ree cr b and fuE ergnctio ofa thwTC b i ll8 consistent with the ner tr th a n f r y m d 1 8 c s contained in the requirements fojmproving emere 1cY prareng e r September 13, 1979 H4RC letter to al prinpoepltS
-37 Section 2.2.2.0 -
Onsite Operational Suport Center Ansareanto be designated as the onsite operational support center shall be An area to be shall be separate from the control room and shall be the petoblhich t h peratiofls support personnel 'will report in an pesaelishhich en a
Co ncations with the control room shall be provied.y tueatironc plamn shall be revised to reflect the existence of the center and to establish the methods and lines of communication and management.
(Complete implementation by January 1, 1980.)
The current station-design complies with the stated position requirement as discussed below. An onsite operationa upport center scurrently available on the first floor of the Administration and Control uilding.
Communication with the control room via in-plant telephones is currently aviae.tIn addition, the San Onofre Emergency dlan will be revised to available. in and function of the center by mid-19 8 consistent with the near term requirements for imprvi ngemeoenaig power plants.
in the September 13, 1979 NRC letter to all opera
-38 Section 2.2.3 -
Revised Limiting Conditions for Operation of Nuclear Power Plants Based Upon Safety System Availability No action is required on this position requirement at this time as discussed in the NRC letter dated September 13, 1979 to all operating nuclear power plants.
- 3.
RESPONSE TO ADDITIONAL FOLLOWUP RECOMMENDATIONS IDENTIFIED IN THE SEPTEMBER 13, 1979 LETTER FROM D. G. EISENHUT TO ALL OPERATING NUCLEAR POWER PLANTS
-39 Secton 31 -Instrumentationl to Monitor Containment Conlditionls During the Course of an Accident 3.1.1 -
Containment Pressure Indication positionfalleprv A cntiuOU iniction of containment Pressure shalb providei thee contol oofi. Masuemet ad indication capability shall incud e s the otmes the Meap surem n o f the containment for con lc t au i me s.the
~~design pr fo tel n minus five psig foral cotimn.
Th eign andssualificaio poisions Of Regu l
Guie me.97 ~ldn de i adquali fication r l re ui~'
a d testability, s a l b e.
( c e u e Complete implementatinb aur,18.
Ceonanen prsueidcto instrumentation will be installed to meet th t ted position requirements.
r n l d d i The esin citeia to be utilized for the instrumettolr nlddi Appndi
- 1. Theimpemetaton schedule for the instrumentalini icue inAppendi Th dx 2. Based on the implementation schedule, gnern ancdd rcrnt pefr t ill be completed by December 19, 1980.
In.r additioncmetionfof te construction efforts is expected to requir appri#oimt letfour moth of which the last two months require a station appoxa tely cour Y onstcto wich does not require a station outage isschedul d incgeneDeemey1,190 construction will1.
be completed during the first utage of sufficientduain or uigth rfeln otage which is now scheduled for September-OCtober 91-
-40 3.1.2 -
Containment ydrO ont potatiot the containment A continuous indication of hydrogen concentration inesontan ant atpereshall be provided In the control roo.
M~easurement capability shaphbeaprovided over the range of 0 to 10% hydogesign and qualification shal be itive and negative ambient pressure. Th fication, redundancy povisos oRelatorY Guide 1.979 ncluding qualificain reudcy d
ait ha be ie (S heduludComplete implementation by January 1, 1981.)
Response
iment hydrogen concentration indication instrumentation will be Containmn hyrgncne psto euirements, installed to meet the stated arsition requir The esin crtera to be utilized for the instrumentation are included in Apedeign1 rTheimlmnainsede for the instrumentation is ilue inAppendi 2. Based on the implementation sceue.egnei included in A efforts will be completed by December 19t 1980. In addi and procurement efforte costucin efforts is eopce ths requirestio tion, completion of the construction last two m t require a station approximately four months of which the lat two nh require a station tg outae.
ccordingly, construction which does not eur tto outed isusheue d tcommence December 19, 1980; construction will be COMPlee ding heu irto t sufficient duration, or during the refueling outage which is now schedul
-41 3.13 Cotaimet Water Level Indication 3.1.3 -
otimn position A continuous indication of containment water level shall be provided in the control room for all plants. A narrow range instrument shall be provided for PWRs and cover the range from the bottom to the top of the containment sump.
Also for PWRs, a wide range instrument shall be provided and cover the range from the bottom of the containment to the elevation equivalent to a 500,000 gallon capacity. For BWRs, a wide range instrument shall be provided and cover the range from the bottom to 5 feet above the normal water level of the suppression pool. The wide range instrumentation shall meet the design and qualification provisions of Regulatory Gui de 1.97, includinlg qualification, redundancy, and testabil ity.
The narrow range instrumentation shall be qualified to meet the requirements of Regulatory Guide 1.89 and shall be capable of being periodically tested.
(Schedule: Complete implementation by January 1.
1981.)
Resoonse Containment wide range and narrow range water level instrumentation will be installed to meet the stated position requirements.
The design criteria to be utilized for the instrumentation are included inl Appendix 1. The implementation schedule for the instrumentation is included in Appendix 2. Based on the implementation schedule,' engineering and procurement efforts will be completed by December 19, 1980. In addition, completion of the construction efforts is expected to require approximately four months of which the last two months require a station outage.
Accordingly, construction which does not require a station outage is scheduled to commence December 19, 1980; construction will be completed during the first outage of sufficient duration, or during the refueling outage which is now scheduled for September-October, 1981.
-J4 2 Section 3.2 -
Reactor Coolant System High Point Vents Position Each applicant and licensee shall install reactor coolant system and reactor vessel head high point vents remotely operated from the control room. Since these vents form a part of the reactor coolant pressure boundary, the design of the vents shall conform to the requirements of Appendix A to 10 CFR Part 50 General Design Criteria. In particular, these vents shall be safety grade, and shall satisfy the single failure criterion and the requirements of IEEE-279 in order to ensure a low probability of inadvertent actuation.
Each applicant and licensee shall provide the following information concerning the design and operation of these high point vents:
- 1. A description of the construction, location, size, and power supply for the vents along with results of analyses of loss-of-coolant accidents initiated by a break in the vent pipe.
The results of the analyses should be demonstrated to be acceptable in accordance with the acceptance criteria of 10 CFR 50.46.
- 2.
Analyses demonstrating that the direct venting of noncondensable gases with perhaps high hydrogen concentrations does not result in violation of combustible gas concentrations limits in containment as described in 10 CFR Part 50.44, Regulatory Guide 1.7 (Rev. 1), and Standard Review Plan Section 6.2.5.
- 3.
Procedural guidelines for the operators' use of the vents. The information available to the operator for initiating or terminating vent usage shall be discussed.
(Schedule: Submit design description by January 1,1980; complete installation by January 1, 1981.)
Response
A design for reactor coolant system high point venting to meet the stated position requirements is currently being evaluated. A description of the design will be provided by January 1, 1980.
However, implementation of the design and the associated procedures for use will be deferred pending completion of the integrated assessment of potential modifications identi fied by review of design and operation in connection with the Systematic Evaluation Program (SEP).
SEP Review Topics which are expected to have a direct bearing on implementation of the stated position requirement include:
Topics III-4.C, III-5.A and 111-6.
For example, Topic III-4.C, Internally Generated Missiles, will evaluate the effects of postulated internally generated missiles on equipment and structures. Consideration of such missiles were not evaluated as part of the design basis for San Onofre Unit 1. Since the design of the venting systems will require consideration of internally generated missiles, this review topic must be
completed prior to initiating engineering work. In addition, Topic 111-6, Seismic Design Considerations, will include the specification of a new seismic spectra for the purpose of seismic reevaluation. Since the design of a venting system will require seismic analyses which utilize the new response spectra, this review topic must be completed prior to initiating engineering work. Similarly, completion of Topic III-5.A, Effects of Pipe Breaks on Structures, Systems and Components Inside Containment will also result in new requirements and/or design criteria which affect the design of a venting system.
Section 3.3 -
Emergency Preparedness Improvements 3.3.1 - Emergeflcr Plan Conformance to Reulato r Guide 1.101 position:
pergency plans to satisfy Regulatory Guide 1.101, with Upgrade licensee emegec paststiefeution level criteria based special attention to the development of uniform action mid-1980.)
on plant parameters. (Schedule:
Complete Implementation by ResDonse:
The San Onofre Emergency Plan complies with the requirements contained in Regulatory Guide 1.101.
The plan will be revised by mid-19 80 to make use of the upgraded uniform action level criteria based on plant parameters.
These action level criteria are currently under development and will meet the intent of the requirements issued by the NRC.
-45 3.3.2 -
Incorporation of Plant Instrumentation into Emergency Plan Action Level Criteria Position:
Assure the implementation of the related recommendations of the Lessons Learned Task Force involving instrumentation to follow the course of an accident and relate the information provided by this instrumentation to the emergency plan action levels.
This will include instrumentation for post-accident sampling, high range radioactivity monitors, and improved in-plant radioiodine instrumentation. The implementation of the Lessons Learned Task Force's recommendations on instrumentation for detection of inadequate core cooling will also be factored into the emergency plan action level criteria.
Resnonse:
Information provided by plant instrumentation which is used to follow the course of an accident will be factored into the San Onofre Emergency Plan action levels by mid-198O.
Instrumentation for post-accident sampling, high range radioactivity monitoring, inplant radiodine monitoring, and detection of inadequate core cooling will be considered for inclusion in the action level criteria.
3.3.3 - Emergency Operations Center Requirements Position:
Determine that an emergency operations center for Federal, State and local personnel has been established with suitable communications to the plant, and that upgrading of the facility in accordance with the Lessons Learned Task Force's recommendations for an in-plant technical support center is underway. (Schedule:
(a) Designate location and alternate location and provide communications to plant by mid-1980, (b) Upgrade Emergency Operations Center in conjunction with in-plant technical support center by January 1, 1981.)
Resoonse:
(a) An emergency operations center for Federal, State and local personnel has been established at the San Clemente City Hall. The emergency opera tions center has been provided with a two-way automatic ring down circuit to the San Onofre Unit 1 technical support center. SCE, in concert with local authorities, is also conducting a search for an appropriate alternate emergency operations center. This effort is scheduled for completion by mid-1980.
(b) The emergency operations center will be further upgraded if additional requirements become known. For the present, the established emergency operations center meets all defined in-facility requirements.
33.4 - Improved Offsite Monitoring Capabilities Position:
Assure that improved licensee offsite monitoring capabilities (including additional thermoluminescent dosimeters or the equivalent) have been provided for all sites.
(Schedule: Complete implementation prior to mid-1980.)
Response
Additional TLD monitoring locations will be established by mid-1980 to meet this requirement.
3.3.5 - Compatibility of Federal, State, Local, and Utility Emergency Plans Position:
Assess the relationship of State/local plans to the licensees' and Federal plans so as to assure the capability to take appropriate emergency actions.
Assure that this capability will be extended to a distance of ten miles.
This item will be performed in conjunction with the Office of State Programs and the Office of Inspection and Enforcement. (Schedule: (a)
Against current criteria by mid-1980, (b) Against upgraded criteria by January 1, 1981.)
Response
(a) The NRC has assessed the State of California/Local Emergency Response Plans and concurred on August 15, 1978. In addition, the NRC has recently assessed the relationship of these plans to the San Onofre Emergency Plan and Federal plans. Southern California Edison is now awaiting the results of that review and will cooperate with Federal, State and local agencies to provide the capability to take appropriate emergency actions. Any revisions determined to be required for the San Onofre Emergency Plan based on this review will be completed by mid-1980.
(b) SCE has contacted State, County and local authorities and requested that a joint revision of all emergency response plans be conducted to meet upgraded criteria by January 1, 1981.
This will establish new emergency planning zones in accordance with forthcoming State and Federal requirements. Southern California Edison will cooperate with the Office of State Programs and the Office of Inspection and Enforcement to ensure compliance with this requirement.
-9 3.3.6 -
Test Exercises of Aporoved Emergency Plans Position:
Require test exercises of approved emergency plans (Federal, State, local and licensees), review plans for such exercises, and participate in a limited number of joint exercises. Tests of licensee plans will be required to be conducted as soon as practical for all facilities and before reactor startup for new licensees. Exercises of State plans will be performed in conjunction with the concurrent reviews of the Office of State Programs. As a preliminary planning bases, assume that joint test exer cises involving Federal, State, local and licensees will be conducted at the rate of about ten per year, which would result in all sites being exercised once each five years. Revised planning guidance may result from the ongoing rulemaking, (Schedule:
(a) Conduct Test of licensees emergency plans by mid-1980, (b) Conduct Test of State emergency plans by mid-1980, (c) Conduct Joint Test exercise of emergency plans (Federal, State, local, licensee) for all operating plans within 5 years.)
ResDonse:
(a) A test of the San Onofre Emergency Plan will be scheduled for completion by mid-1980.
(b) Southern California Edison will request that a test of the California Emergency Response Plan be conducted concurrently with the San Onofre Emergency Plan test.
(c) Southern California Edison will cooperate with the NRC and Federal Agencies to schedule and conduct a joint test exercise of emergency plans within 5 years.
APPENDIX 1:
SUMMARY
OF DESIGN CRITERIA
APPENDIX I DESIGN CRITERIA FOR SHORT-TERM MODIFICATIONS The following design criteria is applicable to the modifications described in the attached table as identified against the respective modifications.
- 1. Category A Structures, Systems and Comoonents Seismic Category A structures, systems, and components shall be designed for no loss of function when subjected to the design basis earthquake (DBE).
These structures, systems, and components shall also be designed to remain within the allowable stress limits when subjected to the oeratine basis earthquake (OBE).
The maximum free-field ground-motion acceleration for the DBE and OBE shall be at least O.5g and 0.25g, respectively, based on Section 9.2 of the San Onofre Unit 1 FSAR.
Analysis of the dynamic loads of seismic category A piping is accomplished using response spectrum or time-history approaches, which utilizes the natural period, mode shapes and appropriate damping factors of the particular system. The adequacy of design shall be such that there is no loss of function during and after the prescribed seismic disturbance, i.e., OBE, and DBE. Damping shall be taken at 2%.
- 2. Single Failure Design Items requiring single failure design shall be capable of withstanding a single active failure without loss of function.
- 3. Electrical Classification to IE This classification is applied to electrical equipment and systems that are essential to emergency reactor shutdown, containment isolation, reactor core cooling, and containment and reactor heat removal, or otherwise are essential in preventing significant release of radioactive material to the environment.
Class IZ systems and equipment criteria is described in IEEE Standard 308-1974.
- 4. Containment Building The highest containment flood to result from/an accident is the 10' elevation. All safety related systems and equipment, to be located inside the containment below this elevation, shall be designed to operate in a submerged condition.
- 5. Requiremen t
eencv Power All electrically operated equipment to be installed shall be operable from an emergency power source.
-2
- 6. Environmental Conditions All equipment to be installed shall be designed to withstand the environmental conditions of normal operation.
Normal Operation A.
In Containment at Reactor Coolant Loop
- 1. Pressure 14.7 psia
- 2. Temperature 70-120oF
- 3. Relative Humidity 50-100%
- 4. Integrated Dose 6 X 106 rads B. In Containment Separated from Reactor Coolant Loop
- 1. Pressure 14.7 psia
- 2. Temperature 70-1200
- 3.
Relative Humidity 50-100%
- 4. Integrated Dose 4 X 103 rads C. In Reactor Auxiliary Building and Spent Fuel Handling Building
- 1. Pressure 14.7 psia
- 2. Temperature 36-1040F (36-1400 in Turbine Lube Oil Area)
- 3. Relative Humidity 0-90%
- 4. Integrated Dose 3 X 103 rads D. All Other Areas
- 1. Pressure 14.7 psia
- 2. Temperature 36-104oF
- 3. Relative Humidity 0-90%
- 4. Integrated Dose 103 rads
- 7. Post LOCA Operation Equipment inside and outside the containment, which is required to be operable during and subsequent to a LOCA, shall be capable of operation in the following conditions.
-3 A.
Inside Containment
- 1. Pressure 50 psig
- 2. Temperature 70-291 0F
- 3. Relative Hmidity 50-100%
- 4. Integrated Dose 1 x 10 rads B. Outside Containment
- 1. Pressure 14.7 psig 0
- 2. Temperature 36-104 F
- 3. Relative Humidity 0-90%
- 4. Integrated Dose 3 X 107 rads
SE A Sn A
ton ensoe Rs SR A
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controls 2.1..
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APPENDIX 2:
SUMMARY
OF PLANT MODIFICATIONS SCHEDULE
- 19.
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-5 ENGINEER1M 10-19 12-28 PROCU[U 1 II CWSTRUCTIM NO~TES:
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-17 12-28T I
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