ML20054M198
| ML20054M198 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 07/16/1981 |
| From: | Mccaffrey B LONG ISLAND LIGHTING CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0630, RTR-NUREG-0737, RTR-NUREG-630, RTR-NUREG-737, TASK-2.F.2, TASK-2.K.3.27, TASK-TM SNRC-595, NUDOCS 8207120087 | |
| Download: ML20054M198 (15) | |
Text
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LONG ISLAND LIGHTING COM PANY
,(/ECN@
SHOREHAM NUCLEAR POWER STATION
.... /a/ovu;su P.O. DOX G10. NORTH COUNTRY ROAD + WADING RIVER. N.Y.11792 July 16, 1981 SNRC-595 Mr. Ilarold R.
Denton, Director Office of nuclear Reactor Regulation U.S.
Nuclear Regulatory Commission Washington, D.C.
20555 SilOREIIAM NUCLEAR POWER STATION - Unit 1 Docket No. 50-322
Dear Mr. Denton:
Enclosed herewith are sixty (60) copies of LILCO responses to specific NRC ccncerns which were previously identified as requiring additional information to complete NRC review.
Attach-ment A provides a list of the specific responses included.
If you require additional information or clarification, please do not hesitate to contact this office.
Very truly yours, Ofigirial signca liv B.
R.
McCaffrey Manager, Project Engineering Shoreham Nuclear Power Station CC/mh Enclosures cc: J.
Iliggins bec:
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Youngling (w/ attach)
A.
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Pedersen Dist. List #14 (w/o attach)
Eng. File /SR2...A21.010 (w/ attach) f.fI]koh I
8207120087 810716 A
fff PDR ADOCK 05000322 E
PDR FC 4935
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Attachment A Additional information is provided for the following items:
1.
SER Open Item No.11 - Supplemental ECCS Calculations with.NUREG-0630 Model 2.
SER Open Item No.16 - Leak Testing of Pressure Isolation Valves 3.
SER Open Item No. 38 - Fracture Prevention of Containment Pressure Boundary 4.
SER Open Item No. 47 - Control System Failures 5.
SER Open Item No. 55 - Q-List 6.
SER Open Item No. 64 - Regulatory Guide 1.88 7.
NUREG-0737 Item II.B.7 - Analysis of Hydrogen Control 8.
NUREG-0737 Item II.F.2 - Identification of and Recovery from Conditions Leading to Inadequate Core Cooling 9.
NUREG-0737 Item II.K.3.27 - Provide Common Reference Level for Vessel Level Instrumentation A
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Item # 11 - Supplemental ECCS Calculations with NUREG-0630 Model The application of NUREG-0630 to the BWR LOCA model is a generic issue for all BWR's. GE and the' NRC met on July 8,1981 to establish a resolution approach and assure LOCA model conservatism and therefore, conformance to 10CFR50 Appendix K.
The NRC concluded that this issue would be finally resolved pending completion of a sensitivity study by GE to further demonstrate model conservatism.
Submittal of these results to the NRC is targeted for July 31, 1981.
It is expected that these studies will confirm the conservatism of the existing model (used in the Shoreham analysis) and that no impact upon the Shoreham LOCA analysis results is appropriate.
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S Item #16 - Leak Testing of Pressure' Isolation Valves t
LILCO will test each Category A valve which forms the interface with the ~RCS
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and low pressure systems to a leak rate acceptance criteria of 1.0 GPM water.
Refer to our previous response, SNRC-577 dated 5/27/81 for the definition of.
'.c the valves.
If alternative low pressure air or nitrogen tests are used in accordance with 10CFR50, Appendix J, then the methodology and basis for correlating the low pressure air or nitrogen tests to 1.0 gallon per minute at RCS pressure will be provided to the NRC for concurrence.
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Item #38 - Fracture Prevention of Containment Pressure Boundary The Shoreham Nuclear Power Station, Unit 1, is located approximately 41 N latitude where tiormal mean temperature for the worst winter month of January is +29.5 F.
Based on extreme values of temperature for Long Island, an extended normal extreme temperature of 0 F has been used as the design basis, which will be equalled or exceeded (higher temperature) for 99 percent of the time during the coldest three consecutive months.
Except for a portion of the walls (about 26 ft.) and the roof over the operating floor, the reactor building external surface is constructed of reinforced concrete :ith a minimum wall thickness of two feet.
Due to the high heat retention capacity and low leakage characteristic of the reactor building reinforced concrete walls, the inside air temperature is relatively unaffected by daily variations in outside air temperature when the reactor buildingnormalventilationsysteml/isnotoperating.
We have established on the basis of heat balance calculations that the reactor building interior temperature should be maintained at 65 F after postulated failure of the RBNVS heating system under the conditions cited by GDC 51.
Auxiliary space and local heating will be provided in the event that extreme temperature conditions prevail.
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3I efer to FSAR Section 9.4.2 for design details of the Reactor Building Normal R
Ventilation System (RBNVS) i i
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Item # 47 - Control System Failures Section 7.7 of the Shoreham SER requests that a review be conducted to identify any power sources or sensors which provide power or signals to two (2) or more control systems, and to, demonstrate that failures or s
malfunctions of these power sources or sensors will not result in consequences outside the bounds of the Chapter 15 analyses or beyong the capability
/
of operators or safety systems. '
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In order to respond to this request, a four (4) part program has been initiated as follows:
1 a.
Create a list of all control grade systems and indicate those whose failure may impact reactor pressure, reactor water level or critical power ratio.
b.
Identify any common power sources and common sensors from this list of s'ystems whose failure might impact the reactor.
c.
Postulate.. th'e failure of each bus with common Control Systems,
and sensors and define an event scenario.
'3 d.
Evaluate each scenario and determine if the existing Chapter 15' cvents are.either identical or bound the postulated bus failures.
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Completion of this ev'aluation and submit 1!al to the NRC is targeted for
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Septamber 30, 1981.
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r SNPS Item # 55 - Q-List (Rev. to SNRC-577, May 27,1981) k
- 10. HPCI and RCIC Initiation Levels Response: All changes to the HPCI and RCIC have the same classification as the systems in which they are located.
11.
Isolation of HPCI and RCIC Response: All changes to the HPCI and RCIC have the same classification as the systems in which they are located.
- 12. Challenges to and Failure of Relief Valves Response: There is no hardware change related to Item 12, thus, no modification to Table 3.2.1-1 is required.
- 13. ADS Actuation Response: There is no hardware change related to this item, thus, no modification to Table 3.2.1-1 is required, i
- 14. Restart of Core Spray and LPCI Response: There is no hardware change related to this item, thus, no modification to Table 3.2.1-1 is required.
- 15. RCIC Suction Response: The RCIC system will be modified to provide automatic switchover from the CST. The modification involves logic changes only and, as such, is included in Table 3.2.1 1 item XII part 10. No change to the table is required.
- 16. Space Cooling for HPCI and RCIC Response: There is no hardware change related to this item, thus, no modification to Table 3.2.1-1 is required.
17.
Power on Pump Seals Response: There is no hardware change related to this item, thus, no modification to Table 3.2.1-1 is required.
- 18. Common Reference Levels Response: There is no hardware change related to this item, thus, no modification to Table 3.2.1-1 is required.
- 19. ADS Valves, Accumulators and Associated Equipment and Instrumentation Response: There is no hardware change related to this item, thus, no modification to Table 3.2.1-1 is required.
SER 01 55 7/16/81 i
Item # 64 - Regulatory Guide 1.88 NRC Request The proposed exception to Regulatory Guide 1.88, August 1974, needs clarification.
It is not clear whether the exception to R.G. 1.88 and the 2-hour fire rating is intended to be applicable to the construction phase, operations phase, or both. Therefore, provide us with a description and/or alternatives to this Regulatory Guide and clearly indicate to which phase your description will be applicable.
LILC0 Position Suitable facilities for the storage of records which satisfy the NRC concern are described in Regularoty Guide 1.88.
Acceptable alternatives to the fire protection rated provisions of R.G.1.88 are given in Standard Review Plan NURES-75/087, Section 17.1.
LILC0 proposes a 2-hour fire rated l
records storage vault conforming to NFPA No. 232 Class 1 for permanent-type records. A 2-hour vault meeting NFPA No. 232 is recognized in SRP Section 17.4 as an acceptable alternative to R.G.1.88.
This alternative
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will be applicable to the construction and operational phase records.
FSAR Appendix 3B will be amended to clearly indicate the above, s
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II.B.7 Analysis of Hydrogen control NRC Position Certain LWR concainment structures having small volumes may have to be inerted to prevent their being overpressurized as a consequence of burning hydrogen during a severe accident involving extensive reaction between fuel cladding and reactor coolant.
Some containment strd'dtures, particularly those with a large volume and high design pressure, may not need inerting.
.In other containment structures, it.may be appro-priate to use features and procedures other than inerting to cope with the generation of hydrogen.
In Commission papers (SECY-80-107 and 80-107A), the staff discussed interim hydrogen control requirements for small containment structures, such as BWR Mark I and II, and the bases for continued operation and licensing of nuclear plants pending the rulemaking proceeding.
A rulemaking is being prepared that,'in part, will establish hydrogen
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control measures to deal with accident conditions involving large amounts of hydrogen generation in all types of containment structures.
LILCO Position Primary Containment Inerting System The Primary Containment Inerting System is designed to prevent the primary containment from being overpressurized as a consequence of burning hydrogen during a severe accident involving extensive reaction between fuel cladding and reactor coolant.
Design Basis The design bases for the Primary Containment Inerting System are:
1.
The system is nonsafety related, except where it ties into existing safety-related systems.
l 2.
The Primary Containment Inerting System is capable of reducing the oxygen concentration in the primary. containment drywell and suppression chamber from a normal concentration of about 21 percent to less than 4 percent (by volume) in 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.
j 3.
The system will maintain the oxygen concentration to less than 4 percent during normal operation 'oy allowing the addition of i
nitrogen as necessary to compensate for any increase in the t
l oxygen concentration.
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The Primary Contain=cnt Incrting System supplies all cas-operated valves in the primary containment, precludina the addition of amf oxygen to
.:e con:1iamant n augh val 1 oper-ation c le_%2gc.
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The system will be isolated from the primary containment' during normal operation except when used to lower oxygen concentration.
The system will be isolated whenever an accident signal is present by automatic closure of the containment isolation valves.
System Design
The Primary Containment Inerting System (Fig. 6.2.6-1) establish [s and maintains an oxygen deficient atmosphere ( 4 percent volume) in the primary containment during normal operation.
The system provides an increased amount of time (14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />) after a LOCA to start the hydrogen recombiners before the oxygen concentration reaches 5 percent.
The system interfaces and works in conjunction l
with the Primary Containment Atmospheric Control system (Fig.
6.2.5-1).
The system consists of the following subsystems:
1.
Main Inerting System - is used prior to or during startup to reduce the oxygen concentration from about 21 percent to less than 4 percent by volume.
Makeup Mitrog'n Supply System
.is used to supply nitrogen 2.
e during normal operatio'n to compensate for small increases in oxygen concentration over long periods of time.
3.
Nitrogen Supply to other systems - hitrogen will be used to supply gas-operated valves in the containment, post-accident sampling, TIP purge, and emergency containment purge supply.
Main Inerting System 4
Liquid Nitrogen stored in an 11,000 gallons storage vessel is-vapor-ized and heated in an electric vaporizer, regulated by a pressure /
temperature control manifold and delivered to the primary containment i
atmosphere at a rate of 1000 scfm, The inerting system piping.cbanects to the reactor building ventilation system piping, the suppression chamber, and to the combustible gas control system piping to the dry-well.
Each inerting supply line contains redundant primary contain-ment isolation valves.
In the event the liquid nitrogen stor' age vessel is out of service, an emergency connection is provided.
This connection allows a liquid nitrogen truck to bypass the storage vessel and supply liquid nitrogen directly to the electric vaporizer.
During the inerting procens the oxygen le.ci is acnitored by redundant, sa 'ty-related oxygen analyzers of the Prim'ry Corinnent ?.'~~c-baric r
c-31 systen.
SNPS-1 FSAR Makeup Nitrogen Supply System g
During normal operation, when the oxygen. concentration in the containment approaches 4 percent, nitrogen'ia added to the containment atmosphere through the makeup system.to lower the oxygen concentration.
i Nitrogen is supplied from the Liquid Nitrogen storage vessel, vaporir.ed in a set of ambient vaporizers, heated by an electric 3
heatet to 700F, regulated by a temperature / pressure control manifold and delivered to the containment at a rate of approxi-
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mately 100 scfm.
Nitrogen from the Primary Containment Inerting System also supplies other systems in part.
All. gas-operated valves in the primary containment are supplied with nitrogen to avoid oxygen leakage into the containment as a result of valve actuation and leakage.
Nitrogen from the Primary Containment Inerting System is also used for the following purposes:
1.
Supply for Emergency Containment Purge 2.
Supply for Traveling Incore Probe Blanket e
3.
Supply for Post Accident Sampling System Purge /Dilht/on
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4.
Supply for Integrated Leak Rate Test 5.
Supply for Drywell Floor Relief Valve Test s
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FIG. 6 2.5-1 PRIMARY CONTAINMENT ATMOSPHERIC CONTROL SYSTEM SHOREHAM NUCLEAR POWER STATION-UNIT 1 FINAL SAFETY ANALYSIS REPORT REVISION 22-JULY 1981
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i II.F.2 INSTRUMENTATION FOR DETECTION OF INADEQUATE CORE COOLING The NRC has found the applicant's response to be satisfactory except that emergency operating procedures have not been submitted and no commitment has been made to provide incore thermocouples.
The applicant commits to provide the emergency operating procedures by July 31, 1931. With respect to incore thermocouples, the applicant commits to implement the generic resolution of this issue between the BWR Owners Group and the NRC.
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II.K.3.27 Provide Coamon Reference Level for Vessel Level Instrumentation flRC Position Different reference points of the various reactor vessel water level instruments may cause operator confusion. Therefore, all level instruments should be referenced to the same point.
Either the bottom of the vessel or the top of the active fuel are reasonable reference points.
LILC0 Position tiUREG-0737, Item II.K.3.27, " Common Water Level Reference", requires that all reactor pressure vessel water level indicator scales be based on a common reference zero. The intent is to reduce a perceived potential for operator confusion due to the different reference points of the various reactor vessel water level instrun.ents.
Although LILCO believes the current BWR water level indication system is fully adequate to allow plant operators to respond properly under all postulated reactor conditions, a common water level reference will be provided.
The common reference will be implemented with minimal changes by modifying-the two fuel zone instrument indications to reflect the same reference zero (the bottom of the dryer skirt) as the other nine water level indications.
The instrument face plates will be changed to indicate a level consistent with the instrument zero used for all other reactor pressure vessel water level indications.
The use of this common reference will not change any existing trip signals, alarms, or control functions; changes to operator training and procedural modifications will be minimal. This common water level reference is a reasonable reference point as requested by the flRC and will eliminate any confusion due to different reference points.
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