ML19331D665
| ML19331D665 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 08/29/1980 |
| From: | Novarro J LONG ISLAND LIGHTING CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0578, RTR-NUREG-578 SNRC-502, NUDOCS 8009030437 | |
| Download: ML19331D665 (80) | |
Text
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'i 1200 ampfag. SHOREHAM NUCLEAR POWER STATION
.- o ,..,_ m u .j P.O. BOX 618 NORTH COUNTRY ROAD
- WADING RIVER. N.Y.11792 SNRC-503 August 29, 1980 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation Light Water Reactors, Branch 4 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 SHOREHAM NUCLEAR POWER STATION - UNIT 1 RESPONSE TO NUREG 0578 REQUIREMENTS
Reference:
D. B. Vassallo letter to "All Pending License Applicants", dated November 9, 1979 D. B. Vassallo letter to "All Pending Operat-ing License Applicants", dated September 27, 1979
Dear Mr. Denton:
Forwarded herein are fifteen (15) copies of Long Island Lighting Company's commitments to meet the requirements outlined in NUREG 0578 "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations" as clarified in NRC letters, dated September 27 and November 9, 1979. We are currently evaluating the requirements of NUREG 0694 "TMI- l Related Requirements for New Operating Licenses" and plan to ad-dress all the issues contained therein in a separate submittal to the NRC in the near future. However, since NUREG 0694 envelops the requirements of NUREG 0578, we hereby request your concurrence and/or any comments on the enclosed material as soon as possible. We consider that the actions taken at Shoreham, as described in the enclosed document, when finalized, will incorporate the Lessons Learned from the TMI-2 Accident. 80090aoy3/ THIS DOCUMENT CONTAINS P00R QUAL.lTY PAGES FC-493 5
Mr. Denton August 29, 1980 Page 2 If you have any questions, please do not hesitate to contact this office. Very truly yours, n -) ~5: \
. 1:
J. P. Novarro Project Manager
. Shoreham Nuclear Power Station t
LG/mt Enclosures cc: Mr. J. Higgins, NRC Site Trailer j .
- O 1 :
- SHOREHAM i NUCLEAR POWER STATION l DOCKET NO. 50 322 RESPONSES TO l
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NUREG 0578 l SHORT TERM LESSONS LEARNED .O REQUIREMENTS AUGUST 1980 Late o
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l l a SNPS-1 1
; RESPONSE TO NUREG 0578 l 1
O V TABLE OF CONTENTS INTRODUCTION RESPONSES , j 2.1.1. . . . . . . Emergency Power Supply Requirements for the Pressurizer Heaters, Power-Operated Relief Valves and Block Valves and Pressurizer Level Indicators in PWRs 2.1.2. . . . . . . Performanco Testing for BWR and PWR Relief and Safety Valves 2.1.3.a. . . . . . Direct Indication of Power-Operated Relief Valve and Safety Valve Position for PWRs and BWRs 2.1.3.b. . . . . . Instrumentation for' Detection of Inadequate Core Cooling in PWRs and BWRs
, 2.1.4. . . . . . . Diverse and More Selective Containment Isola-tion Provisions for PWRs and BWRs 2.1.5.a. . . . . . Dedicated Penetrations for External Recombiners or Post-Accident Purge Systems 2.1.5.b. . . . . . Inerting.BWR Containments 'I 2.1.5.c. . . . . . Capability to Install Hydrogen Recombiner at 4
Each Light Water Nuclear Power Plant 2.1.6.a. . . . . . Integrity of Systems outside Containment Likely to Contain Radioactive Materials (Engineered Safety Systems and Auxiliary Systems) for PWRs and BWRs (l 2.1.6.b. . . . . . Design Review of Plant Shielding of Spaces for l Post-Accident Operations ll 2.1.7.a. . . . . . Automatic Initiation of the Auxiliary Feedwater System for PWRs i 2.1.7.b. . . . . . Auxiliary Feedwater Flow Indication to Steam Generators for PWRs l- 2.1.8.a. . . . . . Improved Post-Accident Sampling Capability 2.1.8.b. . . . . . Increased Range of Radiation Monitors 2.1.8.c. Improved In-Plant Iodine Instrumentation O i 't i
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SNPS-1 RESPONSE TO NUREG 0578 TABLE OF CONTENTS (Continued) RESPONSES (Continued) 2.1.9. . . . . . . Analysis of Design and Off-Normal Transients and Accidents 2.2.1.a. . . . . . Shift Supervisor's Responsibilities 2.2.1.b. . . . . . Shift Technical Advisor 2.2.1.c. . . . . . Shift and Relief Turnover Procedures 2.2.2.a. . . . . . Control Room Access, 2.2.2.b. . . . . .Onsite Technical Support Center 2.2.2.c. . . . . .Onsite Operational Support Center 2.2.3. . . . . . . Revised Limiting Conditions for Operation of Nuclear Power Plants Based Upon Safety System Availability Containment Water Level High Point Venting Containment Pressure Containment H 2 Concentration 1 l f i I i 4 2-
SNPS-1 RESPONSE TO NUREG 0578 INTRODUCTION On March 28, 1979 an accident occurred at the Three Mile Island Nuclear Power Plant - Unit 2 (TMI-2). Shortly thereafter, the Nuclear Regulatory Commission (NRC) formed the Lessons Learned Task Force to identify and evaluate safety concerns originating with the TMI-2 accident that would require licensing actions for operating reactors, pending operating license and construc-tion permit applications. On July 19, 1979, the NRC issued NUREG-0578 "TMI-2 Lessons Learned Task Force Status Report and Short Term Recommendations". During the subsequent reviews of NUREG-0578 by the NRC and the ACRS, the NRC issued a letter on September 27, 1979, on the subj ect of Follow-up Actions Result-ing from the NRC Staff Reviews Regarding the Three Mile Island - O Unit 2 Accident. On November 9, 1979, the NRC issued clarifica-tions to some of the recommendations of NUREG-0578 resulting in modifications to some of the implementation commitments. Contained in this report are Long Island Lighting Company's l Responses and Commitments regarding the implementation of the requirements of NUREG-0578, as clarified in the above-mentioned l letters, in the Shoreham Nuclear Power Station. NUREG-0694 i
"TMI-Related Requirements For New Operating Licenses" is cur-rently under evaluation. Responses to requirements in NUREG-l 0694 not addressed in this report will be forwarded via a separate submittal to the NRC in the near future.
- - . . - . _ . . =. - ,
SNPS-1 RESPONSE TO NUREG 0578 INTRODUCTION (continued) The BWR Owners' Group is actively engaged in the evaluation of many requirements of NUREGs 0578 and 0694. Accordingly, LILCO will continue to monitor closely and participate actively in the efforts of the Owners' Group as well as the NRC and other industry groups. 4 l l
*e 4
SNPS-1 RESPONSE TO NUREG 0578 RESPONSES The LILCO Responses to NUREG 0578 Requirements are contained in individual sections of this report. Each Section has been numbered with the same numbering.. sequence used in the original NUREG 0578 document. The additional four ACRS concerns ap-pear at the end of the report. For completeness and informa-tion, each Section contains the NRC clarification and the BWR Owners' Group Discussion and Implementation Criteria. In-formation referenced in this report has been provided with the applicable section, wherever possible.
1 SNPS-1 RESPONSE TO NUREG 0578 p/ y 2.1. l' Emergency Power Supply Requirements for the Pressurizer Heaters, Power-Operated Relief Valves and Block Valves, i and Pressurizer Level Indicators in FWRs l NUREG 0578 POSITION: Consistent with satisfying the requirements of General Design Criteria 10, 14, 15, 17 and 20 of Appendix A to 10 CFR Part 50 for the event of loss of offsite power, the following positions shall be implemented: Pressurizer Heater Power Supply
- 1. The pressurizer heater power supply design shall pro-vide the cap. ability to supply, from either the off-site power source or the emergency power source (when offsite power is not available), a predetermined num-ber of pressurizer heaters and associated controls necessary to establish and maintain natural circula-tion at hot standby conditions. The required heaters and their controls shall be connected to the emer-gency buses in a manner that will provide redundant
/N power supply capability. \ ] '" 2. Procedures and training shall be established to make the operator aware of when and how the required pres-surizer 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 pres-surizer heaters.
- 3. The time required to accomplish the connection of the preselected pressurizer heater to the emergency busen shall be consistent with the tLaely initiation and i maintenance of natural circulation conditions.
- 4. Pressurizer heater motive and control power interfaces with the emergency buses shall be accomplished through devices that have been qualified in accordance with l safety-grade requirements.
Power Supply for Pressurizer Relief and Block Valves and ( Pressurizer Level Indicators
- 1. Motive and control components of the power-operated re-f-~3 lief valves (PORVs) shall be capable of being supplied ,_
from either the offsite power enurce or the emergency (\~ -) power source when the offsite power is not available. 2.1.1-1
SNPS-1 RESPONSE TO NUREG 0578 ('%
)
- 2. Motive and control comconents associated with the PORV block valves.shall be capable of being supplied from either the offsite power source or the energency power source when the offsite power is not available.
- 3. Motive and control power connections to the emergency buses for the PORVs and their associated block valves shall be through devices that have been qualified in accordance with safety-grade requirements.
- 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. NRC CLARIFICATION: Pressurizer Heater Power Supply
- 1. In order not to compromise independence between the sources of emergency pcser and still provide redundant capability to
;s provide emergency power to the pressurizer heaters, each
.( ) redundant heater or group of heaters should have access to only one Class IE division power supply.
- 2. The number of heaters required to have access to each emergency power source is that number required to maintain natural circulation in the hot standby condition. .
- 3. The power sources need not necessarily have the capacity to provide power to the heaters concurrent with the loads required for LOCA. -
- 4. Any change-over of the heaters from nornal offsite power to emergency onsite power in to be accomplished manually in the control room.
- 5. In establishing procedures to manually reload the pressurizer ,
heaters onto the emergency power sources, careful consider,ation must be given to: j
- a. Which ESF loads may be appropriately shed for a given p situation. -
- b. Reset of the Safety Injection Actuation Signal to permit the operation of the heaters.
fN + h c. Instrumentation and criteria for operator use to prevent overloading a diesel generator. t 2.1.1-2 i
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SNPS-1 RESPONSE TO NUREG 0578 O -
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- 6. The Class IE interfaces for main power and control power are to be protected by safety-grade circuit breakers. ^(See also Reg. Guide 1.75)
- 7. Being non-class IE loads, the pressurizer heaters must be automatically shed from the emergency power sources upon the occurrence of a safety injection actuation signal. (See item 5.b. above)
Power Supply for Pressurizer Relief and Block valves and Pressurizer ! Level Indicators
- 1. While the prevalent consideration from TMI Lessons Learned is being able to close the PORV/ block valves, the design should retain, to the extent practical, the capability to open these valves.
- 2. The motive and control power for the block valve should be i supplied from an emergency power bus different from that which supplies the PORV.
. 3. Any change over of the PORV and blocl. valve motive and control power from the normal offsite power tu the emergency onsite power is to be accomplished manually in the control room.
- 4. For those designs where instrument air is needed for operation, the electrical power supply requirement should be capable of being manually connected to the emergency power sources.
4 i-BWR OWNERS' GROUP DISCUSSION: 1 As discussed in NED0-24708, natural circulation in the BUR is strong and inherent in all off-normal modes of operation, inde-pendent of any powered system, as long as sufficient inventory is maintained. This is because even in normal operation the BWR is essentially an augmented natural circulation machine. Because the BWR operates in all modes with both liquid and steam in the reactor pressure vessel, saturation conditions are always maintained irrespective of system pressure (the BWR does not have a pressurizer). Thus there is no need'for emergency power to maintain natural circulation or to keep the system pressurized. I The power-operated relief valves in BWR's are already powered by emergency power. They have no block valves. The reactor vessel level indication instrument channels for safety system activation and control are already powered by
/~} emergency power. *
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. r 2.1.1-3
SNPS-1 RESPONSE TO NUREG 0578 BWR OWNERS' GROUP IMPLEMENTATION CRITERIA: For the reasons stated above, there is no need for action in response to Recommendation 2,1.1 for any General Electric BWR. LILCO'S RESPONSE: LILCO endorses the BWR Owners
- Group position, This requirement is not applicable to BWR plants such as Shoreham.
\ 2.1.1-4
SNPS-1 RESPONSE TO NUREG 0578 2.1.2 Performance Testing for BWR and PWR Relief and Srfetv Valves NUREG 0578 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 referenced in Regulatory Guide 1.70, Revision 2. The single failures applied to these analyses shall be chosen so that the dynamic forces on 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. NRC CLARIFICATION:
- 1. Expected operating conditions can be determined through the use of analysis of accidents and anticipated operational occurrences referenced in Regulatory Guide 1.70.
- 2. This testing is intended to demonstrate valve operability under various flow conditions, that is, the ability of the valve to open and shut under the various flow conditions should be demonstrated.
- 3. Not all valves on all plants are required to be tested. The valve testing may be conducted on a prototypical basis.
- 4. The effect of piping on valve operability should be included in the test conditions. Not every piping configuration is required to be tested, but the configurations that are tested should produce the appropriate feedback effects as seen by the relief or safety valve.
l l
- 5. Test data should include data that would permit an evaluation
, of discharge piping and supports if those components are not ' tested directly. l l 6. A description of the test program and the schedule for testing should be submitted by January 1, 1980.
- 7. Testine; shall be complete by July 1, 1981. ..
2.1.2-1
l l SNPS-1 l RESPONSE TO NUREG 0578 xms/ I l BWR OWNERS' GROUP DISCUSSION: The BWR Owners' Group has been performing detailed anhlysis
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and evaluation of accidents and anticipated operational occur-rences referenced in Regulatory Guide 1.70, Rev. 2 assuming single active component failure or single operator error. The results and conclusions of this work is anticipated to be sub-mitted to the Staff in September 1980. In addition, the Owners' Group nas acknowledged the alternate shutdown cooling event and has committed to flow testing of S/RVs under the low pressure liquid conditions anticipated for this event. LILCO'S RESPONSE: LILCO is participating in the S/RV performance verification test program being conducted on a generic basis by the BWR Owners' Group. The S/RVs utilized by Shoreham are being included in the test program. The Owners' Group has authorized General Electric Co. to proceed with the program which provides for fa-cility fabrication and testing to be performed by Wyle Labora-tories. Further details regarding this test program will be j submitted to the NRC by the Owners' Group. O General Electric, under contract with LILCO, has performed low pressure S/RV testing for Shoreham. This test was performed to demonstrate an alternate means of core cooling using the S/RVs as back up to shutdown cooling. The result of this test will be shared with the BWh Ownc.rs' Group and submitted to the NRC. l l l O- . 2.1.2-2
SNPS-1 RESPONSE TO NUREG 0578 2.1.3.a Direct Indication of Power-Onerated Relief Valve and f' Safety Valve Position for PWRs and BURS NUREG 0578 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. NRC CLARIFICATION:
- 1. The b.asic requirement is to provide the operator with unam-biguous indication of valve position (open or closed) so that
, appropriate operator actions can be taken.
- 2. The valve position should be indicated in the control room.
An alarm should be provided in conjunction with this indication.
- 3. The valve position indication may be safety grade. If the position indication is not safety grade, a reliable single channel direct indication powered from a vital instrument bus es may be provided if backup methods of determining valve position
( ) are available and are discussed in the emergency procedures as \~ / an aid to operator diagnosis and action.
- 4. The valve position indication should be seismically qualified consistent with the component or system to which it is attached.
If the seismic qualification requirements cannot be met feasibly by January 1,1980, a justification should be provided for less than seismic qualification and a schedule should be submitted for upgrade to the required seismic qualification.
- 5. The position indication should be qualified for its appropriate environment, (any transient or accident which would cause the relief or safety valve to lift). If the environmental qualifi-cation for this position indication will not be completed by January 1, 1980, a proposed schedule for comnletion of the environment qualification program should be provided.
BWR OWNERS' GROUP DIECUSSION-BWR safety and relief valves (S/RV) are arranged in three ways in the various operating reactors:
- 1. Valve discharges piped to the containment suppression pool; Ov 2.1.3.a-1
SNPS-1 RESPONSE TO NUREG 0578 O v
- 2. Valve discharges manifolded and piped to suppression pool;
- 3. Discharging directly to the drywell free volume, in pressure suppression containments, or to the containment free volume in dry containments.
The configuration of the valve discharge, and the operator's ability to diagnose and act on stuck-open valve events, will determine what information is to be provided in the control room. The environment experienced by the installed instrument-ation during a stuck-open valve event will determine the proper qualification requirements. Valve Discharges Individually Piped to the Suppression Pool All dual-function safety / relief valves and most relief valves are configured this way. Given a stuck-open valve, the contain-ment pressure will not increase because of the submerged discharge. There is benefit in direct indication, not only because the oper-ator would be given an early warning of S/RV discharge, but because he can attempt to reset a stuck-open valve from the control room. Most such valves have no external stem, which e recludes direct position indication. BWR OWNERS' GROUP IMPLEMENTATION CRITERIA: Valve Discharges Individually Pines to the Suppression Pool The Owners' Group considers two types of monitoring to be accept-able methods of positive valve indication: pressure switches in the valve discharge lines and acoustic monitors. A suitable pressure switch system is outlined in the Appendix (enclosed ,
. herein), in response to an NRC request in the September 24, 1979, ! Region I meeting.
Either type of system will be designed to the following broad requirements:
- 1. There will be at least one sensing device per discharge line;
- 2. Sensing devices may be either inside or outside the drywell;
- 3. Sensing devices and other components need not be qualified for a LOCA (pipe break) environment, but only for the environment expected during S/RV discharge to the suppression pool; l
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2.1.3.a-2 1 l l
SNPS-1 RESPONSE TO UUREG 0578 I ( ) 4. All components will be seismically qualified;
- 5. The system will be powered by one division of emergency power;
- 6. With sensing devices inside the drywell, non-class IE electrical penetrations may be used if insufficient IE penetrations are available.
LILCO'S RESPONSE: relief There are a total valves (S/RV) in the ofShoreham eleven (11) dual function Reactor System. safety The S/RVs in-stalled in this facility are of the Target Rock two-stage pilot operated design. Direct main stem position indication is not accessible in a valve of this type. Accordingly, positive po-sition indication will be provided utilizing pressure trans-mitters on each S/RV discharge line. The discharge of each safety / relief valve is indeoendently piped to approximately five (5) feet from the bottom of the suppression pool. The calculated steady state pressure near the valve dis-charge is in the range of 300 psig when the valve relieves at set pressure. This pressure is sufficiently high that a positive y and unambiguous signal is available with ample margin for toler-ances in calibration and variance in line cressure. When a valve recloses, pressure will return to normal in a fraction of a second. Thus, pressure measurement does not have the slow re-sponse time which characterizes discharge pipe temperature monitoring instrumentation. Since each valve discharge is in-dependently piped, the pressure signal provides unique indica-tion for the associated valve. Nonredundant safety-grade instrumentation will be provided to monitor pressure in the discharge pipe of each safety / relief valve. The transmitters will be located in the secondary con-tainment and connected to the S/RV discharge piping by instru-ment lines penetrating the primary containment. Individual dis-play and trip set point instrumentation will be provided for i each safety / relief valve in the main control room. A common alarm will also be provided in the control room to promptly alert the operator when any S/RV is open. The display instru-mentation will be located as close as possible to the safety / l relief valve control station in the main control room. In addition to being qualified for the environment exoected l during events resulting in safety / relief valve discharge to the suppression pool, the instrumentation will meet seismic
- Category I requirements in accordance with IEEE 344-1971 and ,_
l ()bepoweredfromaClassIEpowersupply. , ! 2.1.3.a-3 l
SNPS-1 RESPONSE TO NUREG 0578 The existing temnerature monitoring instrumentation will be l retained for its original function, de:ection of valve leakage
. conditions as backup / confirmatory indication for the pressure i instrumentation being provided.
l i i 1 i t i r I .i 1 1 I i ; 2.1.3.a-4
- - - _ _ _ _ . . _ - _ - . , - , _ _ _ _ _ _ _ _ _. -._._ _ _,, , , . . . . . - , _ _ _ _ , . _ _ _ _ _ _ . _ - . . _ _ _ _ _ . . ~ _ . _ . -
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2,.l.3A 2 . APPENDIX to OWMERS' GROUP FOSITIO.; 2.1.3A n USE OF DISCHARGE PIPING FRE55URE 5%ITCHis [Vl FOR S/R VAUIE POSITIG" M0?!ITORIGG_ Main staa positica is not accessible on the manufacturers Asigns of safety / relief valves in SWR service. General Electric has eveluated several conce ts including magnetic.or proximity suitenes, ace'.stic devices, temperature, and pressure switches. The use of pressure switches on the discharge lines has been selected as the most simple, direct and ;~ oven tecnnique for monitoring valve pas,ition. The Safety / Relief valvePressure discharge is piped to the torus, near the valve discharge discharging belcw the water line. it is in the range can be straightferaardly calculated and tested: This pressure is of 250 psig when the reactor is at rated p,c.ssure.sufficiently When the high with ample targin for tolerances in set point calicration. valve re-closes, pressure returns to normal 'in a fracticn of a second. 75.us : prm :ra : r-C 5s en hwe the slew re.9 ease tier which characterizes tamperature conitoring. O Test data are available confirmingThesethe transient dau were andobtained steady-state curingresponse extensive ( )s of S/R'l discharge line pressure. in-plant ceasurecents of suppression-pcol load 1ng resulting trem s relief valve actuaticns. selection of set points. Pressure switches are available in industry which are suitable for this service. Similar devices are'used rcutinely for the protection of plant l j and equipment. Plant perscnnel will be familiar with tne calioration, No development testing is . testing, and ma1ntenance of these devices. required to prove a satisfactcry device, other than qualification tests which would be required for any device. 1 With the use of pressure switches, no device is ecuntec on or near the safety / relief valve. The technique will It work for all types of piped will have no effect on valve BWR safety / relief valves in service. { performance. The pressure switches may be tocated at sc e distance frca the safety / relief valve where they willWhere not besuitacle subjected to severe piping penetrations temperature or vibration ccnditicns. are available it is possible to locate the switches outside the drywell. The pressure switches will be qualified for a 212cF,100% humidity environment. This is adequate for the intended service even if the
' pressure switches are inside the crywell because actuaticn of the' S/R valves, inadvertent or planned, will not cause these envircr ental 3
fm - ( l . RJ _
l ,12,1.3A . ccnditions to be exceeded. In the event of a small pipe break the. safety / relief valves in the ADS system would be initiated early in the transient, before degradaticn of the suitches could have cccurred. In the event of a large pipe break the safety / relief valves are not recuired to operate. tio failure mcde has been identi"ied that would result in an (rroneous indicatica that tne valve was opea. . The signals from pressure switches may be interfaced with indicating lights, control roca annunciators, an event counter, or the processEach computer.. Any one or all of these functions may be imolemented. safety / relief valve can be monitored independently of the other valves. D e O e e G
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SNPS-1 RESPONSE TO NUREG 0578 2.1.3.b Instrumentation for Detection of Inadecuate Core Cooling in PURs and BWRs NUREG 0578 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 instru-mentation 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 Cond.itions , 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 condition. Operator instruction as to use of this meter shall include consideration that is not to be used exclusive of other related plant parameters. ()-
- 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 cora cooling. A description of the functional design requirements for the system shall also be included. A description of the procedbres to be used in developing these procedures, and a schedule for installing the equipment shall be provided.
l NRC CLARIFICATION:
- 1. The analysis and procedures addressed in paragraph one above will be reviewed and should be submitted to the NRC 'for re-view..
- 2. The purpose of the subcooling meter is to provide a continuous indication of margin to saturated conditions. This is an '
important diagnostic tool for the reactor operators.
- 3. Redundant safety grade temperature input from each hot leg (or use of multiple core exit in T/C's) are required.
n 4. Redundant safety grade system pressure measures should be " l ( provided. t v J 2.1.3.b-1
SNPS-1 RESPONSE TO NUREG 0578
- 5. Continuous display of the primary coolant saturation conditions should be provided.
- 6. Each PWR should have: (A.) Safety grade calculational devices and display (minimum of two meters) or (B.) a highly reliable single channel environmentally qualified, and testable system plus a backup procedure for use of steam tables. If the plant computer is to be used, its availability must be documented.
- 7. In the long term, the instrumentation qualifications must be i required to be upgraded to meet the requirements of Regulatory Guide 1.97 (Instrumentation for Light Water Cooled Nuclear Plants to Assess Plant Conditions During and Following an Accident) which is under development.
- 8. In all cases appropriate steps (electrical, isolation, etc.)
must be taken to assure that the addition of the subcooling meter does not adversely impact the reactor protection or engi-neered safety features systems.
- 9. The attachment provides a definition of information required on the subcooling meter.
BWR OWNERS' GROUP DISCUSSION: Additional hardware to identify inadequate core cooling on BWRs has not been determined to be necessary at this time. Licensees' procedures will identify the diverse methods of determining inade-quate core cooling, using existing instrumentation. The results of analysis being performed in response to 2.1.9 will be factored into procedures as required, after the analysis is complete. Because the BWR operates in all modes with both liquid and steam in the reactor pressure vessel, saturation conditions are always
; maintained irrespective of system pressure. Thus there is no need for a subcooling meter in the BWR.
BUR OWNERS' GROUP IMPLEMENTATION CRITERIA:
}
- 1. Analyses and operaqpr guidelines for the detection and mitigation of inadequate core cooling are currently being developed per 1
Requirement 2.1.9 and questions from the Bulletins and Orders
! Task Force. These studies include an evaluation of currently i installed reactor vessel water level instrumentation, and the j possible use of other instrumentation, to detect inadequate core cooling. The need for further measures , if any, will be addressed
- after these analyses and operator guidelines are complete. Imple-mentation of emergency procedures and retraining will be done C, on a schedule consistent with those established with the Bulletins "
and Orders Task Force. - 2.1.3.b-2
SNPS-1 RESPONSE TO NUREG 0578 1
- 2. A subcooling meter, as required by Enclosure 6 of NUREG 0578 Implementation Letter of September 13, 1979 will not be provided.
LILCO'S RESPONSE: LILCO concurs with the BWR Owners' Group position.
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- 1. The BWR Owners' Group, of which LILCO is a member, and General Electric, have participated in the preparation of a report entitled, " Additional Information Required for NRC Staff Generic Report on Boiling Water. Reactors",
'NEDO-24708, August 1979. The NEDO-24708 document, to-gether with information subsequently supplied by the BWR Owners' Group, presents generic operator guidelines for dealing with scenarios which have the potential for leading to inadequate core cooling. The operator guide-lines were developed using " state-of-the-art" analytic techniques. In addition to the generic guidelines, NEDO-24708 describes instrumentation and methods that can be used by reactor operators to detect inadequate core cooling. Safety and backup systems that can pre-vent or mitigate the consequences of inadequate core s j j cooling are also addressed in NEDO-24708. Procedures specifically applicable to Shoreham, for detection and mitigation of inadequate core cooling, will b~e developed prior to Shoreham's startup. The operator training pro-gram for Shoreham will address the use of Shoreham pro-cedures to detect, and deal with, inadequate core ecol-ing.
- 2. The BWR, unlike the PWR, operates with both liquid and steam in the reactor pressure vessel. Since a satura-tion condition is always maintained, regardless of sys-tem pressure, the concept of a saturation margin meter is not applicable to BWRs.
l l l b~_- - 2.1.3.b-3
SNPS-1 RESPONSE TO NUREG 0578 2.1.4 Containment Isolation Provisions for PURs and BWRs NUREG 0578 POSITION:
- 1. All containment isolation system designs shall comply with the recommendations of SRP 6.2.4; i.e., that there be diversity in the parameters sensed for the initiation of containment isolation.
- 2. All plants shall give careful reconsideration to the definition of essential and non-essential systems, shall identify each system determined to be essential, shall identify each system determined to be non-essential, shall describe the basic for selection of each essential system, shall modify their contain-ment isolation designs accordingly, and shall report the results of the re-evaluation to the NRC.
- 3. All non-essential systems shall be automatically isolated by the containment isolation signal.
- 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 O valves. Reopening of containment isolation valves shall require deliberate operator action.
NRC POSITION CLARIFICATION:
- 1. Provide diverse containment isolation signals that satisfy safety-grade requirements.
, 2. Identify essential and non-essential systems and provide results to NRC.
- 3. Non-essential systems should be automatically isolated by l containment isolation signals.
l
- 4. Resetting of containment isolation signals shall not result in the automatic loss of containment isolation.
l BWR OWNERS' GROUP DISCUSSION: There is diversity in the parameters sensed for the initiation of BWR containment isolation. Following an isolation, deliberate operator action is required to open valves in most cases. 2.1.4-1
SNPS-1 RESPONSE TO NUREG 0578 s BWR OWNERS' GROUP IMPLEMENTATION CRITERIA:
- 1. Diversity of parameters sensed for the initiation of contain-ment isolation shall be provided in accordance with SRP 6.2.4.
- 2. A review shall be made of all systems penetrating primary containment to identify all essential systems. The basis of such classification shall be documented and supplied to the NRC.
- 3. All systems not identified as essential will be reviewed. If automatic isolation is not provided, juscification for not isolating will be presented to the NRC.
- 4. Licensees will review and modify isolation control systems and administrative controls, as appropriate, such that no isolation valve will open when the isolation logic is reset. Those plants that have valves that will automatically open when the isolation logic is reset will change the isolation logic to prevent the valves from opening when reset. Admin-istrative controls to prevent valves from reopening will be implemented by 1/1/80; logic modifications will be implemented by 1/1/81.
LILCO'S RESPONSE: A review of all systems penetrating the containment has been performed to:
- a. ensure that diverse containment isolation signals that satisfy safety-grade requirements are provided;
- b. identify essential and non-essential systems;
- c. ensure that non-essential systems are automatically isolated by containment isolation signals; and
- d. ensure that resetting of containment isolation signals does not result in the automatic loss of containment isolation.
Fenetrations not provided with automatic isolation or diversity of parameters sensed for the initiation of containment isolation signal are under evaluation. For tlsse penetrations, either design modifications will be made to provide automatic I isolation and/or diversity of initiating signal, or appropriate justification will be provided. The penetrations determined not to require automatic isolation or diversity of signal,
, along with the justification for this approach will be ..
) identified in a supplement to this report.
j , 2.1.4-2
SNPS-1 RESPONSE TO NUREG 0578 The design of the isolation system for the lines penetrating the Shoreham primary containment conforms to the intent of 10CFR50, Appendix A, General Design Criteria 54, 55, 56 and 57. A descrip-tion of the isolation provision for each containment penetration is provided in FSAR Table 6.2.4-1, enclosed herein. A number of specific signals are used for isolation of various process and safety systems. A summary of the containment isolation signals is given on page eight of this table. The detail # of the lines penetrating the containment are presented on FSAR Figures 6.2.4-1, 2 and 3, also enclosed in this report. Essential and non-essential systems, as defined for containment isolation purposes, are identified in Table 2.1.4-1. Essential systems are those that may be needed within ten minutes of a LOCA, a normal reactor scram or a scram system failure. All other systems are designated as non-essential. All non-essential systems with possible release path are either isolated auto-matically by isolation signals, by check valves that would prevent flow out of containment, or by remote manual operated valves which are closed during normal operation. The design review.of control systems for automatic isolation valves demonstrated that the resetting of the isolation signal (s) will not result in the automatic reopening of containment isolation j valves. This criterion is met in all cases except the outboard d feedwater testable check valves. These valves, upon receipt of an isolation signal, receive a spring assist in the close direction. Resetting of the isolation signal will remove spring assist, but will not provide an opening force. Therefore, a design modification based on this criteria is not warranted. ' m/
) 2,1.4-3
j TABL" ^31 ,4-1
- SYSTEMSCb.sdIFICATION 4
PRIMARY CONTAINMENT PENETRATION I NUMBER SYSTEM ESSENTIAL N0_N-ESSENTIAL i l X-1A,B,C,D Main Steam X I Main Steam Line Drain and MSIV-Leakage Control System I X-2A Feedwater X
- X-2B Feedwater X X-3 Main Steam Line Drain X X-4 RWCU Line from RPV X X-5 RIIR Shutdown Cooling from RPV X 4 X-6A,B RHR Injection Line to Recirculation X System Return X-7A,B RilR-Containment Spray Drywell X i
l X-8A,B RllR-Containment Spray Suppression X l Chamber X-9A,B,C,D RllR Pump Suction X X-10A RllR Test Line Return to Suppression X i' Chamber, Suppression Pool Cleanup Return, X RIIR Steam Condensing Discharge, X RHR Minimum Flow, X Core Spray Test Line, and X Core Spray Minimum Flow X
P RIILF' ^ ' CONTAI.$:T [ ;
)
PENETR/u AON
- z'--
NUMBER SYSTEM ESSENTIAL NON-ESSENTIAL X X-10B RHR Test Line Return to Suppression Chamber, RCIC Minimum Flow, X HPCI Minimum Flow, X RHR Steam Condensing Discharge, X
. RHR Minimum Flow, X Core Spray Test Line, X Core Spray Minimum Flow, and X Relief Valve Discharge from RHR Supply to RCIC Pump Suction X X-11 RHR - Head Spray Line to RPV X
-X-12 HPCI Turbine Steam Inlet Line X X-13 HPCI Turbine Exhaust X X-15' HPCI Pump Suction X X-16 RCIC Turbine Steam Inlet Line X X-17 RCIC Turbine Exhaust X - X-18 RCIC Vacuum Pump Discharge X X-19 RCIC Pump Suction X - X-20A,B Core Spray Pump Discharge to RPV X X-21A,B Core Spray Pump Suction X X-22A,B RBCLCW to Recire. Pump and Mo' tor X Coolers X-23A,B RBCLCW from Recire. Pump and Motor X Coolers 4 O V OV' PRIMAk CONTAINMENT PENETRATION NUMBER SYSTEM ESSENTIAL NON-ESSENTIM-X-24A to H RBCLCW to Drywell Unit Coolers X X-25A,B RBCLCW from Drywell Unit Coolers X X-26 Purge Air to Drywell X X-27 Purge Air from Drywell X X-28 Purge Air to Suppression Chamber X -X-29 Purge Air from Suppression Chamber X X-30 Sample Coolant from RPV X X-31 Equipment Drains from Drywell X X-32 Floor Drains from Drywell X X-36 Standby Liquid Coolant to RPV X X-37A Nitrogen / Air Purge for TIP X X-37B,C,D TIP Drive Guide Tubes , X X-38 TIP Drive Guide Tubes X X-39A,B Instrument Air to Suppression X Chamber
~
X-41 HPCI Vacuum Breaker X X-42 'RCIC Vacuum' Breaker X X-43 RHR Relief Valve Discharge Vacuum Breaker, X RHR Heat Exchanger Vent, x RHR Heat Exchanger (RV) , and y~ HPCI . Steam Heat Supply (to Exchanger RV)RHR X PRIM') ],'
\
CONTAIhedNT s/ PENETRATION NUMBER SYSTEM ESSENTIAL NON* ESSENTIAL X-44 Containment Atmospheric Control from X Suppression Chamber, and Drywell Floor Seal Pressurization X X-45 . Containment Atmospheric Control from X Suppression Chamber, and Drywell Floor Seal Pressurization X X-46 Containment Atmospheric Control from X Drywell X-47 Containment Atmospheric Control from X Drywell CRD Insert and Withdraw Lines X
-XS-S HPCI Steam Supply to RllR lleat Exchanger, X RllR lleat Exchanger Vent, and X RIIR llent Exchanger (RV) X XS-6 Suppression Pool Cleanup / Pump Down X XS-7 Containment Atmospheric Control to , X Suppression Chamber ,
XS-8 Containment Atmospheric Control to X Suppression Chamber XS-ll Drywell Service Air , X XS-12 Containment,Drywell Radiation X Monitoring Subsystem XS-13 Containment Drywell Radiation X Monitoring Subsystem g e O
PE O fY CON h .JENT PENETRATION NUMBER SYSTEM ESSENTIAL NON-ESSENTIAL XS-20 Containment Atmospheric Control to X Drywell XS-21 Containment Atmospheric Control to X Drywell XS-22 Containment Vent to RBNVS X B-7 Instrument Air to Drywell X(1) i D-5 Instrument Air to Drywell X(l) F-10 Recire. Pump Seal Injection X '
- F-ll Recire. Pump Seal Injection i
i I i i I W Even though instrumentation air is non-essential, its supply to the drywell is
- desirable. Hence it is not isolated. Check valves provide isolation in the event of loss of instrument air, l
l i i e
i SNPS-
'i % TABLE FROCESS PIPELINES l
(Numbers in parentheses are keyed to notes on p Valve Location Primary Number Valves Nominal Relative to Containment of Per Pipe Size Primary Penetrations Lines Isolated (22) g Lines Line (In.) Containment X-1A,B,C D Main Steam 55 4 1 24 Inside 1 24 Outside Main Steam Line Drain and MSIV- 55 4 1 2 Outside Leakage Control Systm 55 4 1 2 1/2 Outside X-2A Feedwater 55 1 1 18 Inside 1 18 Outside I-2B Feedwater 55 1 1 18 Inside 1 18 Outside X-3 Main Steam Line Drain 55 1 1 3 Inside 1 3 Outside X4 RWCU Line frce RPV 55 1 1 6 Inside 1 6 Outside X-5 RHR Shutdown Cooling from RPV 55 1 1 20 Inside 1 20 Outside
- 4,B
, RHR Injection Line to Recircul- 55 2 1 24 Inside ation Syst e Return 1 2 Inside 1 24 Outside X-7A,B RHR - Containment Spray Drywell % 2 1 10 Outside 1 10 Outside X-8A,B rim - Containment Spray suppression % 2 1 6 Outside Chamber 1 16 Outside 1 16 Outside X-9A,B,C,D RHR Pump Suction % k 1 20 Outside X-10A rim Test Line Return % 1 1 16 Outside to Suppression Chamber, Suppression Pool Cleanup Return, 1 2 6 Outside R!m Steam Condensing Discharge, 1 1 4 Outside RHR Minimum Flow, 1 1 4 Outside Core Spray Test Line, and 1 1 10 Outside Core Spray Minimum Flow 1 1 3 Outside
( i O
1 FSAR l 6-2 1 1 ( l /ING MtIMARY COfiTADUTrf l l ws 7 and 8; signal codes are listed on page 8.) Valvaand/or Oper; tor Power Power Closing Normal Type to Open to Close Icolr.tlon Time (See) Status ( 6.99 ) . J5,6) (5,6) Signal (10) _ (8,9) Remarks A0 Globe Air /AC/DC Air / Spring B,C,D,E,P,R,T,RM 3-5 Open (1) A3 Globa Air /AC/DC Air / Spring B,C,D,E,P,R,T,RM 3-5 Open (1) FO Globe AC AC B,C,D,E,P,R,T,RM h Open MO Globe AC AC RM 6 Closed (19) Check Flow Reverse Flcr Reverse Fluw N/A Open VTC Flow ReverseFlow/F,G,RM N/A Open (11) R p g Flow /AC/ Check Flow Reverse Flow Reverse Flow N/A Open VTC Flow ReverseFlow/AC/ Reverse Flow /F,G,RM N/A Open (11) Spring FO Gata AC AC B ,C ,D ,E, P,R ,T,M 16 Open FO Gata DC DC B,C,D,E,P,R,T,RM 16 Open bD G:t:3 AC AC A,J.RM 30 Open
>0 Cat? DC DC A,J,W,Y,RM 30 Cm FO Gat] AC AC A,F,U,RM 23 Closed M0 Gata DC DC A,F,U,RM 23 Closed VTC Flow Reverse Flow Reverse Flow N/A Closed (3)
MO Gate AC AC A,RM 19 Closed 60 Gat 3 AC AC RM 24 Closed (12) MO Gati AC AC F,G,RM 51 Closed (2) FO Anglo AC AC F,G,RM 10 Closed (2) FO Globe AC AC F,G,RM 71 Closed (2) 60 Gat] AC AC F,G,RM 71 Closed (2) MO Globe AC AC F.G,RM 79 Closed (2) MD Gat] AC AC RM 106 Open (13) FO Globe AC AC F,G,RM 79 Closed (2) 50 Cat] AC AC A,F,RM 31 Closed 60 Gata AC AC F,G,RM 20 Closed AC RM 20 Open (16) bD Gata AC MD Glob 3 AC AC F,G,RM 67 Closed FO Gat] AC AC RM 16 Open (16) F 6.2.41 Revision 16 - April 1979 1 of 8 1, i
i 4 TALLE Valve L< Primery Numh r Valves Nominal Relat Containment of Per Pipe Size Prii Penetrations Lines Isolated (22) GDC Lines Line (In.) Conta. X -lOF RHR Test Line Return to 50 1 1 lo Out, Cuppression Chamber, RCIC Minimum Flow,. 1 1 2 Out. HPCI Minimum Flow, 1 1 4 Out, RHR Cteam Condensi v Discharge, 1 1 4 Out: FHR Minimum Flow, 1 1 h Out, Core Opray Test Line, 1 1 10 Out, Core Cprny Minimum Flow, and 1 3 Outi Eelief Valve Discharge from RHR 1 1 2 Out; fupply to RCIC Pump Suction X-ll RHR - head Cpray Line to RPV 55 1 1 4 Ins. 1 4 Out: X-12 HPCI Turbine Steam Inlet Line 55 1 1 10 Ins: 1 1 Ins: 1 10 Out: 1 1 Out X-13 HPCI Turbine Exhaust 50 1 1 18 Out: 2 18 Out: X-14 Epare - - - - - X-15 HPCI Pump Uuction 56 1 1 16 Out X-lo RCIC Turbine Sterc Inlet Line 55 1 1 3 Ins. 1 1 Ins: 1 3 Out. 1 .1 Out. X-17 RCIC Turbine Exhaust 56 1 1 8 Out 2 8 Out; X-18 RCIC Vacuum Pump Discharge 50 1 1 2 Out: 1 2 Out: X-19 ECIC Purp Suction 5u 1 1 o Outt 1 a s
3
'L S-1 FOAR e
2 4-1 (CONT'D) xntion Valveand/or .v2 to Operator Power Power Closing Normal rary Type to Open to Close Isolation Time (Lec) Ctatus .nment (u,22) ($,6) ($,c) Signni (10) (8,9) Remarks
- id 2 MO Globe AC AC F G.E; Closed 79 (2)
Lide 10 Glote DC DC RI: 18 Closed (16) tida NO Globe DC DC RM 2r) Closed (16)
- ide MO Gate AC AC F,0,RM 20 Closed sid) 1.D Gate AC AC RM 20 Open (10)
- 1de MO Globe AC AC F,G .PJ4 07 Closed ilde 10 Gate AC AC bl lu Open (15) ti o Relief Valve High Differ- Cpring U/A N/A Closed ential Pressure
.de 10 Gate AC AC A , F , U, PJ4 20 Closed
- id a 10 Globe DC DC A , F , U, RM 13 Closed de 10 Gate AC AC K,RA 11 Open (7)
.de M0 Globe AC AC K,RM 12 Open (7)
- ide MO Gate DC DC K,Rn 43 Closed (7)
.id2 MO Globe DC DC K,M1 12 Open (7) ide FD Gate DC DC PR 102 Open id2 Check Flow Reverse Flow Reverse Flow N/A Closed (15) idi 10 Gate DC DC X ,M; 71 Closed .Da M0 Gate AC AC K,RM lo Open (7) fa 10 Glote AC AC K,RM 12 Open (7) ide MO Cate DC DC K,RM lb Closed (7) hdi FO Globe DC DC K,El 12 Open (7) 6da c0 Gate DC DC RM 38 Open de Check Flow Reverse Flow Reverse Flow N/A Closed (1.5) Bd> FO f top Check Flow /DC Rev. Flcw/DC Rev. Flow /hM 13 Closed (13,21) Zde Check Flow Reverse Flow Reverse Flow N/A Closed Ad2 10 Gate DC DC RM 31 Closed c.2.4-1 Revicion 9 - Decer.ber 1977 2 cf 8 , J
e 8'
}
TABLE 6. Valve Iocatice Primary rember Vr.1ves W aal Relative to Contairanent of Per Pipe Size Primary Penetrations Lines Isolated (22) g Lines Line (In.) Containment X-20A,B Core Spray Ptamp Discharge to RPV 55 2 1 10 Inside 1 2 Inside ', 1 10 Outside X-21A,B Core Spray Pump suction 56 2 1 14 outside i X-22A,B RBCIrw to Recire. Ptarp and 57 2 1 4 outside < Motor coolers 4
- X-23A,B RBCLCW frcan Recire. Ptamp and 57 2 1 4 outside
, Motor Coolers j X-2ld to H RBCLCW to Drywell Unit Coolers % 8 1 3 Inside 1 3 outside X-25A,B y from Drywell Unit Coolers 56 2 1 4 Inside 1 4 Outside X-26 Purge Air to,Dryvell 56 1 1 18 Inside s 1 18 Outside f X-27 Purge Air from Drywell :,t 1 1 18 Inside 1 28 Outside X.28 Purge Air to Suppression Chamber 56 1 2 18 Outside X-29 Purge Air from Suppression Chamber 56 1 2 18 outside X-30 Sample Coolrat frcan RPV 55 1 1 3 Inside 1 3 Outside X-31 Equipment Drains frcan Drywell 56 1 2 3 Outside X-32 Floor Drains from Drywell 56 1 2 4 outside X-33 Spare - - - - -
- X-34 Spare - - - - -
X-35 Spare - - - - - X-36 Standby Liquid Coolant to RPV 55 1 1 1 Irmide 1 1 Outside 2 1 Outside L l l l
,.s--+ a m - 1 b ,__w i
l rPS-1 FSAR 2.4.1 (CONT'D)
; Valvaand/or Operator Power Pcwer Closing Normal Type to Open to Close Isolation Vime (Sec) Status (6.P2) _($ dL J ,6) signal (10) (8,9) Remarks VTC Flow Reverse Flow Reverse Flow N(A Closed (3) 10 Globe AC AC RM Jo Closed ID Gate AC AC RM 43 Closed (18)
W Gate AC AC RM 76 Open l MD Cate AC AC RM 23 open M0 Gate AC AC RM 23 open Check Flow Reverse Flow Reverse Flow N/A Oper 10 Gate AC AC F,0,Z,RM 16 Open 10 Gete AC AC F,G ,Z ,RM 20 Open MO Gate AC AC F,G,Z,RM 20 Open A0 Butterfly AC/ Air Spring L,RM 5 Closed (17) A0 htterf3,y AC/ Air Spring L,RM 5 Closed (17) A0 Butterfly A Spring L,EM S Closed (17) A0 Butterfly AC Spring L,RM 5 Closed (17) A0 Butterfly AC/ Air Spring L,RM 5 Closed (17) AO Butterfly AC/ Air Spring L,RM 5 Closed (17) A0 Globe AC ir Spring B,C,RM 15 open A0 Globe A :':pring B,C,RM 15 Open 10 Gate AC AC A,F,RM 16 Open 10 Gate AC AC A,F,RM 16 Open (15) (15) (15) Check Flow Reverse Flow Reverse Flow N Closed Check Flow Ro arse Flow Reverse Flow N Closed Deplosive AC N/A RM Instantaneous Closed 6.2.4 1 Revision 16 - April 1979 3 of 8 i l
p 4 k SNP: TABLE 6.2, Valve Location Prir:ary Naber Valves Nominal Relative to Containment of Per Pipe Size Primary Penetrations Lines Isolated (22) g Lines Line (In.) Containment X-37A Nitrocen/AirPurgeforTIP 57 1 1 3/8 outside X-37D,C,D TIP Drive Guide Tubes 57 3 1 3/8 oute*4e 1 3/8 Outude X-38 TIP Drive cuide Tubes 57 1 1 3/8 outside 1 3/8 Outside X-39A B Instrument Air to Suppression 56 2 1 1 Outside Chamber 1 1 Outside X h0 Spare - - - - - X 41 HPCI Vacuum Breaker 56 1 1 2 Outside 2 18 Outside X h2 RCIC Vacuu:n Breaker 56 1 1 11/2 Outside 2 8 Outside X 43 RHR Relief Valve Discharge Vacuum 56 1 N/A 5/A N/A Breaker, rim Heat Exchanger Vent, 2 2 1 Outside R!m Heat Exchanger, and 2 1 1 Outside HPCI Steam Supply to RHR Heat Exchanger 2 1 6 Outside X kk Containment Atmospheric Control 56 1 1 6 outside from Suppression Chanber, and 1 4 Outside Drywell Floor Seal Pressurization 57 1 1 1/2 Outside X h5 Containment Atmospheric Control 56 1 1 6 Outside frm Suppression Chamber, and 1 4 Outside Drywell Floor Seal Pressurization 57 1 1 1/2 Outside X k6 Contaiment Atmospheric Control 56 1 1 6 Inside fran Drywell 1 6 Outside X-47 Containment Atmospheric Control 56 1 1 6 Inside from Drywell 1 6 Outside CRD Insert and Withdraw Lines 55 137 1 3/4 outside 137 1 1 outside 1 i
a
'I l.1 FSAR s
4 1 (C0!rt'D) Valvoand/or Operator Power Iower closing Nermal Type to Open to close Isolation Time (Gec) Status (6,22) (5,6) (5,6) Signal (10) @L Renarks cieck Flow Reverse Flow Reverse Flow N/A Open Ball AC Spriig - 05 Closed (1k) Explosive N/A DC RM Instantaneous Open (14) Chear Ball AC Spring - 05 Closed (14) Explosive N/A DC RM Instantaneous Open (14) Shear Check Flow Reverse Flow Reverse Flow N/A Open MO Globe AC AC F G,RM 5 Closed l
. . . . . . (;$)
m Globe DC DC F and X, RN 13 Open Check Flow Reverse Flow Revers'e Flcw N/A Closed MD Globe DC DC F and X, EM 16 y Check Flow Reverse Flow Reverse Flow N/A Close! N/A N/A N/A N/A N/A N/A Mo Globe AC AC RM 10 Closed R: list Valve High Fressure Spring N/A N/A Closed Ralier Valve High Pressure Spring N/A N/A Closed M0 Gats AC AC RM 31 Closed m Gets AC AC RM 20 Closed W Globe AC AC RM 6 Open m Gate AC AC RM 31 Closed W Gets AC AC RM 20 Closed m Globe AC AC RM 6 Open m Gata AC AC RM 31 Closed ' 2 Gata AC AC RM 16 Closed W Gats AC AC RM 31 Closed m Gato AC AC RM 16 Closed Globe Manual Manual N/A / MA Open (20) - Globe Manual Manual N/A N/A Open (f.c) 6.2.4 1 Revision 16 % g
.i
-l l
i I
)
.. - _ _ _ _ . _ .._ - , . . . _ . - - _ - -, ., , , , , . . - - - . _ - . . , m.. _. _ m . _ _ _ -
f f k mLt e g Valve Locatic. Primary :Priber Valves Ncciinal Relative tr Containment of Per Pipe Size Primary Penetrations Lines Isolated (22) GT Lines Line ] In.) Containment XS-1 Spare - - - - - XS-2 Spare - - - - - XS-3 Spare - - - - - XS-4 Spare - - - - - XS-5 !{PCI Steam Supply to RHR IIeat 56 2 1 6 Outside Dtchanger, RJul Heat Dcchanger Vent, and 2 2 1 Outside RHR Heat Dcchanger 2 1 1 Outside XS-6 SuppressionPoolCleanup/PumpDown 56 1 2 10 Outside XS-7 Containment Atmospheric Control to 56 1 2 6 Outside Suppression Chamber XS-8 Containment Atmospheric Control to 56 1 2 6 Outside Suppression Chamber XS-9 Spare - - - - - XS-?O Spare - - - - - XS-ll Drywell Service Air 56 1 1 11/2 Inside 1 11/2 Outside XS-12 Containment Trywell Radiation 56 1 1 11/2 Inside Monitoring Subsyster 1 11/2 Outside XS-13 Containment Drywell Radiation 56 1 1 1 Inside Monitoring Subsyste:1 1 1 Outside XS-14 Spare - - - . _ XS-15 Spare - - - - - XS-16 Spare - - - - XS-17 Spare - - - - - (
a lNPS-1 FCAR 2,L1 (CC'C'D) P o Valveand/cr Cperater Power I cver Clcsirg Ners1
?/pe to Cpet to Clcre a tics 1:c'
(',72) (5, r. ) Ti=e (See) Status ($, ) Firal (101 f,9) Finarr s (ic) (15) (15) (15) R211ef Vilve High Oprire N/A C1cset N/A Pressure 10 Glete A0 A0 F24 10 Cicsed R lier Valve High Sprird I:/A Clcsed N/A Pressv e 10 Gste AC AC A .F,P24 51 cicsed 10 Gate AC A0 F24 32 Clcs ed 10 Gate AC A0 F24 32 Cicsed (15) (15) 01eck Flcv Peverse Flev Reverse Flev N/A C1csed ( L) Gata Ma:r d Manual N/A Iceked N/A (L) 01csed
;-O Globe AC AC F.3,RX lb Cpen 10 Globe AC AC F,3,F24 14 Cpen 10 Globe A0 A0 F,G ,P24 14 10 Globe Cret A0 A0 F,G,F21 14 Cpen
. . . . . . (15)
(15) (15) (15) 6.2.L1 hevisiar.16 . arril 1979 5 cf 8 ! a
~
ii u s:2 l Valve Locatiol Primary Number Valves Nominal Relative to Containment of Per Pipe Size Primary Penetrations Lines Isolated (22) J C Lines Line (In.) Contairs.enO XS-19 Spare - - - - - XS-19 Spare - - - - - X3-20 Contairment Atmospheric Control 56 1 1- 6 A-4 ne to Dr/vell 1 6 Outside XS-21 Contairment Atmospheric Centrol 56 1 1 6 Inside to Dryvell 1 6 Outside XS-22 containment Vent to RIWYS 56 1 1 6 Inside 1 6 Outside 15-23 Spare (Reserved for RPV Internal - - - - - Inspection) XS-216 Spare - - - - - XS-25 Spare - - - - - IS-26 Spare - - - - - XS-27 Spare - - - - - XS-28 Spare - - - - - XS-29 Spare - - - - - IS-30 Spare - - - - - B-7 Instrument Air to Dryvell 56 1 1 11 Inside 1 11 Outside D-5 Instrument Air to Dryvell 56 1 1 112 Inside 1 112 Outside F-10 Recire. Pu=p Seal Injection 55 1 1 3 Inside 1 3 Outside F-ll Recire. Pu::p Seal Injection 55 1 1 3/l+ Inside 1 3/h Outside t i
u 'iPS-1 FSAR , '>.2.ta-1 (COPIT'D) 1 Volve and/or Operator Power Power Closirg Norr.al Type to Open to Close Isolation Sta*.u-(6,22) Time (See) _ (5,61 (5,6) Signal (10) (19)_ .cgrge (13) l (13) l 60 Cate AC AC ILM 32 Closed FO Gate AC E RM 32 Closed FO Gate AC AC RM 32 Closed 50 Gate AC AC RM 32 Closed AO Butterfly AC/ Air Spring L,PM 5 Clored (17) AO Butterfly AC/ Air Spring L,RM 5 Closed (17) l (15) (15) (15) (15) (15) (15) (15) (15) Check Flev Reverse Flow Beverse Flow Open N/A FO Globe AC AC RM 75 Open Check Flow Reverse Flow Reverse Flow N/A Open FO Globe AC AC RM 7.5 Open l Check Flow Reverse Flow Reverse Flow Open N/A Check Flow Reverse Flow Reverse Flow N/A Cpen Check Flow Reverse Flow Reverse Flow N/A Open Check Flow Reverse Flo.- Reverse Flow Open N/A ! 6.2..-l Revision it. - April 1979 6 of 8 f a
p I ! TALLE r These notes are keyed by number to correspond to numters in parentheses. 11. Special air testable check valves designed for remote testing durir
- 1. Main steam isolation valves require that both solenoid pilots mechanical operability of the va]
te deenergized to close valves. Accumulator air pressure plus will cauze only a partial moveme: spring set together close valves when both pilots are de- with only a minor effect on flow, energized. Voltage failure at only one pilot will not cause the actuator spring force will e2 valve closure. The valves are set to fully cloa" .n less than when the feedwater system is ava: 5 sec. providing a positive closure diff when the feedwater flow is not as 2 Containment spray to dryvell and suppression chamter and RHR test line return to suppression chamter isolation valves will 12. This valve will open when toth a have the capability to be manually reopened after automatic and an accident af gnal are preser cloadre. ".his setup will permit containment spray, for high drywell pressure conditions and/or suppression water cooling. 13. The motor operatur of this valre When autonatic signals are not present,these valves may be operating conditions. opened for test or operating convenience. 14 Traversing In-Core Probe (T 9 ) Sy
- 3. Testable check valves are designed for remote opening with zero differential pressure across the valve seat. The valves will h' hen the TIP system cable is inse close on reverse flow even though the test switches may be tute opens automatically so that positioned for open. The valves will open when pump discharge A maximum of four valves may be o pressure exceeds reactor pressure even though the test switch the calibration, and any one guid may te positioned for close, per year,
- b. This line is only needed during maintenance. Service air supply If closure of the line is require is disconnected during plant operation by administrative control, by a containment isolation signal retracted and the ball valve clos
- 5. AC motor operated valves required for isolation functions are of cable withdrawal. To ensure f powered from the emergency AC power buses. DC operated isolation cable fails to withdraw or a ball valves are powered from the station batteries, shear valve is installed in each manual si 6nal, this explosive val 6 All motor operate isolation valves will remain in the last seal the guide tube, position upon fai;ure of valve power. All air-operated isolation valves will close upon air failure. 15. All unused penetrations (designat.
welded.
- 7. Signal B opens, signal K overrides to close.
- 16. Valve will close on system high f.
8 Power operated valve can be opened or closed by remote manual switch for operating convenience during any mode of reactor operation except 17. Isolation signals A or F will ini' when automatic signal is present (see Note 2). ventilation system which in turn valves.
- 9. Normal status position of valve (open or closed) is the position during normal power operation of the reactor. 18. This valve will open when toth a the valve and an accident signal a
- 10. The specified closure rates are as required for containment isolation only.
k I
4 NPS-1 FSAR _2.k-1 (CONT'D) NOTES with c positive closing feature are 19. Pressure sensors, sensing steam line pressure are used for interlock
; normal operation to assure va di;c. The remote testing feature control to prevent inadvertent valve opening at high steam line pressures (above 35 p818 ).
t of tha disc, into the flow stream, Up on riclipt of an isolation signal, 20. Control Rod Drive (CRD) Insert and Withdraw Lines th;r c2nla a slight reduction in flow labl2 or cluse the valve to close, Criteria 55 concerns those lines of the reactor coolant pressure boundary
.renti 1 pressure on the *a?ed disc, penetrating the primary reactor containment. The CRD insert and with- ;ilntl2.
draw linea are not part of the reactor coolant pressure boundary. The classifica\ ion of the insert and withdraw lines is Quality Croup 2, and low reactor pressure vessel pressure therefore, designed in accordance with ASME Section III, Class 2. The t. basis to which the CRD lines are designed is commensurate with the safety Lmportance of isolating these lines. Since these lines are vital to the L1 krylock:d open during normal scram function, their operability is of utmost concern. stem In the design of this system, it has been accepted practice to omit automatic valves for isolation purposes as this introduces a possible rted, tha ball valve of the selected failure mechanism. As a means of providing positive actuation, manua} shutoff valves are used. In the event of a break on these lines, the th; pr;be end cable may advance, manual valves may be closed to ensure isolation. In addition, a ball pen:d Ct cny one tine to conduct check valve located in the insert line inside the CRD is designed to e tube is used, at most, a few hours automatically seal this line in the event of a treak.
- 21. This hD stop check valve is normally in a closed position due to its i during c;libration, as indicated check valve feature, but its hD is in the open position. The MO th) c:,bla is automatically provide
- a backup to close the valve to provide additional high n; tutomatically after completion leak tight integrity, s:lation czpability, if a TIP valve f;ils to close, an explosive, Lini. Upon receipt of a remote ve will thear the TIP cable and ed "Sparo") are capped and seal
- Low,
.i:;ts th2 reactor building standby c lat a th7 purge air isolation .ow diffar;ntial pressure across tre pr sint. C.2.4-1 Revision 16 - April 1979 7 of 8 P L d
F 9 S: l TALLE 6,
)
N(n'ES (CONT'D) ISOLATIC 22 Atbreviations used in table: Signal AO - Air operated A' Reactor vessel low water MO - Motor operated E' Reactor vessel los water at this level, and recirc VTC - Pneumatic testable check valve C* High radiation main ste RHR - Residual Heat Removal System D' Line break main steam 3 RPV - Reactor Pressure Vessel E' Line break - rain steam 1 RCIC - Reactor Core Isolation Cooling System F' High drywell pressure RWCU - Reactor Water Cleanup G Reactor vessel low water HPCI - High Pressure Coolant Injection level) GDC - General Design Criterion J' Line treak in reactor wat RECILN - Reactor Building Closed Loop Cooling Water K+ Line break in steam line diaphragm pressure) TIP - Transversing Incore Probe CRD - Control Rod Drive L Reactor building standby fCIV - Main Steam Isolation Valve M High radiation signal dow P' Low main steam line press R Low condenser vacuum T High temperature in Turbi U High reactor vessel press W' High temperature at outle X Low steam pressure Y Standby liquid control sy Z Low level in RBCIC4 head RM* Remote manual switch from These are the isolation f other functions are give i MI
9 PD-1 FfAR 2El,,(CONT'D) o N SIGNAL CODES Description Lev:13 - (A scram will occur at this level) l L:v:12 - (The reactor core isolation cooling system and the high pressure coolant injection system will be initiated ulation pumps are tripped) na line Lne (high steam flow) ine (stsam line tunnel high temperature) L val 1 - (The core spray systems and the low pressare core injection mode of RHR systems will be initiated at this er cloup system - high space temperature, high differential flow, high differential temperature Lo/from turbine (h!gh steam line space temperature, high steam flow, low steam line pressure or high turbine exhaust rintilation system initiation astream cf primary containment purge filter train l es at inlat to turbine (RUN mode only) w Building are L cf elecnup system nonregenerative heat exchanger item ectuated Lank main control room inctions of the primary containment and reactor vessel isolation control system; i fcr information only. c.2.4-1 Revision 9 - Decenber 1977 8 of 8 I
~
V CC2 CCI XII RHR-HEAD SPR AY LINE TO RPV b' ~
; b &
X2 A FEEDWATER LINE : :: W
- ~~
X2B FEEDWATER LINE >f--* CCt_ CC1 X20A CORE SPRAY PUMP DISCHARGE TO RPV
- E ' -- b -
CC2 _CC1 X208 CORE'SPR AY PUMP DISCHARGE TORPV
- 8 --
b - CC2 CC1 4-('
- a X36 STANDBY LIQUID COOL ANT TO RPV CC4 CC:
O X30 SAMPLE COOLANT FROM RPV < :: : LJ C 2 g X6A RHR INJECTION LINE TO _ 8 1 -- 2, _
' " ~
RECIRCULATION SYSTEM RETURN LINE X4 RWCU LINE FROM RPV : CC5 kCCI
- $ =
CC4 CC3 Fil RECIRCULATION PUMP _
~
h
~ "
SEAL INJECTION WATER E LEVEL \ f\ - k.
I I I l J 9 f i i I I I l E f I i I 1 0 l I l I l l { J i i l O I 4 e-,----.- v., , -.--- -
,% , - ---- . . - - -- - --- - - v
cci_ , ce r cci
^~
ces y MSIV-LE AKAGE CONTROL S"
- (TYPICAL ARRANGEMENT FOR
". tva. o r r cci ece 4 OUTBOARD MSIV'S) r ;: b : XID MAIN STE A l C_C L.lC 5 E : M AIN STE AM Ll J tr a or 4 (C'- CC' -
r $ = XIC MAIN STEAM l cci ces
@ $ r MAIN STEAM LINq gP; cei, ccr r : : XIB M AIN STEAM CC' C C3 h $ = M AIN STEAM LINL:
O' 'g cci cce
' J},p g cci cc4 r -- -
Xi A M AIN STE AM j TO 1111 __ h, $ r MAIN STEAM LINI SUPPRESSION 6d ce, cc. POOL y . T -~p : X3 MAIN STE AM CC1 CCr REACTOR cci ec.
, / VESSEL -- E z X12 HPCI TURBl?
~'
j cci
=~
cce INLET LINE _c al _ X16 RCIC TURBIN? eci ' ec, INLET LINE l
~ -
hg _ 8
, X68 RHR INJECT RECIRCUL c7 Tr
> f " n i g c e s. _c c e K - g,
.. I E X5 RHR SHUTDO
~
COOLING LINE F
~
*- F lO RECIRCUL AT REACTOR --+<
SEAL INJECTIO, RECIRCULATION PUMPS ccr cc4 l l
~~
- xxt-- CRD INSERT AND WITHDRAW LINES O 3 04-- (T YP. FOR 137 UNITS)
DRYWELL UPPRESSION CHAMBER AIR SPACE ~~~~ SUPPRESSION POOL _-
LEGEND PSTEM Ded-GLOBE VALVE (CLOSED) W-GLOBE VALVE (OPEN)
<3-GATE VALVE (OPEN) HPCI-HIGH PRESSURE I DRAIN COOLANT INJECTION
>4- G ATE VALVE (CLOSED )
RCIC-REACTOR CORE i CN - CHECK VALVE ( O PE N ) ISOLATION COOLING Di- CHECK VALVE (CLOSED) RHR-RESIDUAL HEAT REMOVAL R-MOTOR OPERATOR R PV-REACTOR PRESSURE VESSEL
- IR OPER ATOR CRD-CONTROL ROD DRIVE
' DR AIN
- SAFETY / RELIEF VALVE RWCU- REACTOR WATER CLEANUP PUMP MSIV- M AIN STE AM DRAIN M - EXPLOSlVE VALVE INE DR AIN b-LEAK TEST CONNECTION (L.T.C.)
;E STEAM NOTES
- 1. M AIN STEAM PIPING UP TO ISOLATION VALVES WAS PURCH ASED TO B31.1 AND ANALYZED TO ASME III (CODE CLASS 1)(CCI).
)ON LINE TO
- 2. ALL L.T.C. VALVES (EITHER GATE OR GLOBE)
{ STEM RETURN LINE AND LINE SIZES ARE 3/4 INCH, ASME III CC2 AND H AVE AT LEAST ONE OF THE TWO VALVES IN SERIES LOCKED CLOSED. h NOM RPV JON PUMP 3 F IG. 6.2.4 -1 CRITERION 55 CONTAINMENT ISOLATION VALVES SHOREHAM NUCLEAR POWER STATION-UNIT 1 FINAL SAFETY AN ALYSIS REPORT REVISION 9 - DECEMBER 1977
X258-RBCLCW-FROM DRYWELL UNIT [mh LEGEND D MS - GLOBE VALVE (CLOSED) tx4 - GLOBE VALVE (OPEN) RBCLCW-TO DRYWELL UNIT COOLERS 94 - G ATE V1LVE (CLOSED) r 4 - GATE VALVE (OPEN) g - ANGLE VALVE (CLOSED) F' - C H E Cit VALVE ! CLOSED) W - CHECK VALVE (OPEN) XS22-CONTAINMENT VENT 1 h40 - BLTTERFLY VALVE (CLOSED) d - RELIEF VALVE XTA RHR-CONTAINMENT SPRd X4TCONTAINMENT ATMOSPHE-I X-MOTOR OPERATOR 1- AIR OPERATOR XS13 CONTAINMENT DP h -LE AK TEST CONNECTION (L.T C.) RADIATION MONITORINT HPCI HIGH PRESSURE COOL ANT INJECTION XS11 DRYWE RCIC-RE ACTOR CORE ISOL ATION COOLING RHR-RESIDUAL HE AT REMOVAL X27 PURGE AIR FROM @
@ STR AINER v43 CONTAINMENT ATMOSPHERIC *
~ " "
XS8 CONTAINMENTATMOSPHERIC o
-4 L - FLANGE CONNECTION
- 4. RESTRICTING ORIFICE X32 FLOOR DRAINS FROM DR' rov W- AC OR LDING CLOSED LOOP X29 PURGE AIR FROM SUPPRESSION I , X8A RHR-CONTAINM LC-LOCKED CLOSED ' SUPPRESSION CHAMB XS-6 SUPPRESSION POOL CLEANUP / PUMPDOWN X18 RCIC Vf CUUM PUMP DISC SUPPRESSION POOL _E A ',
CLEANUP RETURN RHR STE AM CON- _ A DENSING DISCHARGE w : X42 RCIC VACUUM B0!! RHR MINIMUM FLOW -+ d a X17 RCIC + TURBINE EXH AUST XIO A RHR TEST LINE RETURN r E E I TO SUPPRESSION CH AMBER I X2iA CORE SPRAY PUMP CORE SPRAY TEST LINE 5 X19 RC;i PUMP
~*- X9ARHR PUM O) CORE SPRAY MINIMUM FLOW r E X9C RHR PUK
i
? 7
@OOLERS IW
, h M* -*- E w M-4 5 - X25 A-RBCLCW - FROM DRYWELL UNIT COOLI ~~
X2CE L st M r EL-4 + X24A~ l p X24F -*--i l. W @~ I~ ~ Wl 19 l
. blW l X248
> RBCLCW-TO DRYWELL UNIT COO 3 l l
) X24G iM y >*--*-e f 5-4 + X24C X24H ' 1 k4 ; -- & I X24D _
\ ( REACTOR VESSEL f Ig@g
[ Q N(I k b n Y DRYr! ELL " , ,
, , , - X78 RHR-CONTAINMENT SPR AY DRYWELL l
}IC CONTROL : H ^Mc M4 H r X46 CONTAlnMENT ATMOSPHERIC CONTROL IC CONTROL $ N-* = $ 5 c XS21 CONTAINMENT AT MOSPHERIC CONTR OL W! ELL . SUBSYSTEM
~- M L g1y -
XS12 CONTAINMENT DRYWELL RADIATION MONITORING SUBSYSTEM l T l
-*-+c:g- E : B-7 INSTRUMENTAIR TODRYWELL LL SERVICE AIR y 'r y DRYWELL
+-+c: E, c D-5 INSTRUMENT AIR TODRYWELL l j
+y m s-f fYOELL y ? : X26 PURGE AIR TO DRYWELL ONTT,OL --*-- N- N : r
- 55m X44 CONTAINMENT ATMOSPHERIC CONTROL ONTROL -*--5
- U ce4 N j[ .
g _ X31 EOUIPMENT DRAINS
}No FROM DRYWELL
/ CELL $ : XST CONTAINMENT ATMOSDHERIC CONTROL
- AMBER *- M.
SUPPRESSION CH AMBER
- AIR m {S=%.p ., c X28 PURGE AIR TO SUPPRESSION CHAMBO 7 SPRAY ',
.c c .
8 1 - "
- SPACE '
} 8 X8B RHR-CONTAINMENT SPRAY.
T ,I 1 SUPPRESSION CHAMBER
'C'
%_g .1 I I INX43 RHR-RELIEF VALVE DISCHARGE VOCUUM BRE AK7
'* YS5 HPCI STC
- DR A l(
L.O SRGE = :: 5 - E _ X 39 A INSTRUMENT AIR TO 7 9 cce SUPPRESSION CHAMBER X398 INSTRUMENT AIR TO SUPPRESSION CHAMBER AKER --
- --fk - -
l E'. X41 HPCI VACUUM BREAKER 1(b T
,_1_ g yN".G Qj . m m
X13 HPCI _ TURBINE EXHAUST L WATER l XIOA . LEVEL y XIOB
= M b=_ *1 X 218 CORE SPR AY SUCTION -* M w
_ y 7 A, ___ PUMP SUCTION , E-: c SUCTION --*-H M r X15 HPCI PUMP SUCTION .. % SUCTION -* --M - a d E X98 RHR PUMP SUCTION f - SUCTION - *--5 d* U V a M X90 RHR PUMP SUCTION N ' ggg g g gg FROM RHR SUPPLY =
-* S C T SUPPRESSION POOL i
ERS NOTES
- 1. THESE ARE ALL ASME 22 CODE CL ASS 2 (CC2)
SYSTEMS UNLESS OTHERwlSE NOTED.
- 2. ALL L.T.C. VALVES (EITHER G ATE CR GLCBE)
ERS AND LINE SIZES ARE 3/41NCH, ASWE ". CC2 AND H AVE AT LE AST ONE LOCKED CLOSE D VALVE. NN c RHR HE AT EXCHANGER VENT 3 RHR HEAT EXCHANGER E3 _ XS5 HPCI SUPPLY TO RHR HEAT EXCHANGER ' - RHR HEAT EXCHANGER VENT 4 H ( RHR HE AT EXCHANGER
? HPCI SUPPLY TO RHR HEAT EXCHANGER ; AM ,
1 M-- RCIC MINIMUM FLOW
- L
---- M H PCI MINtMUM FLOW Nw- RHR STEAM COND: NSING D!SCH ARGE
- H-
- RHR MINIMUM FLOW
~
_ X108 RHR TEST LINE RETURN h TO SUPPRESSION CHAMBER
,_A- 0 i CORE SPRAY FIG. 6.2.4-2 TEST LINE CRITERlON 56 CONT AINMENT COGE SPRAY ISOL ATION VALVES MINIMUM FLOW "HOREHAM NUCLE AR POWER STATION - UNIT 1 ; PUMP 'INAL SAFETY ANALYSIS REPORT N LINE R E VISICN 16- A p e n t 1979,
I I ! l
}
{ REACTOR' PRESSURE
- O CS cc X22 A-RBCLCW-TO RECIRCUL ATION PUMP _
' Q l AND MOTOR COOLERS T ces -
CC3 X23A-RBCLCW-FROM RECIRCULATION PUMP Q di AND MOTOR COOLERS M X37A NITROGEN / AIR PURGE FOR TIP X378 TIP DRIVE GUIDE TUBES : b X37C TlP DRIVE GUIDE TUBES = b X370 TIP DRIVE GUIDE TUBES = r X38 TIP DRIVE GU!DE TUBES = = b
-8 St X45-DRYWELL FLOOR SEAL PRESSURIZATION E ::
f WAT EF O
T n ~ t I VESSEL N .s h cc3 a - - s T T l I Q _ X 22 8 -RBCLCW- TO RECIRCUL
' ' ~
AND MOTOR COOLERS m c,3 x s> 1 g _ X23B- RBCLCW- FROM RECIRC AND MOTOR COOLERS REACTOR RECIRCULATION PUMPS DRYWELL JPPRESSION CHAMBER tLEVEL AIR SPACE :: E : X44-DRYWELL FLOOR SE AL SUPPRESSION POOL i
LEGEND TIP-TRAVERSING INCORE PROBE Cod - GLOBE VALVE (OPEN) - BALL VALVE (CLOSED) pse - GLOBE VALVE (CLOSED) -EXPLOSlVE VALVE
>4 - GATE VALVE (OPEN) h - LEAK TEST CONNECTION
( L.T. C.)
>4 - G ATE VALVE (CLOSED)
JR - MOTO R OPER ATOR RBCLCW-REACTOR BUILDING CLOSED LOOP COOLING WATER
-PUMP
- - - DRYWELL FLOOR SE AL NOTES:
- 1. ALL PENETRATIONS, PlPING ANDISOLATION VALVES ARE ASME E CODE CLASS 2 (CC2)
- 2. ALL L.T.C. VALVES (EITHER GATE OR GLOBE) AND LINE SIZES ARE 3/4 INCH, ASME E CC2 AND H AVE AT LEAST ONE LOCKED CLOSED VALVE.
T GTION PUMP .MLATION PUMP
~
PRESSURIZATION CRITERION 57 CONTAINMENT ISOL ATION VALVES SHOrCH AM NUCLEAR POWER STATION-UNIT 1 FIN AL SAFETY AN ALYSIS REPORT REVISION 9-DECE MBER 1977
SNPS-1 RESPONSE TO NUREG 0578 m k ,,) 2.1.5.a Dedicated Penetrations for External Recombiners or Post Accident Purge Systems NUREG 0578 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 recom-biner or purge systems that are dedicated to chat 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 aresized to satisfy the flow requirements of the recombiner or purge system. NRC CLARIFICATION:
- 1. This requirement is only applicable to those plants whose licensing basis includes requirements for external recombiners or purge systems for pcst-accident combustible gas control of the containment atmosphere.
- 2. An acceptable alternative to the dedicated penetration is a
(} ( j combined design that is single-failure proof for containment isolation purposes and single-failure proof for operation of i the recombiner or purge system.
- 3. The dedicated penetration or the combined single-failure proof alternative should be sized such that the flow requirements for the use of the recombiner or purge system are satisfied.
- 4. Components necessitated by this requirement should be safety grade.
- 5. A description of required design changes and a schedule for accomplishing these changes should be provided by January 1, 1980. Design changes should be completed by January 1, 1981.
BWR OWNERS' GROUP DISCUSSION: None BWR OWNERS' GROUP IMPLELENTATION CRITERIA: None
\
J ^ 2.1.5.a-1
SNPS-1 RESPONSE TO NUREG 0578 LILCO'S RESPONSE: The Shoreham design presently incorporates redundant external recombiners for the control of combustible gases inside the primary containment. Two 100 percent capacity hydrogen recombiners are currently in-stalled. The system is safety-related, Seismic Category I and designed in accordance with ASME III, Code Class 2. The re-combiners are located in the reactor building outside the pri mar containment. Four dedicated penetrations are provided f eac recombiner as shown in the enclosed FSAR Figure 6.2.5-1. b Two isolation valves are provided for each primary containment penetration in accordance with the redundancy and single failure . requirements of General Design Criteria 54 and 56 of Appendix A to 10 CFR 50. The combustible gas control system is considered non-essential. However, automatic isolation is not provided, since all isolation valves in this system are closed during normal operation. Ov W This Figure is provided for information only. The hydrogen analyzers have been provided with dedicated penetrations and are no longer connected to the hydrogen recombiner pene-trations as shown in this Figure. Thus, the hydrogen recom-O biners are provided with ded2.cated penetrations. This Figure will be updated in a future amendment to the FSAR. 2.1.5.a-2
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- iia 4se fie- 2 9 h*C5 60.
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-h f
s _. 1% [o. ss <s,< f( FIG. 6.2.5-1 y,*n'g,' $ g.q,.,3 .,ser t PRIMARY CONTAINMENT ' N " '* U ATMOSPHERIC CONTROL SYSTEM SHOREHAM NUCLE AR POWER STATION-UNIT 1 b FINAL SAFETY ANALYSIS REPORT REVISION 16- APRIL 1979
SNPS-1 RESPONSE TO NUPIG 0578 2.1.5.b Inerting BWR Containments NUREG 0578 POSITION: It shall be required that the Vermont Yankee and Hatch 2 Mark I ; BWR containments be inerted in a manner similar to other operating BWR plants. Inerting shall also be required for near term OL ; licensing of Mark I and II BWRs. i NRC CLARIFICATION: None BWR OWNERS' GROUP DISCUSSION: None BWR OWNERS' GROUP IMPLEMENTATION CRITERIA: None O LILCO POSITION: In accordance with IRC letter from D. B. Vassallo to 3.11 :pending operating license applicants, dated September 27, 1979, the pro-posed rule making on inerting has been delayed and no action is required at this time. 2.1.5.b-1
l SNPS-1 RESPONSE TO NUREG 0578
/ 2.1.5.c Canability to Install Hydrogen Recombiner at Each Light (sT Water Nuclear Power Plant NUREG 0578 POSITION:
- 1. All licensees of light water reactor plants shall have the capability to obtain and install recombiners in their plants within a few days following an accident if containment access is impaired and if such a system is needed for long-term post-accident combustible gas control.
- 2. The procedures and bases upon which the recombiners would be used on all plants should be the subject of a review by the licensees in considering shielding requirenents and personnel exposure limitations as demonstrated to be necessary in the case of TMI-2.
NRC CLARIFICATION:
- 1. This requirement applies only to those plants that included Hydrogen Recombiners as a design basis for licensing.
- 2. The shielding and associated personnel exposure limitations
['N x associated with recombiner use should be evaluated as part of licensee response to requirement 2.1.6.b, " Design review for Plant Shielding."
- 3. Each licensee should review and upgrade, as necessary, those criteria and procedures dealing with recombiners use. Action taken on this requirement should be submitted by January 1, 1980. -
BWR OWNERS' GROUP DISCUSSION: None BWR OWNERS' GROUP DELEMENTATION CRITERIA: None g-~S 2.1.5.c-1 - N.) ~
. SNPS-1 RESPONSE TO NUREG 0578 LILCO'S RESPONSE:
The Shoreham Design employs permanently installed hydrogen re-combiners. These recombiners are 100 percent redundant, safety grade and have been designed to deal with quantities of hydro-gen that may be generated during and after a LOCA as predicted by ECCS analyses discussed in FSAR Section 6.3.3. Following an accident, each hydrogen recombiner is controlled from the main control room, and no access to the equipment nor local manipulation by plant personnel is required for its operation. In addition, the hydrogen recombiners are located in an area where personnel access for other purposes is not required during accident conditions. Thus,- operation of this equipment will not contribute to the personnel exposures. Refer to response to Requirement 2.1.5.a. No further action is required. v i 2.1.5.c-2
SMPS-1 RESPONSE TO NUREG 0578 2.1.6.a Integrity of Systems Outside Containment Likelv to Contain Radioactive Materials (Engineered Safety Systems and Auxiliary Systems) for PWRs and BWRs NUREG 0578 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-as-practical 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 the NRC.
- 2. Continuing Leak Reduction s Establish and implement a program of preventive maintenance J to reduce leakage to as-low-as-practical levels. This program shall includa. periodic integrated leak tests at a frequency not to execed refueling cycle intervals.
NRC CLARIFICATION: Licensees shall, by January 1, 1980, provide a summary description of their program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident. Examples of such systems are given on page A-26 of NUREG-0578. Other examples include the Reactor Core Isolation Cooling and Reactor Water Cleanup (Letdown function) Systems for BWRs. Include a list of systems which are excluded from this program. Testing of gaseous systems should include helium leak detection or equivalent testing methods. Consider in your program to reduce leakage potentiql release paths due to design and operator deficiencies as discussed in our letter to you regarding North Anna and Related Incidents dated October 17, 1979. BWR OWNERS' GROUP DISCUSSION: None
./
) -
2.1.6.a-1
SNPS-1 RESPONSE TO NUREG 0578 BWR OUNERS'. GROUP IMPLEMENTATION: , Practical leakage reduction measures will be investigated for systems which may contain radioactive fluids post-LOCA. Such systems as the reactor core isolation cooling system, high-pressure coolant injection system, core s. ray system, residual heat removal system, and waste disposal system will be examined. This examination will include a. study of valve. stem packing leakoffs, rotating seals on equipment, gasketed connections or joints, drains piped to open connections, and reactor drainage system. ! Those components in the above systems from uhich leakage may be measured will be identified and measured leakage from these components will be reported to NRC. A periodic leak inspection
- program will be implemented on these components.
LILCO's RESPONSE: A surveillance testing program, in accordance with 10CFR50 Ap-pendix J, " Reactor Containment Leakage Testing for Water Cooled () Power Reactors", and the plant Technical Specifications, will be implemented at Shoreham. This testing program includes per-formance of Type A tests to measure the overall integrated pri-mary containment leakage rates; Type B tests to detect and measure local leakage from certain containment penetrations and components; and Type C tests, to measure containment isolation valve leakage rates. These tests will be performed during pre-operational testing and periodically at test intervals re-quired by 10CFR50 Appendix J. Periodic surveillance testing will be performed on items tuch ! as Main Steam Isolation Valves (MSIV) and Air Locks to main-tain leakage within the allowable limits as specified in the 1 plant's Technical Specifications. In addition, system hydro-static tests, and inspections will be performed as required by ASME Section XI. During these tests appropriate corrective actions will be implemented as required. Additional systems such as the MSIV leakage control system and the primary to secondary containment leakage detection and i leakage return system have been incorporated in the plant design in order to minimize and control leakage to the maximum extent
-possible.
[ v v 2.1.6.a-2
SNPS-1 RESPONSE TO NUREG 0578 The MSIV leakage control system (MSIV-LCS) colleccs post LOCA leakage from the MSIV's to a maximum of 100 standard O' cubic feet per hour for all main steam lines. This system may be manually actuated by the operator 20 minutes after an accident. The MSIV-LCS consists of physically separated redundant blowers which route any leakage from the closed MSIV's to areas served by the Reactor Building Standby Ventilation System (RBSVS). These blowers maintain the steam lines at a pressure slightly below atmospheric thus assuring that any leakage will be directed to the RBSVS filters prior to release to the atmosphere. The primary to secondary containment leakage detection and re-turn system will assist in identifying and controlling post LOCA Emergency Core Cooling Systems (ECCS) leakage. Any , abnormal leakage is detected by a level switch in the EL 8 -0" floor drain sump which will actuate an alarm in the main control room at high sump level. In addition rgdundant safety related level detectors are provided on el. 8,'-0 , which will alarm in the control room when the floor water level (in the de-tector area) exceeds approximately 1/2 inch corresponding to approximately 2,000 gallons. The leakage return portiton of the system consists of a self-priming leakage return pump with a capacity of 180 gpm which include recirculation of 50 gpm. This pump will be manually started as required and will operate to return postulated ECCS leakage to the suppression pool. The
'l pump will be powered from the emergency power supply and will , be seismically qualified. The use of the leakage return system during post LOCA conditions will allow sufficient time for operator action to identify and isolate suspected leakage paths while continuing to maintain suppression pool watey inventory and preventing excessive buildup of water on 01. 8 -0" of the reactor building.
An additional leakage detection program is presently under l evaluation. The program will include measures to reduce and i maintain leakage to as low as practical for systems outside primary containment that could contain highly radioactive fluids during a serious transient or accident. Major features of the l program currently under consideration are as follows: , 1. Preparation of system list, identifying methods that I may be used to test systems, the system involved, I frequency of testing. A preliminary list of the systems affected is prescribed in Table 2.1.6.a-1.
- 2. The preparation of guidelines for evaluating a) leakage from systems, identified in 1 above, into the secondary containment through valve stems, pump seals, fittings, relief valve discharge lines, drains, vents and instru-ment loops and b) leakage through valve seats into inter-facing systems outside of the secondary containment. ._
{G} 2.1.6.a-3 1 _ _ _ ..
1 SNPS-1 RESPONSE TO NUREG 0578
/' 3. The implementation of a periodic visual inspection pro-A Sram. These inspections shall be performed,on accessible portions of applicable systems during system operational l testing, or by evaluation of leakage at lower pressures !
during operation. ' l
- 4. A leak testing program shall be implemented on specific valves or connections on the systems which provide an interface to equipment or systems located external to the secondary containment and which can bypass secondary containment. This testing could be accomplished by hydrostatic leak testing of the individual valves or evaluating the total accumulated leakage during the system ASME XI hydrostatic testing.
- 5. Various methods to detect and control leakage from gaseous systems outside containment shall be evaluated.
- 6. Records shall be maintained on the tests and inspections performed on the system listed on table 2.1.6. a-l.
These records shall be used to identify chronic and generic leakage problems in order to implement mod-ifications and/or corrective mainenance measures. q . L J' Inspection and Enforcement Bulletin 79-21 is currently under evaluation. Appropriate protective measures, as identified therein or modification as applicable to Shoreham, will be implemented as necessary. /O (,/ 2.1.6.a-4 p.- ~ -n- '" %"'~ ' & " ' e ,w
.T **Y
SNPS-1 RESPONSE TO NUREG 0578 TABLE 2.1.6.a-1 PRELIMINARY LISTS OF SYSTEMS TO BE CONSIDERED FOR PERIODIC LEAKAGE INSPECTION AND CONTROL
- 1. Core Spray (CS)
- 2. High Pressure Coolant Injection (HPCI)
- 3. Residual Heat Removal (RHR)
- 4. Reactor Core Isolation Cooling (RCIC)
- 5. Hydrogen Recombiners (combustible Gas Control)
- 6. Reactor Water Cleanup
- 7. Coolant Sampling
- 8. Reactor Building Equipment Drain System
- 9. Reactor Building Floor Drain System
- 10. Reactor Building Standby Ventilation System
SUPS-1 RESPONSE TO NUREG 0578 2.1.6.b Design Review of Plants Shielding and Environmental Qualification of Equipment for Spaces / Systems Which May Be Used in Post-Accident Operations i NUREG 0578 POSITION: i With the assumption of a post-accident release of radioactivity equivalent to that described in Regulatory Guides 1.3 and 1.4 (i.e., the equivalent of 50% of the core radiciodine and 100% of the core noble gas inventory are contained in the primary plant), each licensee shall perform a radiation and shielding design review of the spaces around systems that may, as a result of an ameldent, contain highly radioactive materials. The design review should identify the location of vital areas and equipment, such as the control room, radwaste control stations, emergency power supplies, motor control centers, and instrument areas, in which personnel occupancy may be unduly limited or safety equip-ment may be unduly degraded by the radiation fields during post-accident operations of these systems. Each licensee shall provide for adequate access to vital areas and protection of safety equipment by design changes, increased permanent or temporary shielding, or post-accident procedural controls. The design review shall determine which types of ! corrective actions are needed for vital areas throughout the facility. l NRC CLARIFICATION: i Any area which vill or may require occupancy to permit an operator to aid in the mitigation of or recovery from an accident is desig-nated as a vital area. In order to assure that personnel can perform necessary post-accident operations in the vital areas, we are providing the following guidance to be used by licensees to evaluate the adequacy of radiation protection to the operators:
- 1. Source Term The minimum radioactive source term should be equivalent to the, source terms recommended, in Regulatory Guides 1.3, 1.4, 1.7 and Standard Review Plan 15.6.5. with appropriate decay times based ca plant design.
- a. Liquid Containing Systems: 100% of the core equilibrium noble gas inventory, 50% of the core equilibrium halogen inventory and 1% of all others are assumed to be mixed in the reactor coolant and liquids irjected by HPCI and LPCI.
p .. V 2.1.6.b-1
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SNPS-1 RESPONSE TO NUREG 0578
- b. Gas Containing Systems: 100% of the cere equilibrium noble gas inventory and 25% of the core equilibrium halogen activity are assumed to be mixed in the contain-ment atmosphere. For gas containing lines connected to the primary system (e.g., BWR steam lines) the concentration of radioactivity shall be determined assuming the activity is contained in the gas space in the primary coolant system.
- 2. Dose Rate Criteria The dose rate for personnel in a vital area should be such that the guidelines of GDC 19 should not be exceeded during the course of the accident. GDC 19 limits the dose to an operator to 5 Rem whole body or its equivalent to any part of the body. Uhen determining the dose to an operator, care must be taken to determine the necessary occupancy time in a specific area. For example, areas requiring continuous occupancy will require much lower dose rates than areas where minimal occupancy is required. Therefore, allowable dose rates will be based upon expected occupancy, as well as the radioactive source terms and shielding.
However, in order to provide a general design objective, we are providing the following dose rate criteria with alternatives to be documented on a case-by-case basis. The C,) recommended dose rates are average rates in the area. Local hot spots mav exceed the dose rate guidelines provided occupancy is not required at the location of the hot spot. These doses are design objectives and are not to be used to limit access in the event of an accident.
- a. Areas Requiring Continuous Occupancy: s 15mr/hr. These areas will require full time occupancy during the course of the accident. The Control Room and onsite technical support center are areas where continuous occupancy will be required. The dose rate for these areas is based on the control room occupancy factors contained in SRP 6.4.
- b. Areas Requiring Infrequent Access: GDC 19. These areas may require access on a regular basis, but not continuous occupancy. Shielding should be orovided to allow access at a frequency and duration estimated by the licensee.
The plant Radiochemical / Chemical Analysis Laboratory, l radwaste panel, motor control center, instrumentation l locations, and reactor coolant and containment gas sample stations are examples where occupancy may be l 1 needed often but not continuously. v . 2.1.6.b-2
SNPS-1 RESPONSE TO NUREG 0578 ( ~h t (v/ BWR OWNERS' GROUP DISCUSSION: BWR plants are specifically designed to mitigate major design basis events with no access outside the main control room being required. With this goal in mind, the plants were not specifically designed for any access outside the main control room. To speci-fically design for guaranteed access at any time in most parts of the reactor building is not feasible. However, the current designs may allow for access for short times if the entry time into the area can be selectively chosen. Design changes in shielding will be made if evaluations identify feasible modifications which should significantly enhance desirable access. The guidelines for the evaluations are given below. BWB OWNERS' GROUP IIiPLEMENTATION CRITERIA: A TID 14844 radioactivity release will be assumed into the primary containment. A summation of the radioactivity levels from sump water leakage from process systems in the reactor building will be made. Che next step will be to calculate the source terms for the suppression pool recirculating piping, pumps, and valves
- p. installed in the reactor building assuming that a TID 14844
(' release had occurred. The vital areas will be identified in the reactor building which may need to be entered during an accident recovery period. The shielding in these vital areas will be reevaluated to assess its effectiveness in such a circumstance. The occupancy time limits, taking into consideration transit time, airborne radioactivity levels, and gamma shine intensities, will then be calculated for the vital reactor building areas. LILCO'S RESPONSE: LILCO concurs with the BWR Owners' Group position. A radiation and shielding review is currently being performed for Shoreham to ensure adequate access is provided to vital areas following l an accident. However, based upon the fact that no operator actions other than those which take place in the main control room are critical for the safe shutdown of the plant, only this area, the post-accident sampling station (s), onsite operational support center, and the technical support center, are considered to be vital for continuous post-accident personnel access.b The NRC-prescribed post-accident distribution of radioactivity !and General Design Criteria 19, along with the occupancy time requirements, will be applied to each of the vital areas identified above to assess the dose rate acceptability for plant personnel.
~
1/Studies are presently being conducted by the Owners' Group to
~
varify the NRC-post accident distribution or to develop more realistic / practical assumptions. 2.1.6.b-3 i _ _ _ _ _ _
SNPS-1 RESPONSE TO NUREG 0578 Based on this evaluation, appropriate design changes, such as additional permanent or temporary shielding, and/or post ac-cident procedural controls will be made to cotimize access to ' the vital areas identified. The major part of the Shoreham design assessment is the evaluation of the environmental qualifications of essential equipment. This evaluation will be performed using TID 14844 source terms. O i' 2.1.6.b 4
SNPS-1 RESPONSE TO NUREG 0578 O %J 2.1.7.a Automatic Initiation of the Auxiliary Feedwater System for PWRs NUREG 0578 POSITION: Consistent with satisfying the requirements of General Design Criterion 20 of Appendix A to 10 CFR 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, (s 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 ATUS from the control room.
In the long term, the automatic initiation signals and circuits i shall be upgraded in accordance with safety-grade requirements. l I 2.1.7.a-1 , i l l
SNPS-1 RESPONSE TO NUREG 0578 NRC CLARIFICATION: Control Grade (Short-Term)
- 1. Provide automatic / manual initiation of ATUS
- 2. Testability of the initiating signals and circuits is required.
- 3. Initiating signals and circuits shall be powered from the emergency buses.
- 4. Necessary pumps and valves shall be included in the automatic
> sequence of the loads to the emergency buses. Verify that the addition of these loads does not compromise the emergency diesel generating capacity.
- 5. Failure in the automatic circuits shall not result in the loss of manual capability to initiate the ATWS from the control room.
- 6. Other Considerations
- a. For those designs where instrument air is needed for operation the electric power supply requirement should be capable of being manually connected to emergency power sources.
BWR OWNERS' GROUP DISCUSSION: None i l BWR OWNERS' GROUP IMPLEMENTATION CRITERIA: None t LT.LCO'S RESPONSE: l Shoreham is a General Electric BWR. Since this requirement l is PWR specific, it does not apply to Shoreham. l l 1 m 2.1.7.a-2 l
SNPS-1 RESPONSE TO NUREG 0578 ( 2.1.7.b Auxiliary Feedwater Indication to Steam Generators for PWRs NUREG 0578 POSITION: Consistent with satisfying the requirements set forth in GDC 13 to provide the capability in the control room to ascertain the actual performance of the ATWS when it is called to perform its intended function, the following requirements shall be implemented:
- 1. Safety-grade indication of auxiliary feedwater flow to each steam generator shall be provided in the control room.
- 2. The auxiliary feedwater flow instrument channels shall be powered from the emergency buses consistent with satisfying the emergency power diversity requirements of the auxiliary feedwater system set forth in Auxiliary Systems Branch Technical Position 10-1 of the Standard Review Plan, Section 10.4.9.
NRC CLARIFICATION: A. Control Grade (Short-Term) (
- 1. Auxiliary feedwater flow indication to each steam generator shall satisfy the single failure criterion.
- 2. Testability of the auxiliary feedwater flow indication channels shall be a featute of the design.
- 3. Auxiliary feedwater flov instrument channels shall.be powered from the vital instrument buses.
B. Safety-Grade (Long-Term)
- 1. Auxiliary feedwater flow indication to each steam generator shall satisfy safety-grade requirements.
C. Other
- l. For the Short-Term the flow indication channels should by themselves satisfy the single failure criterion for each steam generator. As a fall-back position, one auxiliary feedwater flow channel may be backed up by a steam generator level channel.
. 2. Each auxiliary feedwater channel should provide an indication
{] of feed flow with an accuracy on the order of t 10%. - l \~J - 2.1.7.b-1 i l , -- - . - _ . - . . - . _ , - . _ . . - . - - .- _ - . _ - - --. ~ .. . . ..
r SNPS-1 RESPONSE TO NUREG 0578 BWR OWNERS' GROUP DISCUS ION: None BWR OWNERS' GROUP IMPLEMENTATION CRITERIA: None LILCO'S RESPONSE: Shoreham is a General Electric BWR. Since this requirement is PWR specific, it does not apply to Shoreham. f ) i l l e 2.1.7.b-2
J SNPS-1 RESPONSE TO MUREG 0578
) /
2.1.8.a Improved Post-Accident Sampling Capability NUREG 0578 POSITION: A design and operational review of the reactor coolant and con-tainment atmosphere sampling systems shall be performed to de-termine the capability of personnel to promptly obtain (less than 1 hour) a sample under accident conditions without in-curring a radiation exposure to any individual in excess of 3 and 18 3/4 Rems to the whole body or extremitics, respectively. Accident conditions should assume a Regulatory Guide 1.3 or 1.4 release of fission products. If the review indicates that personnel could not promptly and safply obtain the samples, ad-ditional design features or shielding should be provided to meet the criteria. A design and operational review of the radiological spectrum analysis facilities shall be bility to promptly quantify (performed less thanto2 determine the capa-hours) certain radio-isotopes that are indicators of the degree of core damage.
,s Such radionuclides are noble gases (which indicate cladding
( \ failure), iodines and cesiums (which indicate high fuel tempera- \s l tures), and non-volatile isotopes (which indicate fuel melting). The initial reactor coolant spectrum should corresaond to a Regulatory Guide 1.3 or 1.4 release. The review should also consider the effects of direct radiation from piping and com-ponents in the auxiliary building and possible contamination and direct radiation from airborne effluents. If the review indicates that the analyses required cannot be performed in a prompt manner with existing equipment, then design modifications or equipment procurement shall be undertaken to meet the cri-teria. In addition to the radiological analyses, certain chemical anal-yses are necessary for monitoring reactor conditions. Pro-cedures shall be provided to perform boron and chloride chemi-cal analyses assuming a highly radioactive initial sam 31e (Regu-latory Guide 1.3 or 1.4 source term). Both analyses shall be capable of betng completed promptly; i.e., the boron sample anal-ysis within an hour and the chloride sample analysis within a shift.
\/ )
2.1.8.a-1
SNPS-1 RESPONSE TO NUREG 0578 {N NRC CLARIFICATION: The licensee shall have the capability to promptly obtain (in less than 1 hour) pressurized and unpressurized reactor coolant samples and a containment atmosphere (air) sample. The licensee shall establish a plan for an onsite radiological and chemical analysis facility with the capability to provide, within 1 hour of obtaining the sample, quantification of the following:
- 1. certain isotopes that are indicators of the degree of core damage (i.e., noble gases, iodines and cesiums and non-volatile isotopes),
2, hydrc3en levels in the containment, atmosphere in the range 0 to 10 volume percent,
- 3. dissolved gases (i.ej , H2 , 0 2) and boron concentration of liquids.
or have in-line monitoring capabilities to perform the above analysis. Plant procedures for the handling and analysis of samples, minor plant modifications for taking samples and a design review and pro-cedural modifications (if necessary) shall be completed by January 1, 1980. Major plant modifications shall be completed by January 1, s 1981. s During the review of the post accident sampling capability consider-ation should be given to the following items:
- 1. Provisions shall be made to permit containment atmosphere sampling under both positive and negative containment pressure.
- 2. The licensee shall consider provisions for purging samples lines, for reducing plateout in sample lines, for minimizing sample loss or distortion, for preventing blockage of sample lines by loose materf.al in the RCS or containment, for appropriate disposal of the samples, and for passive flow restrictions to limit reactor coolant loss or containment air leak from a rupture of the sample line.
- 3. If changes or modifications to the existing sampling system are required, the seismic design and quality group classifi-cation or sampling lines and components shall conform to the classification of the system to which each sampling line is connected. Components and piping downstream of the second isolation valve can be designed to quality Group D and non-seismic Category I requirements.
Q NJ - 2.1.8.a-2
SNPS-1 RESPONSE TO NUREG 0578 The licensee's radiological sample analysis capability should include provisions to:
- a. Identify and quantify the isotopes of the nuclide categories discussed above to levels corresponding to the source terms given in Lessons Learned Item 2.1.6.b. Where necessary, ability to dilute samples to provide capability for measure-ment and reduction of personnel exposure, should be provided.
Sensitivity of onsite analysis capability should be such as to permit measurement of nuclide concentration in the range from approximately 1 AtCi/gm to the upper levels indicated here.
- b. Restrict background levels of radiation in the radiological and chemical analysis facility from sources such that the sample analysis will provide results with an acceptably small error (approximately a factor of 2). This can be accomplished through the use of sufficient shielding around samples and outside sources, and by the use of ventilation system design which will control the presence of airborne radioactivity.
s c. Maintain plant procedures which identify the analysis required, measurement techniques and provisions for reducing background levels. The licensees chemical analysis capability shall consider the presence of the radiological source term indicated for the radiological analysis. 1 In performing the review of sampling and analysis capability, consid-eration shall be given to personnel occupational exposure. Procedural changes and/or plant modifications must assure that it shall be possible to obtain and analyze a sample while incurring a radiation dose to any individual that is as low as reasonably achievable and not in excess of GDC 19. In assuring that these limits are met, the following criteria will be used by the staff.
- 1. For shielding calculations, source terms shall be as given in Lessons Learned Item 2.1.6.b.
- 2. Access to the sample station and the radiological and chemical analysis facilities shall be through areas which are accessible in post accident situations and which are provided with sufficient shielding to assure that the radiation dose criteria are met.
- 3. Operations in the sample station, handling of highly radio-i active samples from the sample station tc the analysis s_,/ facilities, and handling while working with the samples -
2.1.8.a-3 l
SNPS-1 RESPONSE TO NUREG 0578 in the analysis facilities shall be such that the radiation
\
dose criteria are met. This cay involve sufficient shielding of personnel from the samples and/or the dilution of samples for analysis. If the existing facilities do not satisfy these criteria, then additional design features, e.g., additional shielding, remote handling, etc., shall be pro-vided. The radioactive sample lines in the sample station, the samples themselves in the analysis facilities, and other radioactive lines of the vicinity of the sampling station and analysis facilities shall be included in the evaluation.
- 4. High range portable survey instruments and personnel dosi-meters should be provided to permit rapid assessment of high exposure rates and accumulated personnel exposure.
The licensees shall demonstrate.their capability.to obtain and analyze a sample containing the isotopes discussed above according to the criteria given in this section. BUR OWNERS' GROUP DISCUSSION: The BWR Owners' Group agrees with the intent of the staff's position. BWR OWNERS' GROUP IMPLEMENTATION CRITERIA: A design and operational review of existing reactor coolant and containment atmosphere sampling facilities was completed by January 1, 1980. Modifications will be made to provide the capability to promptly obtain pressurized and unpressurized reactor coolant samples and containment atmosphere samples. Analysis capability shall be provided to identify and quantify (1) certain isotopes that are indicators of core damage (i.e., noble gases, iodines and cesiums, and non-volatile isotopes), and (2) dissolved gases (i.e., H 2 and 0 2 ) and boron concentration of liquids. These modifications will be complete by January 1, 1981. Until the design modifications are complete, procedures will be devised to evaluate the primary coolant system and containment environment activity depending on the accessibility of the sampling stations for particular degraded conditions. LILCO'S RESPONSE: A post Accident Sampling System will be provided for the Shoreham O facility. This system will be housed in a new Post Accident Sample Building adjacent to the Reactor Building. A conceptual - 2.1.8.a-4
SNPS-1 RESPONSE TO NUREG 0578 layout of this building is shown on Figure 2.1.8.a-l. This building will be accessible following an accident (Reactor . Building entry not necessary) to obtain and analyze samples of the reactor coolant, containment atmosphere and suppres-sion pool water. The outer walls and roof of the Sample Building will be shielded to reduce interior radiation levels below acceptable background levels for oersonnel protection. A separate intake filter Heating, Ventilating and Air Con-ditioning (HVAC) System will be provided for this building. During accident conditions the sampling enclosure within this building will be isolated and tied into the Reactor Building Standby Ventilation Building. The Post Accident Sagling System, with the exception of con-tainment isolation valves, will be operated from a control panel located in the Sample Building. This System will pro-vide the following capability:
- 1. Sampling of reactor coolant and suppression pool liquids and containment atmosphere.
l
- 2. On-line gross gamma activity levels monitoring.
- 3. Dilution of liquid, reactor coolant gases, or con-tainment atmosphere samples by either volumetric g or feed and bleed methods, for laboratory analysis including gamma spectrum analysis.
- 4. Either diluted, unpressurized degassed grab samples or pressarized, undegassed, and undiluted grab samples may be obtained.
- 5. Dissolved gas d'etection range from 1% to maximum gas concentration with better than + 10% accuracy.
Cotic'.tuous monitoring of containment hydrogen and oxygen levels is provided as part of the primary containment atmospheric control system. The system will be designed and shielded so that required sagles can be taken inside the facility under worst case con-l ditlons such that the combined dose to the operator from sample fluids and from the accident environment does not exceed 3 Rem whole body or 18 3/4 Rem to the extremities. In addition, the system will be designed to keep routine o7erating, testing and maintenance doses As Low As Reasonably Achievable.
\ -
! 2.1.8.a-5
SIES-1 RESPONSE TO NUREG 0578 O The sample system will be designed to non-safety grade re-quirements,~aut will be supplied with a reliable source of electric power to assure proper operation following an ac-cident. In addition, the sample isolation valves in the reactor building will be safety grade and redundant to comply with containment isolation requirements. Contain-ment isolation valves will be provided with automatic isola-tion signals and override capability from the main control room. O O 2.1.8.a-6
O
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SNPS-1 RESPONSE TO NUREG 0578 ( V 2.1.8.b Increase Range of Radiation Monitors NUREG 0578 POSITION: The requirements associated with this recommendation should be considered as advanced implementation of certain requirements to be included in a revision to Regulatory Guide 1.97, "Instru-mentation to Follow the Course of an Accident", which has already been initiated, and in other Regulatory Guides, which will be promulgated in the near-term.
- 1. Noble gas effluent monitors shall be installed with an extended range designed to function during accident con-ditions as well as during normal operating conditions; multiple monitors are considered to be necessary to cover the ranges of interest.
- a. Noble gas efflyent monitors with an upper range
! capacity of 10D ACi/cc (Xe-133) are considered to be practical and should be installed in all operat-ing plants.
- b. Noble gas effluent monitoring shall be provided for the total range of concentration extending from normal cogdition (ALARA) concentrations to a maxi-mum of 100,aci/cc (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.
- 2. Since iodine gaseous effluent monitors for the accident condition are not considered to be practical at this time, capability for effluent monitoring of radioiodines for the accident condition shall be provided with sampl-ing conducted by absorption on charcoal or other media, follcwed by onsite laboratory analysis.
- 3. In-containmgnt radiation level monitors with a maximum range of 10 rad /hr shall be installed. A minimum of two such monitors that are 7hysically separated shall be pro-vided. Monitors shall be designed and qualified to func-tion in an accident environment.
O m 2.1.8.b-1
SNPS-1 RESPONSE TO NUREG 0578 NRC CLARIFICATION (LETTER OF NOVEMBER 9,1979) ()1. Radiological Noble Gas Effluent Monitors
- a. January 1, 1980 Requirements Until final implementation in January 1, 1981, all operating reactors must provide, by January 1, 1980, an interim method for quantifying high level releases which meets the requirements of Table 2.1.8.b.l. This method is to serve only as a provisional fix with the more detailed, exact methods to follow. Methods are to be developed to quantify release rates of up to 10,000 Ci/see for noble gases from all potential release points (e.g., auxiliary building, radwaste building, fuel handling building, reactor building, waste gas decay tank releases, main condenser air ej ector, BWR main condenser vacuum pump exhaust, PWR steam safety valves and atmosphere steam dump valves and BWR turbine buildings) and any other areas that communicate *directly with systems which may contain primary coolant or containment gases, (e.g.,
letdown and emergency core cooling systems and external recombiners). Measurements / analysis capabilities of the effluents at the final release point (e.g., stack) should be such that measurements of individual sources which contribute to a common release point may not be necessary. For assessing radiciodine and particulate releases,
) snecial procedures must be developed for the removal and sm ,/ analysis of the radiciodine/ particulate sampling media (i.e., charcoal canister / filter paper). Existing, sampling locations are expected to be adequate; however, special procedures for retrieval and anal media under accident conditions (ysis e.g., ofhithe h sampling air and surface contamination and direct radiation levels are needed.
It is intended that the monitoring capabilities called for in the interim can be accomplished with existing instrumentation or readily available instrumentation. For noble gases, modifications to existing monitoring systems, instruments, suchset asinthe use of portable shielded high collimators sorange survey"see" that they I small sections of scmpling lines is an acceptable method for meeting the intant of this re Conversion l of the measured dose rate (mR/hr)quirement. into concentration I (gCi/cc) can be performed using standard volume source ! calculations. A method must be developed with sufficient ' accuracy to quantify the iodine releases in the presence of high background radiation from noble gases collected on charcoal filters. SeLaically qualified equipment and equipment meeting IEEE-279 is not required. Ne# 2.1.8.b-2
g _ T SNPS-1 RESPONSE TO UUREG 0578
'N The licensee shall provide the following information on
(_a his methods to quantify gaseous releases of radioactivity from the plant during an accident. - (1) Noble Gas Effluents a) System / Method description including: i) Instrumentation to be used including range or sensitivity, energy dependence, and calibration frequency and technique,
- 11) Monitoring / sampling locations, including methods to assure renresentative measurements
^
and background radiation correction, iii) A description of method to be employed to facilitate access to radiation readings. For January 1, 1980, Control room read-out is preferred: however, if impractical, in-situ readings by an individual with verbal communication with the Control Room is acceptable based on (iv) below.
' ~ '
iv) Capability to obtain radiation readings at
; least every 15 minutes during an accident.
v) Source of power to be used. If normal AC power is used, an alternate back-up power supply should be provided. If DC power is used, the source should be capable of providing continuous readout for 7 consecutive days. b) Procedures for conducting all aspects of the measurement / analysis including: i) Procedures for minimizing occupational exposures. ii) Calculational methods for converting instru-ment readings to release rates based on exhaust air flow and taking into consideration radionuclide spectrum distribution as function of time after ' shutdown. iii) Procedures for dissemination of information. iv) Procedures for calibration,
- b. January 1, 1981 Requirements
~.
By January 1, 1981, the licensee shall provide high range noble gas effluent monitors for each release path. The 2.1.8.b-3 .
. SNPS-1 RESPONSE TO NUREG 0578 noble gas effluent monitor should meet the requirements of Table 2.1.8.b.2. The licensee shall also provide O. the information given in Sections 1. A.1.a.i. , l.A.l.a.ii,
*1.A.1.b.ii, l.A.l.b.iii, and 1.A.l.b.iv above for the noble gas effluent monitors.
- 2. Radioiodine and Particulate Effluents
- a. For January 1, 1980, the licensee should provide the following:
(1) System / Method description including: a) Instrumentation to be used for analysis of the sampling media with discussion on methods used to correct for potentially interfering back-ground levels of radioactivity. b) Monitoring / sampling location. c) Method to be used for retrieval and handling of sampling media to minimize occupational exposure. d) Method to be used for data analysis of individual radionuclides in the presence of high levels of radioactive noble gases. r e) If normal AC power is used for sample collection and analysis equipment, an alternate back-up power supply should be provided. If DC power is used, the source should be capabic of providing continuous read-out for 7 consecutive days. (2) Procedures for conducting all aspects of the measure-ment analysis including: a) Minimizing occupational exposure. b) Calculational methods for determining release rates. c) Procedures for dissemination of information. d) Calibration frequency and technique.
- b. For January 1, 1981, the licensee should have the capa-bility to continuously sample and provide onsite analysis of the sampling media. The licensee should also provide the information required in 2.a. above.
2.1.8.b-4
SNPS-1 ! RESPOUSE TO NUREG 0578 O) ( 3. Containment Radiation Monitors Provide by January 1, 1981, tuo radiation monitor systems in containment which are documented to meet the requirements of Table 2.1.8.b.3. It is possible that future regulatory requirements for emergency planning interfaces may necessitate i identification of different types of radionuclides in the ! containment air, e.g., noble gases (indication of core damage) l and non-volatiles (indication of core melt). Consequently, consideration should be given to the possible installation l or future conversion of these monitors to perform this function. l BWR OWNERS' GROUP DISCUSSION: The Owners' Group recognizes and concurs with the positions as modified in the NRC regional meetings the week of September 24, 1979. BWR OWNERS' GROUP IMPLEMENTATION CRITERIA
- 1. The Owners will implement the requirements of position 2.1.8.b, items 1, 2 and 3 (provided in NRC clarification letter of November 9, 1979), consistent with commercial h availability of equipment.
[G 2. Procedures will be develooed to estboate noble gas and radio-iodine release rates if the existing effluent instrumentation goes off scale. LILCO'S RESPONSE: The effluent monitors in NUREG 0578, as clarified in NRC letter from D. B. Vassallo dated November 9, 1972, Table 2.1.8.b-2, which apply to the Shoreham Nuclear Power Station are: (a) "di-luted containment exhaust", (b) "other buildings", and (c) " build-ings with systems containing orimary coolant or gases". See Figure 2.1.8.b.1 for a simplified diagram of Shoreham's gaseous effluent layout. The maximum anticipated primary containment leakage rate is 0.005 volumes pgr day into the secondary containment which has a volume of 2 X 100 cubic feet. The primary containment exhaust in highly diluted in the secondary containment atmosphere. This mixture will be discharged after passing through high efficiency particulate abselute filters and charcoal ads,rber banks via the Reactor Build-ing Standby Ventilation System (1:3SVS) discharge nipe, at the top of the Station Vent Exhaust. The two Class lE ra3iation monitors ('T (RE-021 and RE-022) serving thisfrom system \g J and adsorbers will have a range 1 X downsbream of themicro-10- to 1 X 10+4 filters - curies /cc. 2.1.8.b-5
SNPS-1 RESPONSE TO NUREG 0578 O The RBSVS monitors are supplied with power from vital instrument buses. These monitors read out in the Control Room and are lo- ! cated in the Control Building to permit access during an accident for collection of their radiciodine and particulate sample media for laboratory analysis. The capability to provide readout of these monitors in the Technical Support Center and in the Emer-gency Off-site Facility is under evaluation. The criteria in Table 2.1.8.b-2 for other buildings and buildings j with systems containing primary coolant or gases are applicable to the Station Vent Exhaust monitor (RE-042). Normal ventilation discharges from the reactor building, the turbine building and the radwaste building are mixed,,thereby providing dilution prior to being exhausted through the Station Vent Exhaust. During an accident when RBSVS is operating, and the Reactor Building Normal Ventilation System (RBNVS) is isolated, the loss of normal reactor building ventilation flow is compensated by opening louvers at the Station Vent Exhaust to aermit 90,000 cubic feet / minute of outside air for dilution. This single discharge point for the combined ventilation flow from all potentially contaminated build-ings will be monitored by a noble gas radiation monitor (RE-042) having a range of 1 X 10-6 to 1 X 10+2 microcuries/cc. e moni-tor is backed up by RE-069 with an upper range of 1 X 10 . In addition, the individual building ventilation flows to the Sta-C_ tion Vent Exhaust are each analyzed by a high range in-line radi-ation 1 X 10N nitor (RE-066, RE-067, RE-068) with an upper range of microcuries/cc. All these monitors, except RE-042, are powered from a vital instrument bus. The normal Station Vent Exhaust monitor (RE-042) is not powered from a vital instrument bus and, due to its location in the secondary containment, it may be inaccessible during an accident. This would preclude obtaining the radiciodine and particulate sample media from the monitor for analysis. However, inability to obtain these samples is compensated for by the fact that the turbine building and radwaste building ventilation flows are each sampled for radiciodine and particulates by the equipment associated with the normal range noble gas monitors for these flows (RE-057 and RE-055). These monitors are both located in the turbine building permitting access for collection of the sample media during an accident in order that laboratory analysis may be nerformed. Adding the results obtained for radioiodine or particulates from the turbine building and radwaste building ventilation flows will give the radiciodine or particulate re-lease.at the Station Vent Exhaust should the secondary contain-ment be inaccessible. Under these circumstances, RBSVS is op-erating and there is no reactor building ventilation contribu-tion to the Station Vent Exhaust. As discussed above, the RBSVS release is monitored separately (RE-021, RE-022). The O monitors associated with the reactor, radwaste and turbine build- . ings ventilation systems are not powered from a vital bus. This is consistent with the design of the monitored systems. More 2.1.8.b-6
SNPS-1 ' RESPONSE TO NUREG 0578 directly, the Station Vent Exhaust monitor's (RE-042) radio-l iodine and particulate sample media can be obtained for anal-ysis if the secondary containment is accessible. Initial calibration will include detector response for a mini-mum of three decades using standard sources of two different j energies and intensities. These calibration curves will be initially generated using both gaseous and solid sources, where practical. Routine calibration of these monitors will be
- in accordance with technical specifications provisions using j solid sources related to the initial calibration.
The conversion of the instrument readings to release rates will be determined using the energy response of the detectors ob-tained during calibration. Accident release rates will then be calculated based on anticipated radionuclide inventories fol-lowing a design basis loss of coolant accident. Actual re-leases may be determined by analyzing a grab sample and cor-recting the release rates calculated. l Background radiation will not substantially affect readings on the RBSVS noble gas monitors (RE-021 and 022) during an ac-cident, due to their location in the control building and the detector's location in a 4$ lead shield. For the Station Vent ! Exhaust Monitor (RE-042), background radiation in the vicinity of the monitor within the secondary containment will not sub-stantially affect the noble gas detector, due to its location in a 4ft lead shield and the fact that the detector is a thin ! beta scintillator. This type of detector is very inefficient for detecting gamma radiation which might penetrate the lead shield, while they are efficient for detecting the beta radia-tion associated with the sample stream's noble gases brought in close contact with the detector. i The capability to readout the Station Vent Exhaust noble gas monitor (RE-042) and the individual building ventilation and Station Vent Exhaust in-line high range monitors (RE-066, RE-067, RE-068, RE-069) at the TSC and the Emergency Off-Site Facility is also under evaluation. The radioiodine and particulate sampling media will be analyzed in the counting room at Shorehap. Charcoal cartridges will be purged with air to remove interfering noble gases. In order to facilitate analysis of the radiciodine and particulate samole media, various features have been included ln the Radiochemistry laboratory and counting e under adverse conditions.quipment designs In addition, to permit analysis consideration is being given to establishing a separate accident laboratory area to include counting equipment at a location on-site and establish-ing backup counting capability at a nearby facility with the required equipment and expertise. Further, procedures will be
,~
2.1.8.b-7
SNPS-1 RESPONSE TO NUREG 0578 I prepared for conducting all aspects of the measurement and analyses correctly and in a manner to minimize personnel ex-posure. Procedures for dissemination of information will also be prepared. Two physically separate monitors will be ingtalled inside the drywell having a range of 1 X 101 to 1 X 10' Roentgens / hour for photon radiation. These monitors will be each powered by a vital instrument bus, will be seismic qualified, and will be designed to withstand the temperatures, pressures, humidity and total radiation in the drywell containment through an accident. Moni-tor readouts will be displayed continuously and recorded on a Category I panel in the Main Control Room. Additionally, two monitors, range.1 to 1 X 106 R/hr, will be mounted one on the outside of the personnel hatch and the other on the outside of the equipment hatch. These monitors will provide containment radiation readings during an accident. These monitors meet the requirements of Table 2.1.8.b.3 with the exception of qualifica-tion to ANSI-N320-1979. For a listing of the radiation monitors with the ranges provided, refer to Table 2.1.8.b.4. O l 2.1.8.b-8 O g .
( ATTACEMENT TO NRC CLARIFICATION TABLE 2.1.8.b.1 INTERIM PROCEDURES FOR QUANTIFYING HIGH LEVEL ACCIDENTAL RADIOACTIVITY RELEASES Licensees are to implement procedures for estimating noble gas and radiciodine release rates if the existing effluent instrumentation goes off scale.
,i Examples of major elements of highly radioactive effluent release special procedures (noble gas).
Preselected location to measure radiation from the exhaust air, e.g., exhaust duct or sample line. Provide shielding to minimize background interference. i Use of an installed monitor (preferable) or dedicated portable monitor (acceptable) to measure the radiation. Predetermined calculational method to convert the radiation level to radioactive effluent release rate. e 2.1.8.b-9
ATTACHNENT TO NRC CLARIFICATION TABLE 2.1.8.b.2 (-% s/ HIGH RANGE EFFLUENT MONITOR
. NOBLE GASES ONLY
. RANGE: (Overlap with Normal Effluent Instrument Range)
UNDILUTED CONTAINMENT EXHAUST 10+5 . mci /CC DILUTED (010: 1) CONTAINMENT EXHAUST 10+4 ACi/CC MARK I BWR REACTOR BUILDING EXHAUST 10+4 Aci/CC PWR SECONDARY CONTAINMENT EXHAUST 10+4 ,4Ci/CC BUILDINGS WITH SYSTEMS CONTAINING PRIMARY COOLANT OR GASES 10+3 #Co/CC OTHER BUILDINGS (E.G., RADWASTE) 10+2 JACi/CC m
\
. NOT REDUNDANT - 1 PER NORMAL RELEASE POINT
. SEISMIC - NO
. POWER - VITAL INSTRUMENT BUS
. SPECIFICATIONS - PER R.G. 1.97 AND ANSI N320-1979
. DISPLAY *: CONTINUOUS AND RECORDING WITH READOUTS IN THE TECHNICAL SUPPORT CENTER (TSC) AND EMERGENCY OPERATIONS CENTER (EOC)
. QUALIFICATIONS - NO O *Although not a present requirement, it is likely that this --
information may have to be transmitted to the NRC. Consequently, - consideration should be given to this possible future require-ment when designing the display interfaces. 2.1.8.b-10 i
. - ~
ATTACHMENT TO NRC CLARIFICATION TABLE 2.1.8.b.3 HIGH RANGE CONTAINMENT RADIATION MONITOR
. RADIATION: TOTAL RADIATION (ALTERNATE: PHOTON ONLY)
. RANGE:
- ,UP TO 10 8 RAD /HR (TOTAL RADIATION)
ALTERNATE: 107 R/HR (PHOTON RADIATION ONLY) SENSITIVE DOWN TO 60 KEV PHOTONS *
. REDUNDANT: TWO PHYSICALLY SEPARATED UNITS
. SEISMIC: PER R. G. 1.97 ,
[ . POWER: VITAL INSTRUMENT BUS
. SPECIFICATIONS: PER R. G. 1.97 REV. 2 and ANSI N320-1978
. DISPLAY: CONTINUOUS AND RECORDING
. CALIBRATION: LABORATORY CALIBRATION ACCEPTABLE d
- Monitors must not provide misleading information to the operators assuming delayed core damage when the 80 kev photon Xe-133 is the O
major noble gas present. _ 2.1.8.b-11
TABLE 2.1.8.b.4 O V RADIOACTIVITY CONCENTRATION RANGES FOR SHOREHAM GASEOUS EFFLUENT RADIATION MONITORS RANGE GASEOUS EFFLUENT MONITOR (microcuries/cc) Reactor Building Standby Ventilation RE-021, RE-022* lx10-6 to 1x10+4 Normal ReactorBuildinb29* Ventilation RE- lx10-6 to lx10-1 Turbine Building Ventilation RE-057* lx10
-6 to lx10-1 Radwaste Building Ventilation RE-055* lx10-6 to Ir.10-i Station Vent Exhaust RE-042* lx10-6 to 1x10+2 Reactor Building Normal Ventilation RE-068 lx10
-2 to lx10+3 Turbine Building Ventilation RE-067 lx10
-2 to lx10+3 Radwaste Building Ventilation RE-066 lx10-2 to 1x10+3
-2 Station Vent Exhaust RE-069 lx10 to lx10+3 1 7 Drywell Monitors 1x10 to lx10 R/hr 0 6 Personnel Hatch lx10 to 1x10 R/hr Equipment Hatch lx10 0 to lx10 6 R/hr i
- Ranges shown for these radiation monitors are for the noble gas portion of the monitor.
2.1.8.b-12
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SNPS-1 RESPONSE TO NUREG 0578 C Improved In-Plant Iodine Ittstrumentation 2.1.8.c NUREG 0578 POSITION: Each licensee shall provide equipment and associated training and procedures for accurately determining the airborne iodine concentration in areas within the facility where plant personnel may be present during an accident, NRC CLARIFICATION: Use of Portable versus Stationary Monitoring Equipment Effective monitoring of increasing iodine levels in the buildings under accident conditions must include the use of portable instru-ments for the following reasons:
- a. The physical size of the auxiliary / fuel handling building precludes locating stationary monitoring instrumentation at all areas where airborne iodine concentration data might be required.
- b. Unanticipated isolated " hot spots" may occur in locations (r3) where no stationary monitoring instrumentation is .ocated.
- c. Unexpectedly high background radiation levels near stationary monitoring instrumentation after an accident may interfere with filter radiation readings.
- d. The time required to retrieve samples after an accident may result in high personnel exposures if these filters are located in high dose rate areas.
Iodine Filters and Measurement Techniques A. The following are short-term recommendations and shall be Lmplemented by the licensee by January 1, 1980. The licensee shall have the capability to accurately detect the presence of iodine in the region of interest following an accident. This can be accomplished by using a portable or cart-mounted iodine sampler with attached single channel analyzer The SCA windowshouldbecalibrctedtothe365keVof{ggA). I. A repre-sentative air sample shall be taken and then counted for 1311 using the SCA. This vill give an initial conservative estimate of presence of iodine and can be used to determine if respiratory protection is required. Care must be taken to assure that the counting system is not saturated as a result of too much activity collected on the sampling cartridge. ,, 'h . 2.1.8.c-1
SNPS-1 RESPONSE TO NUAEG 0578 B. By January 1, 1981: The licensee shall have the capability to remove the sampling cartridge to a low background, low contamination area for further analysis. This area should be ventilated with clean air containing no airborne radionuclides which may contribute to i:. accuracies in analyzing the sample. Here, the sample should first be purged of any entrapped noble gases using nitrogen gas or clean air free of noble gases. The licensee shall have the capability to measure accurately the iodine concentrations present on these samples and effluent charcoal samples under accident conditions. BUR OWNERS' GROUP DISCUSSION: i The Owners' Group recognizes and concurs with the position. BWR OWNERS' GROUP IMPLEMENTATION CRITERIA:
- 1. The Owners will implement the requirements of position 2.1.8.c.
- 2. Procedures will be developed to accurately determine in-plant iodine concentrations.
J LILCO'S RESPONSE: i The iodine concentrations will be determined by utilizing appro-priate in-plant instrumentation. Portable, semi-portable or fixed air samplers will be used to pump a known quantity of air through a charcoal filter. A gross count will then be performed on the charcoal filter cartridgs to ascertain if a significant amount of radioactivityhasgeenabsorbed. is above 9.0 X 10-If the resulting gross activity will be performed. A(ci/cc unidentified, a gamma spectrum analysis i The gamma spectrum analysis will identify the 364 kev peak for I-131 as well as its confirming secondary peak. This analysis along with previous energy and efficiency calibrations of the equipment will permit quantifyin and identifying which nuclide(s)gare thepresent. radioactivity in the The use of samnle respir-atory protection will then be based on the concentration of each identified nuclide present and its maximum permissable concentra-tion as indicated in 10CFR20, Appendix B. 4 2.1.8.c-2
i i i l SNPS-1 RESPONSE TO NUREG 0578 ( . 2.1.9 Analysis of Design and Off-Normal Transient and Accidents
~
NUREG 0578 POSITION: Analyses, procedures and training addressing the following are required:
- 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 additio'n, 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 pro-cedures and in support of an eventual long term verification of compliance with Appendix K of 10 CRF Part 50. In the analysis of inadec . ditions shall be analyzec uate usingcore cooling realistic (,best-estimate) the following methods: con-
- 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 inadec uate core cooling is approached as well as the period of time c uring which inadequate core cooling exists. The calcula-tions shall be carried out in real time far enough that all im-portant phenomena and instrument indications are included. Each case should then be repeated taking credit for correct operator j action. These additional cases will~ provide the basis for de-veloping 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 this appendix). l l The analyses of transients and accidents shall include the de-l sign basis events specified in Section 15 of each FSAR. The .. analyses shall include a single active failure for each system . 2.1.9-1
. SNPS-1 RESPONSE TO NUREG 0578 f~'sI . called upon to function for a particular event. Consequential t \'-
failures shall also be considered. Failures of the operators to perform required control manipulations shall be given con-sideration for per=utations of the analyses. Operator actions that could cause the complete loss of function of a safety sys-tem shall also be considered. At present, these analyses need not address passive failures or multiple system failures in the short term. In the recent analysis of small break LOCAs, complete loss of auxiliary feedwater was considered. The com-plete loss of auxiliary feedwater may be added to the failures aeing considered in the analysis of transients and accidents if it is concluded that more is needed in operator training be-yond the short-term actions to upgrade auxiliary feedwater system reliability. 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 gs \ corrective action. Reactor simulators may provide some informa-(N/ tion in defining the event trees and would be useful in study-ing the information available to the operators. The transient and accident analyses are to be performed for the purpose of identifying apporpriate and inappropriata operator actions relat-ing to important safety considerations such as natural circula-tion, prevention of core uncovery, and prevention of more serious accidents. , The information derived from the preceding analyses shall be included in the plant emergency procedures and operator train-4 ing. It is expected that analyses performed by the NSSS vendors will be put in the form of emergency procedure guidelines and that the changes in the procedures will be implemented by each licensee or applicant. 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 comparisons with the analytical methods being used by the reactor vendors. These comparisons, together with comparisons to data, including LOFT small break test data, will constitute the short-term verification effort to assure the adequacy of the analytical methods being used to generate emergency procedures. fN - 2.1.9-2
SNPS-1 l RESPONSE TO NUREG 0578 l NRC CLARIFICATION: Containment Pressure Indication and Containment Hydrogen Indication
- 1. The containment pressure indication shall meet the design provisions of Regulatory Guide 1.97 including qualification, redundancy, and testability.
- 2. The containment pressure monitor shall be installed by January 1, 1981.
Reactor Coolant System Venting A. General -
- 1. The two important safety functions enhanced by this venting
. capability are core cooling and containment integrity.
For events within the present design basis for nuclear power plants, the capability to vent non-condensible gases will provide additional assurance of meeting the require-ments of 10CFR50.46 (LOCA criteria) and 10CFR50.44 (contain-ment criteria for hydrogen generation). For events beyond the present design basis, this venting capability will substantially increase the plant's ability to deal with N large quantities of non-condensible gas without the loss of core cooling or containment integrity.
- 2. Procedures addressing the use of the RCS vents are required by January 1, 1981. The procedures should define the conditions under which the vents should be used as well as the conditions under which the vents should not be used.
The procedures should be based on the following criteria: (1) assurance that the plant can meet the re.quirements of 10CFR50.46 and 10CFR50.44 for Design Basis Accidents; and (2) a substantial increase in the plants ability to maintain core cooling and containment integrity for events beyond the Design Basis. B. BWR Design Considerations
- 1. Since the BWR Owners Group has suggested that the present BWR designs inherent capability of venting, this question relates to the capability of existing systems. The ability of these systems to vent the RCS of non-condensible gas must be documented. Since there are imoortant differences among BWR's, each licensee should addre'ss the specific design features of his plant.
- 2. In addition to reactor coolant system venting, each BWR licensee should address the ability to vent other systems
(}j g m such as the isolation condenser, which may be required to ," 2,1,9-3
SNPS-1 RESPONSE TO NUREG 0578 ('~h
\-- maintain adequate core cooling. If the production of a large amount of non-condensible gas would cause the loss of function of such a system, remote venting of that system is required. The qualifications of such a venting system should be the same as that required for PWR venting systems.
C. PWR Vent Design Considerations
- 1. The locations for PWR Vents are as follows:
- c. Each PWR licensee should provide the capability to vent the reactor vessel head.
- b. The reactor vessel head vent should be capable of venting non-condensible gas from the reactor vessel hot legs (to the elevation of the top of the outlet nozzle) and cold legs (through head jets and other leakage paths) .
Additional venting capability is required for those portions of each hot leg which cannot be vented through the reactor vessel head vent. The NRC recognizes that it is impractical to vent each of the many thousands of tubes in a U-tube steam generator. However, we believe that a procedure can be developed which assures that cs sufficient liquid or steam can enter the U-tube region ( ) so that deccy heat can be effectively removed from the
\m / reactor coolant system. Such a procedure is required by January 1981.
- c. Venting of,the pressurizer is required to assure its availability for system pressure and volume control.
These are important considerations especially during natural circulation.
- 2. The sizelof the reactor coolant vents is not a critical issue. The desired venting capability can be achieved with vents in a fairly large range of sizes. The criteria for sizing a vent can be developed in several ways. One approach l which we consider reasonable, is to specify a volume of l non-condensible gas to be vented and a venting time, i.e.,
j a vent capable of venting a gas volume of % the RCS in one I hour. Other criteria and engineering approaches should be considered if desired.
- 3. Where practical the RCS vents should be kept smaller than the size corresponding to the definition of a LOCA (10CFR50 Appendix A). This will minimize the challenges to the ECCS since the inadvertent opening of a vent smaller than the LOCA definition would not require ECCS actuation although it may result in leakage beyond Technical Specification
( Limits. On PWRs the use of new or existing valves which .. ['}
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2.1.9-4
SNPS-1 RESPONSE TO NUREG 0578 N {N-) are larger than the LOCA definition will require the addition of a block valve which can be closed remotely to terminate the LOCA resulting from the inadvertent opening of the vent.
- 4. An indication of valve position should be provided in the control room.
- 5. Each vent should be remotely operable from the control room.
- 6. Each vent should be seismically qualified.
- 7. The requirements for a safety grade system is the same as the safety grade requirement on other Short Term Lessons Learned items, that is, it should have the same qualifi-cations as were accepted for the reactor protection system when the plant was licensed. The exception to this require-ment is that we do not require redundant valves at each venting ~_ocation. Each vent must have its power supplied from an emergency bus. A degree of redundancy should be provided by powering different vents from different emergency buses.
(~T 8. For systems where a block valve is required, the block (_/ valve should have the same qualifications as the vent.
- 9. Since the RCS vent system will be part of the reactor coolant systems boundary, efforts should be made to mini-mize the probability of an inadvertent actuation of the system. Removing power from the vents is one step in the direction. Other steps are also encouraged.
- 10. Since the generation of large quantities of non-condensible gas could be associated uith substantial core damage, venting to atmosphere is unacceptable because of the associated released radioactivity. Venting into contain-ment is the only presently available alternative. Within containment those areas which provide good mixing with containment air are preferred. In addition, areas which provide for maximum cooling of the vented gas are preferred.
Therefore the. selection of a location for venting should take advantage of existing ventilation and heat removal systems.
- 11. The inadvertent opening of an RCS vent must be addressed.
For vents smaller than the LOCA definition, leakage detection must be sufficient to identify the leakage. For vents larger than the LOCA definition, an analysis (3) is required to demonstrate compliance with 10CFR50.46. + (/ 9 2.1.9-5
SNPS-1 RESPONSE TO NUREG 0578 BWR OWNERS' GROUP DISCUSSION: The specific requirements and schedules are being developed in a continuing series of meetings between the utility owners' groups and the NRC Bulletins and Orders Task Force. BWR OWNERS' GROUP IMPLEMENTATION CRITERIA: The implementation of emergency procedures and retraining will be done on a schedule consistent with those established with the Bulletins and Orders Task Force. LILCO'S RESPONSE: The NEDO-24708 report, prepared by the BWR Owners' Group, of which LILCO is a participant, contains state-of-the-art anal-yses for various postulated accidents. The pos.tulated ac-cidents considered in NEDO-24708 and other material subsequently submitted to the Bulletins and Orders Task Force by the BWR Owners' Group included:
- a. Small Break Loss-of-Coolant Accident,
- b. Steam Line Break,
( c. Detection and Mitigation of Inadequate Core Cooling,
- d. Feedwater Line Break, and
- e. Other Transients and Accidents analyzed in l Chapter 15 of the FSAR.
The rloove analyses considered various combinations of the safety-related equipment available at the time of the transient or ac-cident with the effect of operator actions also considered. The purpose of performing these analyses was to better understand the course of taese events so as to provide reactor operators with l realistic guidelines. General emergency guidelines, symptom l oriented, have been developed through the efforts of General Electric and the BWR Owners' Group. These guidlines have been submitted to the NRC by the Owners' Group and are being utilized in the development of Shoreham emergency procedures. These pro-cedures will Le completed prior to Shoreham startup. The Shoreham o7erator training program will assure that shift personnel are thoroughly familiar with the emergency procedures and respond adequately to transient and accident conditions. N'
) .
, 2.1.9-6
SNPS-1 s RESPONSE TO NUREG 0578 2.2.1.a Shift Supervisor's Responsibility NUREG 0578 POSITION:
- 1. The highest level of corporate management of each li-censee shall issue and periodically reissue a manage-ment directive that emphasizes the primary management responsibility of the shift supervisor for safe opera-tion of the plant under all conditions on his shift and that clearly establishes his command duties.
- 2. Plant procedures shall be reviewed to assure that the duties, responsibilities, and authority of the shift supervisor and control room operators are properly de-fined to effect the establishment of a definite line of command and clear delineation of the command de-cision authority of the shift supervisor in the con-trol room relative to other plant management person-nel. Particular emphasis shall be placed on the following:
'~
- a. The responsibility and authority of the shift super-visor 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 ac-cident situations to direct the activities of con-trol 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.
h 2.2.1.a-1
~
SUPS-1 RESPONSE TO NUREG 0578 O 3. Training programs for shift supervisors shall emohasize and reinforce the responsibility for safe operat' ion and the management function the shift supervisor is to provide for assuring safety.
- 4. The administrative duties of the shift supervisor shall be reviewed by the senior officer of each utility respon-sible for plant operations. Administrative functions that detract from or are sub irdinate to the management responsibili.y for assuring che safe operation of the plant shall be delegated te othet operations personnel not on duty in the control room.
NRC CLARIFICATION Shift Supervisor Resnonsibility (2.2.1.A) NUREG-0578 Position (Position No.) Clarification Highest Level of Corporate Mgmt. (1.) V.P. for Operations Periodically Reissue (1.) Annual Reinforcement of Company Policy O Management Direction (1.) Formal Documentation of Shift Personnel, All Plant Management, Copy to IE Region Properly Defined (2.0) Defined in Writing in a Plant Procedure Until Properly Relieved (2.B) Formal Transfer of Authority, Valid SRO License, Recorded in Plant Log Temporarily Absent (2.C) Any Absence Control Room Defined (2.C) Includes Shift Supervisor Office Adjacent to the Control Room Designated (2.C) In Administrative Procedures Clearly Specified Defined in Administrative Procedures 2.2.1.a-2
__ _ . = - - _ - - - . .. __ _ - _ - - _. . _ - t [ SUPS-1 , RESPONSE TO NUREG 0578 i
' I i Shift Supervisor Responsibility (2.2.1.A) (Continued)
NUREG-0578 Position (Position No.) Clarification I SRO Training Specified in ANS 3.1 (Draft) Section 5.2.1.8 Administrative Duties (4.) Not Affecting Plant i Safety Administrative Duties Reviewed (4.) On Same Interval as Reinforcement: 1.e., Annual by V.P. for 4 Operations.
; BWR OWNERS' GROUP DISCUSSION:
The Owners' Group agrees with the intent of the staff's position. 1 However, in order to remove any ambiguity from the meaning of the term " accident situation" in item 2.b of the staff's nosition in Appendix A of NUREG 0578*, the entire sentence will be interpreted as follows: The shift supervisor (or equivalent, such as the supervising control operator in some plants), until properly relieved, shall remain in the control room at all times whenever a site or general emergency has been declared to direct the activities of control room operators. BWR OWNERS' GROUP IMPLEMENTATION CRITERIA: The staff's position will be implemented as stated and subject to the interpretation of item 2.b, as discussed above. l LILCO'S RESPONSE: LILCO endorses the BWR Owners' Group position.
- 1. A management directive from the Vice President of Operations will be issued prior to fuel loading and annually reissued to clearly reinforce the Watch En-j gineer's** command duties and to emphasize that the prime responsibility of the Watch Engineer is the i safe operation of the plant.
'
- The shi.ft 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. _
** At the Shoreham facility,.the term Watch Engineer is synony-mous with Shift Supervisor.
2.2.1.a-3
SNPS-1 RESPONSE TO NUREG 0578 l
- 2. Station procedures will be updated so that the duties, responsibilities and authority of the Watch Engineer and Control Room Operators are explicitly defined and include the following items:
- a. The responsibility and authority of the Watch Engi-neer.will 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 ob-jective that the Watch Engineer should not become totally involved in any single operation in times of emergency, when multiple operations are required in the control room, will be reinforced.
, b. The Watch Engineer, until properly relieved, will remain in the control room at all tLmes whenever a site or general emergency has been declared to di-rect the activities of control room operators. ;
Persons authorized to relieve the Watch Engineer ' will be specified in appropriate procedures.
- c. Any time the Watch Engineer is temporarily absent from the control room during routine operations, O the lead control room operator will be designated to assume the control room command function. These temporary duties, responsibilities, and authority i
will be clearly specified in appropriate procedures.
- 3. Training programs for Watch Engineers will emphasize and reinforce the management functions of the Watch En-gineer, which are to provide important safe plant opera-tions.
- 4. The administrative duties of the Watch Engineer shall be reviewed annually by the Vice President of Opera-tions. Administrative functions that detract from or are subordinate to the Watch Engineer's management responsibility for assuring the safe operation of the plant will be delegated, whenever possible, to other operations personnel not on duty in the control room, p)
\ -
2.2.1.a-4
SNPS-1 RESPONSE TO NUREG 0578 s 2.2.1.b Shift Technical Advisor NUREG 0578 POSITION: . Each licensee shall provide an on-shift technical advisor to the shift supervisor. The shift technical advisor may serva 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. NRC CLARIFICATION: s
- 1. Due to the similarity in the requirements for dedication to safety, training and onsite location and the desire that the accident assessment function be performed by someone whose normal duties involve review of operating experiences, our preferred position is that the same people perform the accident and operating experience assessment functions. The performance of these two functions may be split if it can be demonstrated -
the persons assigned the accident assessment role are aware, on a current basis, of the work being done by those reviewing operating experience.
- 2. To provide assurance that the STA will be dedicated to concern for the safety of the plant, our position has been that STA's must have a clear measure of independence from duties associ-ated with the commercial operation of the plant. This would
, minimize possible distractions from safety judgments by the demands of commercial operations. We have determined that, while desirable, independence from the operations staff of the plant is not necessary to provide this assurance. It is necessary, however, to clearly emphasize the dedication to safety associated with the STA cosition both in the STA job description and in the personnel filling this position. It is not acceptable to assign a person, who is nor= ally the immediate supervisor of the shift supervisor to STA duties as defined herein.
~
'- 2.2.1.b-1
SNPS-1 RESPONSE TO NUREG 0578 p 3. V) t It is our position that the STA should be available within 10 minutes of being su=moned and therefore should be onsite. The onsite STA may be in a duty status for periods of time longer than one shift, and therefore asleep at some times, if the ten minute availability is assured. It is preferable to locate those doing the operating experience assessment on-site. The desired exposure to the operating plant and contact with the STA (if these functions are to be split) may be able to be accomplished by a group, normally stationed offsite, with frequent onsite presence. We do not intend, at this time, to specify or advocate a minimum time onsite.
- 4. The implementation schedule for the STA requirements is to have the STA on duty by January 1, 1980, and to have STAS, who have all completed training requirements, on duty by January 1, 1981. While minimum training requirements have not been specified for January 1, 1980, the STAS on duty by that time should enhance the accident and operating experience assessment function at the plant.
BWR CWNERS' GROUP DISCUSSION: Implementation of the Shift Technical Advisor (STA) as proposed _s by the Task Force would place a graduate engineer independent
) and detached from plant operations, in the control room at or shortly following the occurrence of an accident or abnormal transient. Because the STA would not be in the direct operational chain of command and, in fact, would not need to be licensed, he could neither manipulate nor direct licensed operators to manipulate the controls of the reactor plant. He would be empowered to advise operations but not responsible to operations for his advice.
The shift supervisor is correctly charged with the responsibility for safe operation of the plant at all times. During the early phase of an accident, he discharges this responsibility by coordi-nating and directing the response of the control-room staff. The actions of the operators are procedural, being governed by their training and emergency procedures, and during this phase the entire control room staff, including the shift supervisor, is completely occupied with responding to the accident. Plant operating experience indicates that there is a period of time following initiation of
- any accident or transient wherein the shift supervisor has sufficient time to analyze, diagnose, and respond to the condition of the plant but does not have sufficient time to carefully consider an inde-gendent assessment of the accident, resolve any conflicts between his and the independent asse;sment and, on the basis of such assess-ment, decide to alter the procedural actions of the operators.
p .. Q") 2.2.1.b-2
SNPS-1 RESPONSE TO NUREG 0578 Dialogue regarding such an assessment or time spent resolving such conflicts can only distract and delay the shift supervisor and consequently degrade the response of the control-room staff to the accident. Even though the roles of shift supervisor and STA can be care-fully delineated by procedure and training, industrial and military exaerience indicates that a direct-line organization wherein authority and responsibility are interdependent is re-quired to effectively operate in a crisis environment. The proposed STA is empowered to advise operations but not re-sponsible to operations for his advice. His authority and re-sponsibility are not interdependent'. A potential for conflict and confusion exists which cannot be completely eliminated by procedures or training because pro-cedure and training can address only those event sequences which have been postulated in advance. One important lesson learned from the experience at Three Mile Island and at other facilities is that not all event sequences can be postulated in advance. Therefore, an alternative which avoids this po-tential for conflict and. confusion but improves the functions intended by the proposed STA is recommended. Two functions are intended to be improved by the proposed STA:
) (1) accident assessment and (2) operating-experience assess-ment. In order to improve the accident-assessment function while avoiding the degradation in accident response which ac-companies the proposed STA, the course of an accident is con-sidered in three sequential phases: immediate, intermediate and recovery.
The immediate phase extends from the point at which an abnormal condition affecting plant safety can be detected in the control room until the point at which the shift supervisor has suf-ficient time to carefully consider an independent assessment and, on the basis of such assessment, decide to alter the pro-cedural actions of the operators. The intermediate phase ex-tends from the end of the immediate phase until the point at which the Technical Support Center (TSC) is manned and ready. The recovery phase extends from the end of the intermediate phase until the point at which recovery is complete. For the immediate phase, the accident-assessment function can be improved only by upgraded training to enhance the operators' abilities to recognize, diagnose, and respond to accident con-ditions. During this phase, the operators' actions are governed by training and emergency procedures, and by definition there is s insufficient time for the careful consideration of an independent " i assessment which would be required before such an assessment ' could become the basis for altering the procedural actions of the operators.
- 2. 2.1.b- 3
, i SNPS-1 ! RESPONSE TO NUREG 0578 l 1 For the intermediate phase, the accident assessment function can be improved by either of two alternative means. An operator can be educated in science and engineering in order that he might
, provide an assessment which could be considered and acted umon i ! by the shift supervisor. Alternatively, a graduate engineer or !
equivalent can be trained in plant operations and made available i , to the shift supervisor on call in order that he might provide t such an assessment. In either case, the shift supervisor must . have sufficient time to carefully consider the assessment and, l based on such assessment, decide to alter the procedural actions
- of the operators.
i ! For the recovery phase, the accident assessment function can be
- improved by manning the TSC. The collective engineering re-source within the TSC will be able to develop a detailed in-dependent assessment of plant conditions and provide appropriate procedures with which to recover from the accident.
l The operating experience assessment function can best be pro- ! vided by a team which reviews operating experience at the plant i and at plants of like design. Varying team membershi propriate to the operating experience being assessed,p,as ap-assures l accomplishment of this function by the best qualified individuals. BWR OWNERS' GROUP IMPLEMENTATION CRITERIA: The two functions intended to be improved by the proposed STA l will be improved as follows:
- 1. Accident Assessment
- a. Immediate Phases An operator or supervisor in the direct operational chain of command on each shift (normally in charge in the control room) will receive additional specific training in the response and analysis of the plant for transients and accidents. This training will be co-ordinated with the schedule for preparation and review of analysis and guidelines under the NRC Bulletins and Orders Task Force.
All other operators and supervisors will receive ad-ditional training appropriate to their responsibilities in the response of the plant to transients and ac-cidents. These longer term training and qualifica-tion criteria will be provided by the Institute of Nu- " clear Power Operations. - 2.2.1.b-4
d SNPS-1 RESPONSE TO NUREG 0578
, O
- b. Intermediate Phase (Alternatives) t An operator or su?ervisor in the direct operational chain i of command on each shift will receive substantial additional i
education in basic engineering and science sufficient to aid hLn in assessing unusual situations not explicitly covered in the current operator training. OR - A graduate engineer or equivalent trained in the response
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and analysis of the plant for transients and accidents and in plant design and layout, including the capabilities of instrumentation and controls in the control room, will be available to the individual in charge in the control room on call. He may be stationed on or off site as j appropriate to plant location, communication capabilities, op;rator training and education, extent and detail of erfrgancy procedures , etc. l c. Recovery Phase Individuals knowledgeable of and responsible for engineering and management support of reactor operations in the event of an accident will be available on call to staff the On-Site Technical Support Center. J
- 2. Operating Experience Assessment Where it does not already exist, a team will be designated by the licensee to assess the operating experience at his plant or plants and at plants of like design. Team member-ship may vary as appropriate to the operating experience being assessed but will include experience in systems engineering and familiarity with or routine access to persons experienced in the principles of human engineering or human factors.
LILCO RESPONSE: I LILCO endorses the BWR Owners' Group position. However, an on-shift technical advisor (STA) or alternate, in accordance with D, B. Vassa11o's letter dated October 10, 1979 (Enclosure 2, " Alternatives to Shift Technical Advisors," provided herein) will be provided for the Shoreham facility. In addition, other guidelines and criteria developed by O industry groups such as the Institute of Nuclear Power Operations will be evaluated. 2.2.1.b-5
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*EtiCLOSURE 2 ALTERtATIVES TO SHIFT TECH:lICAL ADVISORS The recommendation by the Lessons Learnec Task Force that an on-shift Technical Advisor be required at operating nuclear power plants has received much comment and attention by the ACRS and industry representatives since tiUREG-0578 was published. Several alternative approaches have been suggested. .
The ACRS has advised and the Director of tiRR has decided that alternatives be considered and approved if found by'the staff to satisfactorily accomplish the functions described by the Task Force for the Shift Technical Advisor. As an l aid to evaluating alternatives, a more ccmprehensive discussion of the purpose and basis of the Task Force recommendation is provided below. The discussion is in terms of the two principal functions intended to be accomplished and the characteristics thought to be necessary to effectively accomplish these functions. It is intended that the licensing review staff make use of this discussion in evaluating alternativas proposed by licensees and license applicants. Introduction . As stated in tiUREG-0578, the Lessons Learned Task Force has concluded that the need for improved operations is the most important lesson learned from the accioent at TMI-2. One key element sn far identified is the need to improve the capability in the control room to recognize and diagnose unusual events. Over the next several years, improvements in the capability of the reactor 1 . operations staff to respond to unusual events can and will be sought through improvements in plant design, operating procedures and the qualificaticn and training of operators. Improvements in plant design are expected to include improvements in the area of human factors, especially" improvements in display ' -N 9
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. 2 and diagnostic systems availacle to aid :: erat:rs. For example, the Task Force made a short term rec:mendation fer impr:vement of the means :f assessing inadequate core c:oling. The Task Force also mac'e shert term rc::m endaticns for improvements in emergency procedures and precarations by the plant iperations organizaticn. The purpose of these rec:=endatiens i: to assure that the operators and the onsite operational and technical support persennel are organized both administratively and physically in an effective manner. In addition, improvements in the licensing requirements for operators have been recomended to the Co mission. Over the coming months, it is likely that further increases in qualification and training requirements for operators will be developed by the industry's recently announced Nuclear Operations Institute for implementation over the next several years. Because these changes are necessary but difficult to achieve rapidly, the Lessons Learned Task Force has recomended the use of Shift Technical Advisors as a method of immediately improving the operating staff capabilities for response to off normal conditions and for Q
evaluating operating experience. The consensus of the Task Force is that there are two necessary improvements in the capability to assess the status of a plant during unusual conditions such as a transient or an accident, to realize the significance of the available information such as instrument readings, and to take appropriate action. First, there should be an accident assessment capability based on a comprehensive education in engin-eering and science subjects related to nuclear power plant design and on training and experience in the dynamic response of the specific ;'Isnt. This capability must be rapidly availaole in the control' room in the event of an accicent. Second, there should be a capability to maintain and upgrade safe plant operations thrcugh the cognizance and evaluation of applicable operating experience by an engineering O group with diverse technical knowledge, ex;erience, and ;erspective in relevant ..
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areas such as electrical, mechanical anc
' 3 fluic systems and human factors. The adoition of Shift Technical Advisors to g the plant coerating staff is an acceptable means of supplying both of these .
functions. Alternative manning and organizational schemes will be considered and will be evaluated for satisfaction of the qualifications, training and duty assignment criteria discussed below. Discussion In developing the recommendation for the Shift Technical Advisor, the Task Force concentrated dn the two functions that needed to be provided, namely, an accident assessment function and an operating experience assessment function. The proper performance of these functions requires the provision of certain characteristics described in the following paragraphs. A. Accident Assessment Function ( ,'
- l. General Technical Education
'- The technical education of at least one person in the centrol roem under off normal conditions should include basic subjects in engineering and science.
The purpose of this education is to aid the operator in assessing unusual situations not explicitly covered in the current operator training. The following is a tentative list of areas of knowledge that are considered to 'the desirable: Mathematics, including elementary calculus Reactor physics, chemistry and materials Reactor thermodynamics, fluid mechanics, and heat transfer Electrical engineering, including reactor control theory These areas of knowledge should be taught at the college level and would be equivalent to abcut 60 semester hours. Although a college gracuate engineer would have T.any of these subjects and more that would not be essential, scme
/'~' engineers might be deficient in a few of these specific areas, .,s e.g., reactor --
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4 physics. Althcugh the time te teach these subjects to a licensac senier rsacter cperator c:uld :e as sh:rt as 50 years, :e:encing en the se:pe sne centent of s the subjects, the selection of a graduate engineer wculd likely ce a m:re rapid means of fulfilling this characteristic.
- 2. Reactor Operations Training All persons assigned to duties in the control r::= shculd be trainec in the details of the design, function, arrangement and operation cf the plant systems. This training is necessary to assure that the meaning and significance of instrument readings and the effect of control actions are known. A licensed operator or supervisor of an operator w0uld not be required to have further training in order to fulfill this characteristic. A graduate engineer not previously licensed or trained as an operator or senior operator would require additional training in order to fulfill this characteristic.
( 3. Transient and Accident Response Trainimg
% In addition to the training in normal operations, anticipated transients. .
and accidents presently required of operators and senior operators, one persen in the control roca under off normal conditions should be trained to recognize and react to a wide range of unusual situations including multiple equipment failures and operator errors. This training should not be limited to written procedures or specific accident scenarios, but should include the recogniticn of symptoms of accident conditions such as complex transient responses or inadequate core cooling and possible corrective actions. The purpose of this training is to broaden the ability for prompt rec gnition of and response to unusual' events, not to modify the instinctive, rapid procedural response to l I transients and accidents provided by reactor operators. The training is recuired in recognition of the fact that real accicents inherently are initiated and s acccmpanied by unusual and unex:ectec events. The training is also to er:.Xsize -- n \ J -
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need to fccus on the essential parameters that indicate the stat.,s sf the core c. I ' I and the primary coolant bcundary. This additional training woul: take up to a year to accomplish for a persen not already experienced in nuclear plant transient and accident analysis or evaluation. Both inexpericnced gracuate engineers and currently licensed operators would require additional training to fulfill this characteristic.
- 4. Detachment from Operations The plant response assessment function requires a measure of detachment from the manipulation of controls or imediate supervision of operators. This is intended to provide the perspective and the time for assessing plant conditions and advising on appropriate operator actions. It has been called a safety monitor characteristic. Currently only three operators would normally be in the control room at the time an unusual event occurred, and it is allowed that at times there would be fewer. This number is only enough to satisfy the demands The for prompt control and supervisory actions under off normal conditiens.
time'necessary to make a considered assessment and permit independent monitoring of plant safety require one more person'in the form of the Shift Technical Advisor or some alternative in the control room.
- 5. Independence from Operations In order to provide both perspective in assessment of plant conditions and dedication co the safety of the plant, this function should have a clear measure of independence from duties associated with the cermercial cperation of the plant. In an accident situation where c0rmland authority snould nct be diluted, complete independence is not desirable and is not necessary to the safety assessment function.
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- 6. Availa:ility This :apactlity sneule te rea:ilj avaliable in :ne c:ntroi r::m, nitnin ten min;tes. Having prefers:1y imreciately at all times, cut at mes this capability en duty for each shift is the best appr:a:h.
B. Operating Experience Assessment Functicn
- 1. Indepencence frem Operations A measure of incependence is regJired to provide for effective safety I monitoring of Operating experience at the individual plant and at plants of like design. The assessment of operating experience at the assigned plant and l
other similar plants and the routine monitoring of the safety of plant operations . is usually compatible with and necessary for efficient operations. However, the demands of cc=mercial operation can sometimes distract from or appear to override safety judgments. An independent =enitoring of the safety of plant operations is ' intended to counter-balance the.icmediate and pressing needs of cc=mercial Operaticn. t
- 2. Dedicatien l
Personnel should be dedicated to the function of safety monitoring of operating experience as their primary responsibility and duty. Althougn reacter operating persennel have a commitment to safety that derives from self interest l as well as regulatory requirements, it is only one of two primary responsibilities , the other being the continuous procuction of ;ower. The assignment of safety evaluation of cperating excerience as a primary rescensibility for certain specified individuals will reduce potential c:nflicts and assure acecuate time to discharge the duties. . 1 i , as s_ l l
l
- 3. Diversity of Technical Knowlecge The technical knowledge of those assessing operating experience should be diverse and encompass all technical areas impcrtant to safety. The types of problem that can affect safety include all areas related to the cesign and operation of nuclear power plants; e.g., r.echanical, electrical and fluid systems and reactor physics, chemistry'and metallurgy. Recognition and under-standing of a problem and its significance requires scme kncwledge in the relevant technical specialities and cannot depend solely on the descriptions and judge-ments of the persons identifying and reporting the problem. Because of the broad scope of possible technical areas and the possible interactions of components, equipment and systems, the people engaged in operating experience review should have experience in areas usually designated as systems engineering.
In additien, because of They should also be graduate engineers, or equivalent. the importance of operator actions in the safety of plant operations, familiarity y with or routine access to persens with the principles of human engineering or human factors should be provid6d. Alternatives As discussed in NUREG-0578, several alternative means of providing the accident assessment function were considered by the Lessons Learned Task Force. They were:
- 1. Upgrade the requirements for reactor operators and senior reactor operators to include more engineering and plant response training.
- 2. Provide additional on-snift personnel with science or engineering training and specific traning in plant design and response.
Provide on-call assistance to the control room by identified 3.
;:ersonnel in the plant engineering organi:ation having the training ..
cescribed in alternathe 2.
3 Althougn the Tasa Force initially assumed that the accident assessment functicn woulc be ccebinec alth tne operating excertence assess ent Enctien, it is
' possible that the two functicns could be separatec. . h..e nave saggeste:
tnat people with the education, training, and experience recuirec for ccth the operating experience assessment function and tne safety tenitorir; functicn would be . ore easily obtained and retained if not requirec to acrk on shift. Others believe that such ;:ceple can be retainec if sufficient incentives are provided. The advantages and disadvantages of these alternatives are discussed
';elow. Althcugh no alternativa other than a group of dedicated Shift Technical Advisors has so far been found acceptable, it is possible tnat inncvative impreve-ments in the other alternatives could be found acceptable.
Discussion of Alternatives
- 1. Ucarade the trainino and cualifications of the senior reactor coerator.
This alternative would recuire no change in the cresent nu.t.ber or organization k of control room operators. The debilitating feature of this alternative is that the senior operator would be busy direc:ing the reactor operators or taking actions himself during an accioent rd no,t have sufficient time or perspective to make the desired assessment of plant conditions; f.e., perform the safety monitor function. This arrangement would also not provide a clear independence from cocreercial operation. However, the capability would be readily available when needed. It is unrealistic to expect the ser.ior Operator to fulfill the operating experience assessment function. A separate group could be established to accomplish that function on the day shift when interaction with offsite experts and utility management would be enhanced. If schemes are proposeo to acccmplish the two functicns separately, then they shoulc include c'echanisms f~ a ( .
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' .g.
,m for sufficient coupling of the two to assure continuous fee:back cf and ready f\ { access to the knowledge being acquired in ooerating. experience evaluation.
- 2. Additional on-shift eersonnel This alternative would require the addition of one person to the on-shift control room staff. If the person is to be a Shift Technical Adviser, no license would be required, thus making the position easier to fill quickly. However, detachment from first-line comercial operations decisions can be. attained by either a line or advisory position. For example, instead of the Shift Technical Advisor proposed by the Task Force, there may be acceptable methods of using a Shift Engineer, who no'rmally has authority over a Shift Supervisor., to perform the accident assessment function. Either approach would utilize people on shift so they would be readily available. Since the Shift Engineer would have normal duties other than operating exper'ience assessment, a separate day shift group would be required to fulfill that function if the shift engineer was found to be an acceptable source of the accident assessment (safety monitor) function.
- 3. On-call assistance This alternative would require no additional on-shift personnel. Others have susggested that provision of the recomended technical education and training would be most easily accomplished with this alternative since degreed engineers with intimate knowledge of the plant design basis and accident response character-istics are available in the utility-technical staff. Since these personnel would be remote from the control room, a requirement to be licensed does not appear to be consistent. Knowledg' e of accident response might also be more easily found among vendor personnel who have extensive experience'in accident analysis and systems design. This alternative also provides detachment frcm actual operation V and some independence from ccmmercial cperati:n. Mcwever,M5ese pecple would
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not be readily available when nected. The use of utility Or vender persennel {' not at the site woulo increase the difficulties of c:=unicati:n. Althcugh there is need for backup assistance frem these other organi:stions, it is doubtful that they would be able to provide for the prompt respense needs of l l the accident assessT.ent function and they do not have sufficient plant unique experience and familiarity to satisfy the operating experience assessment
' function.
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p_.__.-___._._. ._ ___.___ _ __ O. SNPS-1 RESPONSE TO NUREG 0578 2.2.1.c Shift and Relief Turnover Procedures NUREG 0578 POSITION: The licensees shall review and revise as necessary the plant pro-cedure for shift and relief turnover to assure the following:
- 1. A checklist shall be provided for the oncoming and off-going 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. A'ssurance 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 miti-gation of operational transients and accidents by a check of the control console (what to check and cri-teria for acceptable status shall be included on the i
j O c. checklist). Identification of systems and components that are in a degraded mode of operation permitted by the Techni-cal Specifications. For such systems and components, the length of time in the degraded mode shall be com-pared with the Technical Specifications action state-ment (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 tech-nicians. Such checklists or logs shall include any equipment under maintenance of test that by themselves could degrade a system critical to the prevention and mitigation of operational transients and accidents or initiate an operational transients (what to check and criteria for acceptable status shall be included on the checklist); and
- 3. A system shall be established to evaluate the effective-
- ! ness of the shift and relief turnover procedure (for l example, periodic independent verification of system alignments).
- ? 2.2.1.c-1
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SNPS-1 RESPONSE TO NUREG 0578 ( NRC CLARIFICATION: No clarification provided. BWR OWNERS' GROUP DISCUSSION: The Owners' Group agrees that knowledge of plant status, expecially for thos systems required to mitigate the consequences of an accident, should be transferred in a systematic manner from one shift to the next. The Group is also convinced that to be most effective as a means of information transfer in the course of a shift or relief turnover, the information must b'e limited to j that which can be summarized on a single list on a single piece
; of paper. Furthermore, the information provided by the list should be reviewed not only by the shift supervisor and control room operators, but by other plant personnel (auxiliary operators, technicians, etc.) as appropriate, thus eliminating the need for separate checklists, as apparently required in the Staff's position.
l l l BWR OWNERS' GROUP IMPLEMENTATION CRITERIA: l
- 1. A checklist will be devised to ensure that control room i status of systems that are required to mitigate the conse-quences of an accident are monitored on a shift turnover l
i basis. This list will include system lineups and alarms located in the main control room. Systems and components in a degraded condition will be identified as required by plant status.
- 2. The checklist will be kept in the control room at all times.
- 3. The checklist will be reviewed by personnel other than the shift supervisor and control room as appropriate.
LILCO'S RESPONSE: The Shoreham Nuclear Power Station has a procedure for operations staff shift relief turnover. This procedure will be reviewed and revised as necessary to assure that the above requirements are addressed.
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1 SNPS-1 RESPONSE TO NUREG 0578 2.2.2.a Control Room Access NUREG 0578 POSITION: l l 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 super-visor, shift supervisor, and control room operators), to tech-nical 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 estab-lished 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 manage-ment personnel not in direct command of operations, includ-i O ing those who report to stations outside of the control room.
NRC CLARIFICATION: No clarification provided. BWR OWNERS' GROUP DISCUSSION:
, The Owners' Group agrees that it is necessary to limit access to the control room and to establish a clear line of authority and responsibility in the control room in the event of an emergency.
i! sj BWR OWNERS' GROUP IMPLEMENTATION CRITERIA: Procedures will be develooed and imolemented which will meet the ' intent of the staff's pos'ition. 2.2.2.a-1 f! u __-..n .
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SNFS-1 RESPONSE TO NUREG 0578 O t LILCO'S RESPONSE: Appropriate procedures will be prepared for the Shoreham Station to limit access to the Control Room to those individuals re-
; sponsible for the operation of the plant such as Operations Super-visors, Watch Engineers, Control Room Operators; technical ad-l visors as requested and predesignated NRC personnel. The pro-cedure(s) will clearly establish the following:
a) The authority and responsibility of the person in charge of limiting access to the Control Room; b) The line of authority and responsibility in the Con-trol Room in the event of an emergency; and j c) The lines of communications and authority for plant ma-nagement personnel not in direct command of operations, including those who report to stations outside the Con-trol Room. -l
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RESPONSE TO NUREG 0578 ( A) 2 . Z'. 2 . b Onsite Technical Suoport Center NUREG 0578 POSITION: l Each operating nuclear power plant shall maintain an onsite tech- ! nical support center separate from and in close proximity to the l 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. plants as necessary to incorporate the role and location of the technical support center. 4 A complete set of as-built drawings and other records, as des-cribed in ANSI N45.2.9-1974, shall be properly stored and filed at the site and accessible to the technical support center under emergency conditions. These documents shall include, but not be limited to, general arrangement drawings, P& ids, piping system isometrics, electrical schematics, and photographs of components installed without layout specifications (e.g., field-run piping and instrument tubing) . (':) hus CLARIFICATION:
'l. By January 1, 1980, each licensee should meet items A-G that follow. Each licensee is encouraged to provide additional upgrading of the TSC (items 2-10) as soon as practical, but no later than January 1, 1981.
A. Establish a TSC and provide a complete description, B. Provide plans and procedures for engineering / management i support and staffing of the TSC, C. Install dedicated communications between the TSC and the ! control room, near site emergency operations center, and the NRC. Provide, between the TSC and the control room, a capability for the transmittal of some data. This requirement could be satisfied by closed circuit television l or process computer printout, D. Provide monitoring (either portable or permanent) for both direct radiation and airborne radioactive contaminants. The monitors should provide warning if the radiation levels in the support center are reaching potentially dangerous levels. The licensee should designate action levels to ~ Os define when protective measures should be taken (such as using breathing apparatus and potassium iodide tablets, or [-
- evacuation t' the control room),
2.2.2.b-1 __ _ _ Z_~n _ _ _ _
SNPS-1 RESPONSE TO NUREG 0578
\
J E. including Assimilate or ensure the licensee's accesstotohave best effort Technical directData, disp fay of plant parameters, necessary for assessment in the TSC, F. Develop procedures for performing this accident assess-ment function from the control room should the TSC become uninhabitable, and G. Submit to the NRC a longer range plan for upgrading the TSC to meen all requirements.
- 2. Location It is recommended that the TSC be located in close proximity to the control room to ease communications and access to technical information during an emergency. The center should be located onsite, i.e., within the plant security boundary.
The greater the distance from the CR, the more sonhisticated
^
and complete should be the communications and availability of technical information. Consideration should be given to providing key TSC personnel with a means for gaining access to the control room.
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- 3. Physical Size & Staffing The TSC should be large enough to house 25 persons, necessary engineering data and information displays (TV monitors ,
recorders, etc.). Each licensee should succify staffing levels and disciplines reporting to the TSC for emergencies of varying severity.
- 4. Activation The center should be activated in accordance with the " Alert" level as defined in the NRC document " Draft Emergency Action Level Guidelines, NUREG-0610" dated September, 1979, and currently out for public comment. Instrumentation in the TSC should be capable of providing displays of vital plant para-meters from the time the accident began (t = 0 defined as either reactor or turbine trip). The Shift Technical Advisor should be consulted on the " Notification of Unusual Event";
however, the activation of the TSC is discretionary for that class of event.
- 5. Instrumentation ,
The instrumentation to be located in the TSC need not meet safety-grade requirements but should be qualitatively compar- /~) able (as regards accuracy and reliability) to that in the (/ control room. The TSC should have the capability to access
- 2.2.2.b-2
SNPS-1 RESPONSE TO NUREG 0578 and display plant parameters independent from actions in the control room. Careful consideration should be given to the design of the interface of the TSC instrumentation to assure ; that addition of the TSC will not result in any degradation of the control room or other plant functions.
- 6. Instrumantation Power Supply The power supply to the TSC instrumentation need not meet safety-grade requirements, but should be reliable and of a quality compatible with the TSC instrumentation requirements.
To insure continuity of information at the TSC, the power supply provided should be continuous once the TSC is activated. Consideration should be given to avoid loss of stored data (e.g., plant computer) due to momentary loss of power or switching transients. If the power supply is provided from a plant safety-related power source, careful attention should be given to assure that the capability and reliability of the safety-related power source is not degraded as a result of this modification.
- 7. Technical Data Each licensee should establish the technical data requirements for the TSC, keeping in mind the accident assessment function that has been established for those persons reporting to the TSC during an emergency. As a minimum, data (historical in addition to current status) should be available to permit the
; assessment of:
Plant Safety Systems Parameters for:
- Reactor Coolant System
- Secondary System (PWRs)
- ECCS Systems
- Feedwater & Makeup Systems
- Containment l In-Plant Radiological Parameters for:
- Reactor Coolant System
- Containment
- Effluent Treatment
- Release Paths Offsite Radiological
- Meteorology
- Offsite Radiation Levels
- 8. Data Transmission l In addition to providing a data transmission link between the "
TSC and the control room, each licensee should review current t
. 2.2.2.b-3 Tr ~ J _ _ _ _ _ :T: T r_ .:r '~' : - ' T:::L L . . ~Li _ ^ ~T:; ' ;T
SNPS-1 RESPONSE TO NUREG 0578 (]-
\~- technology as regards transmission of those parameters identi-fied for TSC display.
Although there is not a requirement at the present time, each licensee should investigate the capability to transmit plant data offsite to the Emergency Operations Center, the NRC, the reactor vendor, etc.
- 9. Structural Integrity A. The TSC need not be designed to seismic Category I require-ments. The center should be well built in accordance with sound engineering practice with due consideration to the effects of natural phenomena that may occur at the site.
B. Since the center need not be designed to the same stringent requirements as the Control Room, each licensee should prepare a backup plan for responding to an emergency from
~
the control room.
- 10. Habitability The licensee should provide protection for the technical support center personnel from radiological hazards including direct radiation and airborne contaminants as per General Design Crit-erion 19 and SRP 6.4.
A. Licensee should assure that personnel inside the technical support center (TSC) will not receive doses in excess of those specified in GDC 19 and SRP 6.4 (i.e., 5 Rem whole body and 30 Eem to the thyroid for the duration of the accident). Major sources of radiation shculd be considered. B. Permanent monitoring systems should be provided to contin-uously indicate radiation dose rates and airborne radio-activity concen,trations inside the TSC. The monitoring systems should include local alarms to warn personnel of adverse conditions. Prpeedures must be provided which will specify appropriate proqective actions to be t~aken in the event that high dose rates or airborne radioactive concen-trations exist. '! C. Permanent ventilation systems which include carticulate
' and charcoal filters sh'uld o be provided. Th'e ventilation
- systems need not be qualificd as ESF systems. The design
.I and testing guidance of Regulatory Guide 1.52 should be followed except that the systems do not have to be redundant, seismic, instrumented in the control room or automatically activated. In addition, the HEPA filters need not be tested ~; _ . - ._Z T l.. 711 TT _ TT- . l~ T -~ ~ ~ _ ;Z _ _ . i_ '- _T ~ _ 1- ._
i SNPS-1 RESPONSE TO NUREG 0578 l l as specified in Regulatory Guide 1.52 and the HEPA's do not have to meet the QA requirements of Appendix B to
, 10 CFR 50. However, spare parts should be readily avail-able and procedures in place for replacing failed com-ponents during an accident. The systems should be de-signed to operate from the emergency power supply.
D. Dose reduction measures such as breathing apparatus and
- potassium iodide tablets cannot be used as a design basis i for the TSC in lieu of ventilation systems with charcoal 1
filters. However, potassium iodide and breathing ap- , paratus should be available. BWR OWNERS' GROUP DISCUSSION: The Owners' Group agrees that it is important to have a technical support center (TSC) desipated whdre " individuals who are know-ledgeable of and responsible for engineering and management sup-ports of reactor operations in the event of an accident" can go to, consistent with the intent to limit access to the control room. Furthermore, it is agreed that it is accropriate that the emergency plants will designate the role and location of the tech- , nical support center. There is, however, one area in particular which needs further discussion. The requirement that the TSC be onsite and in close proximity to the control room is not necessarily the best choice under all ,, circumstances for meeting the intent of the position. The lo-
+
cation of the TSC should be dictated by its accessibility to the
- engineering and matagement personnel who will occupy it, rather t
than b For example, multi y its physical oroximity to the control room. unit sites which share engineerin nel, or so-called outdoor sites which have administrative build-ings detached from the plant, may designate locations which may not be judged as in close proximity to the control room, but make sense from a per'onnel s access viewpoint. Furthermore, "elose l proximity" would only seem to be required as a means of supple-l; menting the transmittal of plant status from the control room to l the TSC, and in that sense then becomes inconsistent with the
! desire to limit access to the control room during emergencies.
Thus, the requirements for close proximity could be eliminated on the basis that the plant status must be monitored from the TSC. !! The Owners' Group also agrees that monitoring equipment may vary from plant to plant, and that there is no single best way in which' to monitor plant status in the TSC. There was agreement that TV monitors which could read and transmit information from , the control room panels to the TSC would meet the requirement to .j display and transmit plant status. It was also agreed that the . { TSC should have two-way communication links with the control room, other onsite telephones, the offsite Emergency Operations Center, t i 2.2.2.b-5
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SNPS-1 RESPONSE TO NUREG 0578 3 (d and the NRC. It was further agreed that the existing direct link between the NRC and the control room would be switched over to the TSC upon its activation in accordance with the intent to IL=it access to the control room. Finally, it was agreed that the staffing and activation criteria for the TSC would be specified in the emergency plan. BWR OWNERS' GROUP IMPLEMENTATION CRITERIA PHASE I (By January 1980)
- 1. A location will be designated in the e=ergency plan. This may be a temporary location.
- 2. Communication links will be established with the control, the onsite Operational ; Support Center, the offsite E=ergency Operations Center, and the NRC. These may be temporary.
- 3. The staffing and activation criteria will be specified in the emergency plan.
- 4. The TSC will have access to the records (system descriptions, arrangement drawings, etc.) in accordance with the revised NUREG 0578 position.
(/ The implementation criteria of Phase II will be issued after further discu.:sions between the Owners' Group and the NRC staff. LILCO'S RESPONSE: LILCO will provide a technical support center (TSC) onsite prior to fuel load. A location will be designated in the SNPS emer-gency plan. The TSC will have com.mnication links with the con-trol room, the onsite Operation Sngport Center, the offsite Emer-gency Operations Facility, and t he NRC. The TSC staffing and activation criteria will be specified in the SNPS emergency plan. The TSC will also have access to system descriptions, arrange-l ment drawings, and other plant records in accordance with the
- Staff's position. For a description of the conceptual design i currently being implemented refer to Appendix A, enclosed herein.
This information was previously submitted to the NRC via LILCO letter SNRC-486 from J. P. Novarro to H. Denton, dated July 21, 1980. xj 2.2.2.b-6
. APPENDIX A
, J.O. No. 11600.02 July 22, 19o0 i 1 i i
/
I i- i e i I DESIGN CRITERIA AND DESCRIPTION i TACHNICAL SUPPORT CANTER SBOREHAM KUCLdAR POWBR STATION - UNIT 1 IDNG ISLAND LIGHTING COMPANY i l l t 'I , 1 + i 6 t
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- . . - . - . - . . - ~ , - . - - . _ . . . . _ . . . - , - - - - . - - - - - - . , - - - - , . - . - . _ - . . - _ - _ _ _ _ . - . - . - - . . - - . . ~ ~ . . . _ - - - - . - . . - - . . . - - - _ - -
+. .
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- TABLE OF COtTh.NTS
- 1. GENERAL CRITritIA AND DESCRIPTIOu 1.1 General Criteria 1.2 General Description
- 2. DESIGN CRITERIA AND DESCRIPTION 2.1 Location / Space 2.2 Structural / Architectural 2.3 Habitan111ty 2.4 Heating Ventilation and Air Concitioning 2.5 Instrumentation 2.0 Electrical Power Supply 2.7 Communication 2.8 Records
- 3. ATTACHMr NTS
- 1. TSC X/Q Calculations Technique
- 2. TSC Integraten Dose Calculaticns
- 3. TdC Log and historical Data File - Inglant Syster.
Parameters
- 4. TSC Log and Historical Datia File - Itaciological/heteur-oloy1 cal Paran.eters N 4. FIGURES
- 1. Site Arrangement Plan
- 2. Secona Floor Plan - rxisting Security Bullaing
- 3. hvAC Senematic Diagram l
l .. i I O 1 I __ l: 11. . . . _ . __ _ ____ _..___ __... ._ .
- ~ ~ ~ ~
l t , -l 1.0 GENERAL CRITERIA AND DESCRIPTION 1.1 ( General Criteria A separate Technical Support Center (TSC) shall be orovided for i use by plant managem ent, technical, and engineering support i personnel. In an emergency, this center shall be used for and potential offsite impact in assessment of plant status support of the control room command and control function. The center should also be used in conjunction with implementation of onsite and offsite emergency plans, including communicaticns with an offsite emergency response center. Provide at the onsite Technical Support Center the as-built drawings of general plant arrangements and piping, instrumentation, and electrical systems. Photographs of as-built system layouts and locations are an acceptable method of satisfying some of these needs. 1.2 General Descriotion The second floor of the security building will be upgraded to serve as the TSC by the addition of filtered ventilation, computer generated system and radiological parameter displays and a backup power supply. The TSC staffing and activation criteria and interaction with the Emergency Operations Facilities will be specified in the Shoreham Nuclear Power Station - Unit 1 (SNPS-1) Energency Plan. The TSC will be operational by fuel load. l 2.0 DESIGN CRITERIA AND DESCRIPTION 2.1 Location / Space l '! 2.1.1 Criteria i The TSC shall be located in proximity to but separate from the i control room, and within the plant security boundary. The
; f acil,.ty shall be of sufficient si=e to accommodate those operating the TSC, NRC, and vendor represent &tives as well as the required equipment and technical data.
2.1.2 Description The existing security building is a separate structure located on the north side of the plant, as shown on the Site Arrangement Plan, Figure 1. The entire second floor of approximately 4,000 sq f t consisting of lecture and classrooms, an office, i library and toilets will be made available as the TSC on a joins basis. The first floor will continue a3 the security facility although it will be within the protected (habitable) environment provided for the entire building due to the TSC requirements. 1. I- ._-
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i The existing floor plan is shown on Figure 2. It will provide ample space for 25 people.
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2.2 Structural / Architectural 2.2.1' Criteria I The TSC need not be designed to seismic Category I requirements. l It shall be well built in accordance with sound engineerinq
! practice, with due consideration to the effects or natural j phenomena which may occur at the site.
2.2.2 Description The existing security building will be modified as necessary to accommodate the functions of a TSC.
- 2. 2 .2 .1 Existing Structure The security building superstructure is of steel framed construct-ion supported on reinforced concrete spread footings.
The. energy efficient curtain wall design utilizes insulated cavity wall construction. The roof deck and intermediate ficor slab are of reinforced concrete construction, with the roofing i material comprised of insulated, built-up asphalt and gravel. 2.2.2.2 Building Modifications *
,O The existing roof level HVAC penthouse will be expanded to
/ accommodate additional mechanical equipment. This penthouse expansion will be of a similar construction as the existing
. security building and will complement the existing architectural j style.
Additional building modifications will include the architectural r , sealing of the building to develop the ability to sustain the psitive internal pressure required for TSC occupation.
, This will be accomplished by providing existing doors and frames with appropriate weather stripping and gaskets. Should such modifications provide inadequate reduction of air leakage, the interior walls will be coated with an impermeable coacing system where necessary.
2.3 Habitability 2.3.1 Criteria The TSC shall be designed to protect personnel from radiological hazards including direct radiation and airborne contaminants in accordance with General Design Criterion 19 and Standard Review Plan 6.r4. Limits of S rem whole body, 30 rem tnyroid, snall not O- 2. l 1 w - w -,.rv 3 -g- . - - ----, -u----- -- - - .- , -- --w.w- - - - - ---,7-- g--- - - - ,e
i be exceeded for the duration of the accident considering ma3or s sources of radiation. s_ Monitoring shall be provided for both direct radiation and airborne radioactive contaminants. The nonitors should provide warning if the radiation levels in the support center are reaching levels approaching the design limits. The licensee should designate action levels to define when protective measures should be taken (such as using breathing apparatus and potassium iodine tablets, or evacuation to the control room) . 2.3.2 Description The security building meets these criteria, as follows:
- 1. Credit is taken for mixed mode release; see Attachment 1 for justification, and
- 2. The TSC atmosphere is filtered through a Charcoal-HEPA i ; filter. See Attachment 2 for a discussion ot the analysis. This is achieved by upgrading the security
- building HVAC system as discussed in Section 2.'4 .
The 30 day integrated doses calculated based on the above are: Total 30 Day Integrated Dose (Rem) Thyroid Gamma Beta i Mixed Mode Release
& 95 percent Halogen Filter 16 0.348 3.75 i These are within the limits of General Design Criterion 19.
Local wall-mounted area radiation monitors will be provided to measure radioactivity within the TSC and a ventilation monitor I with an iodine cartridge will measure recirculated alrhorne levels. Action levels to define wnen protective measures should be taken (including evacuation) will be designated. 1i* 2.4 Heatinu, Ventilation, and Air Conditioning 2.4.1 Criteria Permanent ventilation systems, in'cluding particulate and charcoal filters, shall be provided. These systems need not be qualified as ESF systems. However, the design and testing gulcance of Regulatory Guide 1.52 shall be followed except that the systems need not be redundant, seismic, instrumented in the control room, or automatically activated. In addition, HEPA. charcoal filters need not be tested as specified in Regulatory Guide 1.52 nor meet i
; the QA requirements of 10CFR50 Appendix B. Spare parts shall De readily available. -
l O~J 3. I
'I 2.4.2 Dencription
/ To pressurize the security building a tmosphere , 2,000 to
(-)g 3,000 cfm of filtered outside air will be supplied to the building. Provision has been made in the sy, sten design for 3,000 cfm maximum outside air, with recirculation capanility of up.to 1,000 cfm. 1 A 3,000 cfm capacity C5arcoal-HEPA filter train with booster fan
' will be installed on the roof of the security building inside an extension of the existing equipment room. This filter train will remove, with 95 percent efficiency, the gaseous iodine, methyl ~
iodine, and any particulates from the outside air, recucing concentrations to within acceptable limits. In order to use the outside air of 2,000 to 3,000 cfm for pressurization only, exhaust from the second floor lecture hall, toilets, and locker area will be eliminatec by shutting down roof fans and securely closing dampers. In addition, the main exhaust damper will be closed securely. Procedures shall be provided to
<j ensure all necessary actions are completed upon manning the TSC. 'I i
Existing system controls.will be modified to suit tne new design requirements and to maintain positive pressure following a DBA. A central control center will be provided for remote manual operation of the HVAC system during an accident. This will include push buttons for all the manual-controlled, power operated dampers, startup of the filter Ecoster fan, and direct expansion air conditioning.
't A conceptual study sketch (Figure 3) showing a sche.mtic of the existing security building HVAC system and proposed, modifications is attached. ' 'I Provisions for a direct expansion (DX) refrigeration system have
- been made to meet the requirements of heat gains due to
- t equipment, outside air, lights and power, and personnel occupany, and loss of office and service building chilled water supply.
During an accident, the TSC HVAC will be self contained and.its power source will be from a backup power supply as discussen in Section 2.6. , 1
! Regulatory Guide 1.52, Design Testing and Maintenance Criteria 'l for Atmospheric cleanup System Air' Piltration and Adsorption Units, will be followed as required to meet the criteria as i
stated in Section 2.4.1. Spare parts will be readily available l as will procedures for replacing failed components. i 4. 3 , , _ _ ,__ - c., .,._w. , _ , , - - - -.-.- -
'l 2.5 Instrumentation 2.5.1 Criteria !
d The TSC shall have tne capability to display plant parameters and eauipment status to technical and management cersonnel l responsible for engineering and support of reactor ooerations ! (control room activities) following an accident. The TSC capability to assess plant parameters shall be indeper. dent from l actions in the control room. The TSC equipment is not ' required i to be safety grade or redundant. l l The data between the beginning of the accident (t=0 defined as initial event, e.g. , reactor scram or turbine trip) and the time of activation of the TSC shall not be lost and shall ce available at the TSC. The instrumentation in the TSC shall not degrade plant installed safety grade instrumentation and equipment. 2.5.2 Descriotion The TSC data display will be entirely computer based and uill be provided by way of an enhancement of the existing process computer (PCS) and digital radiation monitoring conputer (RMS) systems. Refer to FSAR Sections 7.5.1.6 and 7.5.2.7 on the process computer system and to Sections 11.4 and 12.3.4 on the digital radiation monitoring system. j y_ The selection of parameters will be based on capabilities to: j 1. Diagnose initial event / accident,
! 2. Evaluate performance of safety related systems,
- 3. Ensure that the plant is in a stable shutdcan condition following an accident,
! 4. Monitor offsite (portable) and onsite radiological data, and
- 5. Monitor meteorological data.
2.5.2.1 In-plant Svstem Parameters Presentation of in plant system parameters will be provided at the TSC by the process computer system. Data will be cresented by a color graphics CRT display with keyboard access. Three high speed typers will be provided for hard copy record. One typer, a KSR (Input / Output) type, will provide user demand request
, capabilities and will receive outputs from existing NSSS and Sd?
post-trip logs as well_ as slanificant in plant systen alarns. , The other two typers will be of the RO (Receive Only) type; one ..
~
l s j S.
- , ,r- ,
- ~
i will provide the same alarms being presented on the tcIn control room alarm typer and the other will provide output or the process
/ computer TSC data log (Section 2.5.2.3) .
I
\
The entire process computer system data base will thus be available, on demand, for TSC display. Attachm=nt 3 provides a listing of specific data points which will be available at the TSC as part of the PCS TSC historical data file and log (Section 2.5.2.3) . Additional data will be provided to the process computer to ensure plant safety parameter availacility at tne TSC. All Class 1S signals required at the TSC will be isolated prior to
~
input to the process computer to ensure that these signals are not jeopardized or degraded by the operation of, or failure of , the process computer system. Qualified isolation devices will, in these cases, be inserted into those required Class 1S circuits to provide this assurance. A new TSC termination cabinet will ne located in the relay room and will be designed as a central gathering point for all required additional date prior to process i computer system input. I 2.5.2.2 Radiological Data Presentation of in plant radiological parameters and meteorological data will be provided at the TSC by the digital radiation monitoring computer system. Data will be presented by
,, a color graphics CRT display with keyboard access. One nigh speed typer, a KSR (input / output) type will be provided for hard (s ,,) copy records by way of user demand requests and outputs frcm the i RMS TSC data log and historical file (Section 2.5.2.3) .
I i i The entire RMS computer system data base, including of t-site done calcuations, will be availatie, on demand, for TSC display. Attachment 4 provides a listing of speciric data points wnich
, will be avilable at the TSC as part of the PRS TSC historical l data file and log (Section 2.5.2.3) .
Additional data, including wide ranges en effluant c ccess i monitors, will be provided to the radiation monitorinc computer to ensure radiological and meteorological parameter availacility at the TSC. 2.5.2.3 TSC Loos and Historical Data Files Two logs. and historical data files will be provided, one cy way i of the process computer system (PCS) and the other by way of the radiation monitoring system (RMS) computer. Pre-event historical l data files and post-event logging of data will provide TSC 'i personnel with the capability to diagnose tne initiating event t and its radiological cons equences , as well as provide an immediate evaluatica of safety systems performance and plan: status. .. D - U 6.
2.5.2.3.1 TSC Loc / Historical File - In-clant Systen Parameters s The process co=puter TSC historical data log will initiate uncu receipt o: an external event signal (t=0) a printcut, en a Tsu typer, of tnose in plant systen pars eters ass gned to t.r : icg (Attach =ent 3) . The log will continue printtng out dcta (tne frequency of a data point printout will depena on its assigned scan rate) until =anually ter=inated. A 5 =tnute pre-event data file (i.e. , history) of these selected TSC log parameters will be stored by the process computer to be recalled to tne TSC, on demand, using the TSO F.SR typer. 2.5.2.3.2 TSC Loc /Mistorical File - Radiolo:: cal /Xeteorcitnical Parameters The RMS co= outer log and historical- files provided will be similar to the process co=puter TSC log and cata file (Sertion 2.5.2.3.1) thus providing both pre- and ptst svent radiological /=eteorological data record. 2.6 Electrical Power suco1v - 2.6.1 Criteria An emergenev. power s u c. c i v. shall be o.rovided for the per=anent ventilation system of the TSC. The power supply to tr.e TSC s instrurentation need not meet safety grade requir==snes, rut i shall be reliable and of a quality compatible with tne TSO s I instr"-antation require =ents. The power sutelv for
\_/ =*
- instru=entation shall te continuous once the TSC is activated.
2.6.2 Descriction r The security . building facilities are presenti.v suppliec frc= a 300 kVA 430-120/203 V transformer through an autceatic transfer switen wnien receives pcwer frc= buses 11C End 11C. Black power fro = an on-site diesel will be connected to the alternate side of the automatic transf er switch instead of the f eed ::a= nus 11C. The normal supply to the transf er switch still has access to both sources of offsite power by virtue of ne transfer senere en tne
, 4 kV switchgear and =anually . through the tie-breaker in the 490 V double-ended load center. This arrange =ent alicws access to two sources of offsite power and a diesel generater and will carry , existing and added FVAC, lighting and other necessary loacs.
I 1 The power require =ents for the added co=puter por.ts and the [ peripheral equipment in -the TSC will be-- suppliec :rc= :ne-i exis ting ec=puter inve rter which is connected to the safety diesels. The inverter serves as an isolation device so thct the computer does not have to be tripped on a IOCA si 3 nal. Power for
, power supplies associated with isolation devices f c-r e c-kl .I inutru=entation will be fed froc the appropriate sare*., relate. . ! buses. ' /'-
(o 7.
( - 2.7 Communications 2.7.1 Criteria Comnunication links shall be established with the control room, the onsite Operational- Support C enter , the offsite R:aergency j Operations Facility, and the NRC. 2.7.2 Des criotion 1 Bell System phone lines will be used for communication with the NRC, the onsite Operational Support Center, and the offsite i Emergency Operations Facility, with optional lines to the Duclear Steam Supplier and the Architect Engineer. An existing line to NAWAS will be available. Communica tion to the control room will be by page/ party, and the ( plant pbx phone systems. 2.8 Records 2.6.1 Criteria l A complete set of as-built drawings and other records, as
! described in ANSI N45.2.9-1974, shall be properly stored and filed at the site and accessible to the TSC under e:aergency l conditions. These documents shall include, but not be limited to, general arrangement drawings, PGIDs, piping system
) isometrics, electrical senematics, and photographs of ccmponents installed without layout specifications (e.g. , field-run piping and instrument tubing) .
2.8.2 Description Critical documents sucn as Emergency Procedures, System Descriptions, and General Arrangement, Flow, Logic and Elementary l Schematic Drawings will be available in the TSC and the balance j will be availaole in the plant Records Center. Il i l l Cr V s. l
- - - __ _ _ - - - . - , - . . - _ , - . , . _ . - . , , , , , ,c-.-y, - ,-_ _-
ATTACHMENT 1 [,) I G SECURITY BUILDING , f I TSC X/O CALCULATIONAL TECHNIQUE -i Murphy and Campe identifies the technique that is to be utilized to evaluate X/Q values to be used in plant habitability calculations (see Standard Review Plan 6.4). The technique identified applies to a Design Basis Accident (DBA) release emanating from some wall of the containment struc ture . Historically, DBA X/Q calculational techniques have been conservatively limited to ground level release criteria exceot for releases from stacks 2 1/2 times the height of the ne: trest adjacent building (see Regulatory Guide (RG) 1.145). The release from the Shoreham DBA is unique in that the reactor building standby ventilation vent fulfills all seismic criteria. Instead of the release. leaking through portions of the primary and secondary containments, it is confinec to er.it through the vertical vent atop the secondary containment structure. This vent is higher than any adjacent building in the plant. Thus, a l more appropriate approach to consider X/Q calculation would be to
! utilize the mixed-mode release concept identirled in RG 1.111, Revision 1, Position C2b. To this end, the governing hurphy and Campe equation can be married with tne RG 1.111 Position C2b b concept (which was developed sponsored by the Atomic Industrial Forum at Millstone) to produce from at:aospheric tracer tests the following working equation:
- h -
f (I~ET)E*P ~ I I" E.f Z x/Q -
-. +
U (;to ,o,+ ) Urto ,c_ Where:
- ET = entrainment coefficient ET
- 1 fOr U ~u o 5 l ET =
2.58-1.58 (W eu - ) for 1 < wc ] s 1.5 V A-1 I _ . . ~ . _ _ _ . . _ _ _. _
,ee--.*w s
- l .
r ET = 0.3-0.06 (Wa g ) for 1.5 < il G5 o 5.0 i ET
= 0 for ving > 5.0 u =
wind speed at 10-m level (Wsec) oy = horizontal dispersion coefficient (m)
't 1 i o, =
vertical dispersion coefficient gn)
; A =
containment building aren (m2)
'j K = 3 '
(S/d)a,+ where S = source to receptor distance (m) d = containment diameter (m) .) h, = etfective stack height (m)
~.
where: he = hs +hp r -ht h3 = heicht of vent ralease (m) hr p = nonbuoyant plume rise (m) ht = height of TSC roof above plant grade (:a) W, = stack exit velocity (m/sec) e A-2
~ * ^
. . ...-,-, . . . - , . . , N , , , . ,-n._-..- .-~ , , , - . . . , , - . . . - . . , .. , , . . . , , , . _ . , . _ . - - , . . , - _ , , . , , . , . , , . , - - - . . , . - , , , - , - , . . ,.
I I It is conservatively assumed that tne 10-m wind speed applies to (N
\
the elevated portion ot the release. RG 1.145 also identifies a fumigation condition ac 11.ating for elevated (or partially elevated) releases durin- accicent conditions. Seabreeze fumigation occurs only wnen tne winds are
, blowing onshore. Examination of the relative locations of the snoreline, containment structure, and TSC, clearly shcws the fumigation frcm the containment can only occur in the opposite i direction from the TSC. Thus, this condition yields a zero X/Q at the TSC.
i
. The final consideration is to identify a 5 percent worst condition for this type of release. In order to introduce more conservatism into the calculational techniue, the meteorological condition producing the highest (worst) X/Q value
- (i.e. 0.01 percent) was assumed to cccur for the first 8 hr cf
'. the accident (0-8 hr period) . J.n additional ecn.t erva tism, nonbuoyant plume rise, due to the momentum of the release, was l presumed to be zero, even though RG 1.111, Revision 1 recommends its consideration. i i The X/Q's shown below were determined as being the highest (worst). values for the TSC. For the mixed-node release , the O-j 8-hr X/Q value maximized during Pasquill stability Class O eteutral) with a wind speed of 10.73 m/sec. For the ground release scenario, the 5 percent X/Q recu':ci frc:. a Pasquill . ['es t stability Class F (stable) and a wind speed of 1 m/sec.
\
Security Buildinc* 0-8 Hr 9-24 Hr 1-4 Day 4-30 Dav Mixed Mode Release X/Q (sec/m3) 8.58x10-5 5.32x10-5 1.39x10-5 3.u3x10-6 Ground Release j X/Q (sec/m3) 1.02x10-3 6.32x10-* 2.24x10-* 4.38x10-5
*Diatance 107 m from source to receptor k.
S A-3 l l
l i . B ATTACHMSMT 2 l
/~
(h) , SECURITY BUILDTUG TSC INTEGRATED DOSE CALCULATION l I ,, Regulatory Guide 1.3 identifies the technique that is to be => utilized to evaluate the integrated dose. The TSC integrated i dose analysis was done based on a LOCA release f rom the primary l containment at a rate of .5 percent volume per day, 10 gph ECCS 1eakage into the secondary containasnt, and MSIV leakage
, corresponding to a Technical Specification value of 11.5 scfh per i' valve. All releases are discharged via the RBSVS System.
The thyroid doses are computed using the conversion farters given in TID 14844 and a Dreathing rate of 3.47x10-4 m3/sec (1.25 m3/hr) . The gamma dosos are computed based on a finite cloud model in the TSC plus a semi-infinite cloud surrounding the building which has an equivalent 4 inch concrete structure. The beta doses are based on the simi-infinite cloud model suggested by the NRC, Regulatory Guide 1.3. The total 30-day integrated LOCA doses from the airborne activity l in the TSC plus gamma penetrating the building are indicated below. The doses are calculated based on mi::ed .no5e r21 esses s with atmospheric dispersion factors (Z/Q 's) as descriced in Attachment 1 and providing a HEPA-cnarcoal HVAC system as delineated in Section 2.4.2 of the TSC Design Criteria and Description. Total 30-Cav Incaarated Doce Jr.em) Thyroid Gamma Beta t Mixed-Mode Release 95% Halogen Filter 16. 1.73 3.70 No filter 319. 1.74 3.75
!U l
L __ __ - . _ , , , _ _ _ . _ - _
ATTACH '.3NT 3 y ) TSC LOG AMD HISTOR! CAL DATA FILE IN-PLA!!T SISTE11 PA?A. *.:.Ti.kS
- 1.0 Core Parameters '
- 1. Control Rod Postion (Core map graphic display)
- 2. Neutron Flux Levels (APRM, TIP) 1.1 Reactor Coolant System Parameters *
- 1. Reactor pressure
- 2. Reactor water level
- 3. Safety and relief valve position 1.2 Power Conversion System Parameters
- 1. Feedwater flow
- 2. Feedwater temperature
- 3. Condensate storage tank level
- 4. Main condenser pressure i
S. Circulating water pumps disch. pressure e I i 1.3 Safety System Parameters *
- 1. RCIC pump disch. flow
) 2. RHR system flow
_ ,/ 3. RHR HX inlet / outlet temperatures
- 4. HPCI pump disch. ficw
- 5. Core spray system flow
- 6. RHR HX - RER SW outlet temperature
, 7. RHR EX - RHR SW flow
- 8. RSCLCW EX outlet temperatures
- 9. R3 flood level 1.4 Containment Parameters *
- 1. Drywell pressure
- 2. Drywell temperature
- 3. Suppression chamber pressure l
- 4. Suppression pool water temperature
- 5. Suppression pool water level
- 6. Drywell hydrocen conc.
- 7. Suppressicn hydrogen conc.
- d. Drywell oxygen conc.
- 9. Suppression chamber oxygen conc.
- 10. Reactor Bldg. pressure 1.5 Service Air
- 1. ADS air header pressure ..
- hecun int c4:nal.s in tne.:e e?.te raries wil. be previfed in t.ua I \ log.
l .G-1 I _ _ _ _ _ _ _ _ _ . , _ . _ _ . . -~ - - - - - . - - -_ _
.c-, .- .--
~
t ' ATTACHME!!T 4 TSC LOG AND HISTOP.ICAL DATA FII E PADIOLOGICAL/ METR.OROLOGICAL PARA *IE""ERS Meteorological Parameters
- 1. Wind direction
- 2. Wind speed
- 3. Temperature 10 meters elevation
- 4. Vertical temp difference between 10 meters and upper levels
, Radiological Parameters
- 1. Main steam line radiation level
- 2. Containment area radiati n high range
- 3. RHR service water discharge radioactivity
.; 4. RBCLCW system radioactivity level
- 5. Control room ventilation activity level
- 6. Radiation. levels in essential equipment areas i 7. Release paths activity (Station vent exhaust and RBSVS)
; 8. Gaseous effluent flow rates A
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- 2. 4cc co.uirinc..ar in C To se. nooe.o 3.. Roor enHaost pampa. ras may sta4Lo.scer. uren tr.apW4 to sepoca s.e4mse FIGURE 3 HVAC SCHEMATIC DIAGRAM
.seas,. voi.uma mor ed- Secomi3,soii.o.~4 - tosiooo
- 4. -
ac ee n.ro c v1 TECHNICAL SUPPORT CENTER SHOREHAM NOCLEAR POWER STATION-UNIT t i 4
i SNPS-1
) RESPONSE TO NUREG 0578 J
s 2.2.2.c Onsite Operational Support Center NUREG 0578 POSITION: An area to be designated as the onsite operationai support center shall be established. It shall be separate from the control room and shall be the place to which the operations support personnel will report in an emergency situation. Communications with the control room shall be provided. The emergency plan shall be revised to reflect the existence of the center and to establish the lines of communications and management. i
; NRC CLARIFICATION:
No clarification provided. BWR OWNERS' GROUP DISCUSSION: The Owners' Group agrees with the position as stated, with the clarification that there may be plant unique situations where 4 it may be more appropriate that more than one location be designated in the emergency plan. As long as these locations are known and the " methods and lines of communication and
- management" are specified in the emergency plan, the intent
! of the position will have been met.
BWR OWNERS' GROUP IMPLEMENTATION CRITERIA: ll The Staff's position will be implemented as stated and subject
; to the clarification on location stated above.
LILCO'S RESPONSE: LILCO endorses the BWR Owners' Group position and will implement the Staff position as stated above prior to fuel load. i ** O 2.2.2.c-1 - k, s P i
SNPS-1 O O RESPONSE TO NUREG 0578 2.2.3 Revised Limiting Conditions for Operation of Nuclear Power Plants Based Upon Safety System Availability i NUREG 0578 POSITION: All NRC nuclear power plant licensees shall provide information to define a limiting operational condition based on a threshold of complete loss of safety function. Identification of a human or operational error that prevents or could prevent the ar.com-plishment of a safety function required by NRC regulations and analyzed in the license application shall require placement of the plant in a hot shutdown condition within 8 hours and in a i cold shutdown condition within 24 hours. The loss of operability of a safety function shall include con-sideration of the necessar electrical power sources, ycooling instrumentation, or seal water, controls, emergency lubrication, operating procedures, maintenance procedures, test procedures and operator interface with the system, which must also be capable of performing their auxiliary or supporting functions. The limiting conditions for operation shall define the minimum s_,) safety functions for modes 1, 2, 3, 4 and 5 operation. The limiting conditions of operation shall require the following:
- l. If the plant is critical, restore the safety function (if possible) and place the plant in a hot shutdown j condition within 8 hours.
- 2. Within 24 hours, bring the plant to cold shutdown.
1 3. Determine the cause of the loss of operability of the safety function. Organizational accountability for the loss of operability of the safety system shall be established.
- 4. Determine corrective actions and measures to prevent re-currence of the specific loss of operability for the particular safety function and generally for any safety
; function.
- 5. Report the event within 24 hours by telephone and con-firm by telegraph, mailgram, or facsimile transmission to the Director of the Regional Office. or his designee.
tj
- 2.2.3-1 1
o
. _ , , - - , , . , ,_ _ , - . , . ~ . . - . . - . . - . ~ . -
SNPS-1 RESPONSE TO NUREG 0578
)
Cx 6. Prepare and deliver a Special Report to the NRC's Director of Nuclear Reactor Regulation and to the Director of the appropriate regional office of the Office of Inspection and Enforcement. The report shall contain the results of steps 3 and 4, above, along with a basis for allowing the plant to return to power operation. The senior corporate execu-tive of the licensee responsible and accountable for safe plant operation shall deliver and discuss the contents of the report in a public meeting with the Office of Nuclear Reactor Regulation and the Office of Inspection and Enforce-ment at a location to be chosen by the Director of Nuclear Reactor Regulation.
- 7. A finding of adequacy of the licensee's Special Report by the Director of Nuclear Reactor Regulation will be required before the licensee returns the plant to power.
NRC CLARIFICATION: No clarification provided. BWR OWNERS' GROUP DISCUSSION: T None [G BWR OWNERS' GROUP IMPLEMENTATION CRITERIA: None LILCO'S RESPONSE: In accordance with NRC letter from D. B. Vassallo to all pending operating license applicants, dated September 27, 1979, the t proposed rule making on limited condition for operation of nuclear power plants has been delayed and no action is required at this time. t i e 2.2.3-2
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C _ . _ . _ _ _ _ _ _ .. p SNPS-1 RESPONSE TO NUREG 0578 Containment Pressure Indication i POSITION: A continuous indication of containment pressure shall be pro-vided in the control room. Measurement and indication ca-pability shall include three times the design pressure of the containment for concrete, four time,s the design pressure for steel, and minus five psig for all containments. The containment pressure measurements shall meet the design and qualification provisions of Regulatory Guide 1.97, including qualification, redundancy and testability. BWR OWNERS' GROUP DISCUSSION: The Owners' Group concurs with the ACRS recommendations for additional instrumentation for containment pressure monitoring. , 'i ( BWR OWNERS' GROUP IMPLEMENTATION CRITERIA:
- 1. The Owners' Group intends to implement containment pressure monitoring which will be designed and in-stalled to meet Engineered Safety System criteria.
, LILCO'S RESPONSE:
'I Currently installed instrumentation provides continuous indi-cation of containment pressure in the control room. The exist-ing pressure transmitters and associated instrumentation will be replaced in order to provide the capability to measure three times the design pressure of the primary containment. The range of the pressure instrumentation will be from -5 to +150 psig. The components provided will meet the design criteria outlined in the proposed Revision 2 to Regulatory Guide 1.97 to the max-imum extent possible.
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l! i; SNPS-1 RESPONSE TO NUREG 0578 4 Containment Hydrogen Monitors j, POSITION: A continuous indication of hydrogen concentration in the con-tainment atmosphere shall be provided in the control room. Measurement capability shall be provided over the range of 0 to 10% hydrogen concentration under both positive and negative ambient pressure. The containment hydrogen concentration measurements shall meet the design and qualification provisions of Regulatory Guide 1.97, including qualification, redundancy, and testability. BWR OWNERS' GROUP DISCUSSION: The Owners' Group concurs with the ACRS recommendations for ad- !; ditional instrumentation for containment hydrogen monitoring. It is the Owners' Gro'up's current interpretation that the h y-drogen monitoring requirement is associated with ECCS per formance and core degradation, rather than with containment at-nosphere control. I ii BWR OWNERS' GROUP IMPLEMENTATION CRITERIA:
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- 1. The BWR Owners' Grouo intends to implement containment i
hydrogen monitoring which will be designed and installed i to meet Engineered Safety System criteria. f LILCO'S RESPONSE: I The hydrogen concentration in the primary containment atmosphere will be continuously monitored by the hydrogen analysis system. This system consists of two redundant sub-systems, each including i two hydrogen analyzers to sample the drywell and the suopression 2 chamber atmospheres. Refer to FSAR Figure 6.2.5-1 included with
- Section 2.1.5.a and footnote on page 2.5.1.a-l. Each analyzer is
[ provided with dedicated instrument penetrations to ensure con-tinuous monitoring. The range of the analyzer will be from 0 to 10 percent hydrogen concentration by volume over a pressure range i of -2 to +60 psig. Monitoring units will be qualified for the ';I environment expected during normal and accident conditions. A dual recorder is currently installed in the main control room for each subsystem. These recorders are seismically qualified in accordance ! with IEEE-344-1971, QA Category I.and.in conformance with IEEE-323- - 1971. The hydrogen analysis system is powered from redundant emer- - i
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l i SNPS-1
. RESPONSE TO NUREG 0578 Containment Water Level Indication f
l POSITION: l 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. j The narrow range containment water level measurement instru-mentation shall be qualified to meet the requirements of Regulatory Guide 1.89 and shall be capable of being periodi-cally tested.
{rA BWR OWNERS' GROUP DISCUSSION: l\w The Owners' Group concurs with the ACRS recommendations for additional instrumentation for containment water level monitoring. For practical reasons, it is not desirable to monitor suppression , u pool water level all the way to the bottom of the s,ppression l pool. This is because an instrument tap at the very bottom could become obstructed by sludge and small debris. The 'l Owners' Group believes that water level monitoring down to the elevation of the lowest ECCS pump suction is more { practical and fully satisfies the intent of the requirement. i - BWR OWNERS' GROUP IMPLEMENTATION CRITERIA:
- 1. The BWR Owners' Group intends to implement containment water level monitoring which will be designed and installed to meet Engineered Safety System criterie.
- 2. The lowest suppression pool water level monitored will be at or below the elevation of the lowest ECCS pump suction.
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- i O LILCO'S RESPONSE:
; LILCO concurs with the Owners' Group position. Accordingly,
! taps to measure water level to the bottom of the suppression
, pool will not be provided. For Shoreham the lower limit will I remain unchanged at the elevation of the center line of the i ECCS suction lines containment penetrations.
In order to provide suppression pool water level measurement with an upper limit of 5 feet above the normal water level, the currently installed instrument taps will be relocated in order to increase the upper limit from 26'6" to 31'6". i f I I l t f l t l (
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! SNPS-1 h RESPONSE TO NUREG 0578 l>
?. Installation of Remotely Operated High Point Vents in the Reactor
} Coolant System '!;c 1; NRC POSITION (1):
i' Each applicant and licensee shall install reactor coolant system ' 2: and reactor vessel head high point vents remotely operated from
; the control room. Since these vents form a part or the reactor ,
'. coolant pressure boundary, the design of the vents shall conform ,: Criteria. In particular, these vents shall be safety grade, and ,, shall satisfy the single failure criterion and the requirements r
- t of IEEE-279 in order to ensure a low probability of inadvertent
'! actuation. ;
ji Each applicant and licensee shall provide the follewing informa- ; tion concerning the design and operation of these aigh point l: vents: . jj 1. A description of the construction, location, size and power 1 supply for the vents along with results of analyses of loss-of-coolant accidents initiated by a break in the vent pipe. 1 The results of the analyses should be demonstrated to be ac-captable in accordance with the acceptance criteria of 10 CFR 50.46.
- 2. Analyses demonstrating that the direct venting of noncondens- I able gaaes with perhaps high hydrogen concentrations does not "
! result in violation of combustible gas concentration limits l in containment as described in 10 CFR Part 50.44, Regulatory Guide 1.7 (Rev. 1), and Standard Review Plan Section 6.2.5. lj !! 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. BWR OWNERS' GROUP DISCUSSION: Domestic BWRs are provided with a number of power operated safety grade relief valves which can be manually operated from the con-trol room to vent the reactor pressure vessel. The point of con-l (1) Enclosure 4 to NRC letter, dated October 10, 1979, from -- (N D. B. Vassallo to all Licensees of Plants under Construction. _ L i
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I' nection of the vent lines from the vessel to these valves is such !; that accumulation of gases above that point in the vessel will not affect natural accumulation of gases of the reactor core. l These power operated relief valves satisfy the intent of the NRC 4 1 position. Information regarding the design, qualification, power
, source, etc., of these valves has been provided to the individual plant Safety Analyses Reports.
The Owners' position is that the requirement of single failure cri-teria for prevention of inadvertent actuation of these valves, and the requirement (stated in the October 11 topical meeting) that i power be removed during normal operation, are not applicable to BWRs. These valves serve an important function in mitigating the
- i effects of transients and in many plants provide ASME code over-
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pressure protection. Therefore, the addition of a second " block" valve to the vent lines could result in a less safe design and in
.: i some cases a violation of the code. Also, inadvertent opening of relief valve 4.n a BWR is a design basis event and is a control- ., lable transis it. (This is discussed in our position of NUREG-ji 0578, item 2.1.2.) '! In addition to the power-operated relief valves, operating BWRs include various other means of high-point venting. Information on ,i which plants are equipped with which features has been provided in
!! N individual plant Safety Analysis Reports, and may be summarized by '! s ,) individual licensees in their NUREG-0578 implementation letters. Among these are:
,, 1. Normally closed reactor vessel head vent valves, operable from the control room, which discharge to the drywell; 1-
- 2. Normally open reactor head vent line, which discharges to a main steam line; i
- 3. Main steam-driven Reactor Core Isolation Cooling (RCIC) Sys-tem turbines, operable from the control room, which exhaust to the suppression puol;
- 4. Main steam-driven High Pressure Coolant Injection (HPCI) Sys-tem turbines, operable from the control room, which exhaust to the suppression pool;
- 5. Isolation condenser primary side vent valves, operable frcm
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the control room, which discharge to containment or a main steam.line. Although the power-operated relief valves fully satisfy the in-tent of the requirement, these other means also provide protection against the accumulation of noncondensables in the reactor pres-f sure vessel. .. 7:X:L ~ :_ D -
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i e (" In the October 11, 1979, topical meeting on this subj ect, three (- procedural questions were raised:
- 1. Where to vent to (suppression pool vs. containment);
- 2. When to vent;
- 3. When not to vent.
Under most circumstances, there would be no choice as to where to vent to or when to vent, since the relief valves (as part of the Automatic Depressurization System), HPCI and RCIC will func-tion automatically in their designed modes to ensure adequate , core cooling, and these will provide continuous venting to the suppression pool. The current assessment is that it would not be desirable to interfere with emergency core cooling functions in order to prevent venting, but the matter will be studied further. The result of a break in the safety / relief valve discharge line, or any of the other systems enumerated above, would be the same as a small steam line break. A complete steam line break is part of the plants' design basis, and smaller-size breaks have been shown to be of lesser severity. A number of reactor sys-tem blowdowns due to stuck-open relief valves (also equivalent to a small steam line break) have confirmed this in practice (see [~s\ Owners' Group position on Requirement 2.1.2). Thus no new anal-i ( yses to show conformance with 10 CFR 50.46 are required. ! Because the relief valves, HPCI and RCIC will vent the reactor cont.inuously, and because containment hydrogen calculations in nornal safety analysis calculations assume continuous venting,
-; no special analyses are required to demonstrate "that the direct venting of noncondensable gases with perhaps high hydrogen con-centrations does not result in violation of combustible gas con-centration limits in containment".
BWR OWNERS' GROUP IMPLEMENTATION CRITERIA: i 1. The Owners' Group believes that adequate reactor coolant system venting is provided by the existing plant design.
- 2. Plant procedures will be provided to govern the operator's use of the relief valves for venting the reactor pressure vessel.
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- 3. No new 10 CFR 50.46 conformance calculations or containment combustible gas concentration calculations are required, l; since systems in the plant's original design and covered j by the original design bases are used. _
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- 4. In response to a request from the October 11, 1979 topical meeting, the use of isolation condenser tube side vents will be considered.
- 5. In response to a request from the October 11, 1979 topical meeting, the effect of noncondensables in HPCI/RCIC turbine steam will be addressed.
LILCO's RESPONSE: LILCOendorses(jheBWROwners'Groupposition. Presented below is a discussion 6f the features provided for Shoreham, which provide protection against the accumulation of noncondensables in the reactor pressure vessel.
- 1. Safety Relief Valves i The Shoreham facility is provided with eleven power operated i safety / relief valves (S/RVs) , which can be manually operated from the control room to depressurize (vent) the reactor pressure vessel (RPV). Seven of the eleven S/RVs comprise the automatic depressurization system (ADS) and are automatically actuated under certain conditions as described in Chapter 15 of the Final Safety Analysis Report. The S/RVs are connected to the four main steam lines which in turn are connected to
g the RPV above the fuel. Each S/RV discharge is piped to a quencher discharge device located at the bottom of the suppression pool. Position indication is provided in the main control room for each S/RV. The S/RVs, Steam Line and ADS are safety grade and conform with Appendix A to 10 CFR 50 General Design Criteria including the single failure criterion and the requirements of IEEE 279, as applicable.
- 2. Reactor Core Isolation Cooling (RCIC) and High Pressure Core Injection (HPCI) Systems The RCIC and HPCI, installed at Shoreham, are provided with
{ steam turbine driven pumps. The RCIC and HPCI turbines are ,i supplied with steam from the RPV through the main steam lines. The exhaust steam from these turbines is discharged l to the suppression pool. The equipment required for initiation of the RCIC and HPCI are completely independent of auxiliary A-C power; they require D-C power, derived from the station battery. These systems are automatically started upon a RPV low water level signal. Controls are provided for remote manual operation from the main control room. (1)This discussion includes the plant specific information re-
,_ quired in Attachment lE to the conference report of the BWR -
Owners'/NRC topic meeting on NUREG 0578 implementation, held - on October 11, 1979. _. _ _ . _ . . __ . _ T ~1 _ _ Z _ _ __ ~ ~ E _ _ ~ . _1
- 3. Normally Open Reactor Head Vent Line A normally open reactor head vent line is provided in the Shoreham design. This line discharges to one of the main steam lines and vents the portion of the RPV above the main steam nozzles. The head vent line is provided with a safety-related motor operated valve powered by an emergency bus and operable from the main control room. This line con-forms to the same design requirements as the reactor coolant pressure boundary.
We consider that the power operated S/RVs, as described in one (1) above, fully satisfy the intent of the reactor coolant sys-tem venting requirement. The alternative path of venting the RPV described in two (2) and three (3) above, however, provide ad-ditionally installed protection against the accumulation of non-condensable gasses in the reactor pressure vessel. Procedures for the proper operation of the S/RVs, RCIC and HPCI will specify operator action to vent the RPV.
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