ML20151K531

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Forwards Equipment Qualification Branch SER Input Re Seismic & Dynamic Qualification of Seismic Category I Mechanical & Electrical Equipment & Pump & Valve Operability Assurance Program
ML20151K531
Person / Time
Site: Nine Mile Point, 05000000
Issue date: 11/26/1984
From: Knight J
Office of Nuclear Reactor Regulation
To: Novak T
Office of Nuclear Reactor Regulation
Shared Package
ML20150F672 List:
References
FOIA-88-356 NUDOCS 8412050571
Download: ML20151K531 (10)
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b Docket No. 50-410 MEMORANDUM FOR: Thomas M. Novak, Assistar,t Director for Licensing Division of Licensing FROM: James P. Knight, Assistant Director Components & Structures Engineering Division of Enginecring

SUBJECT:

SER INPUT ON EQU!PMENT QUALIFICATION FOR NINE MILE POINT 2 Plant Name: Nine Mile Point 2 Docket No.: 50-410 Licensing Stage: OL Responsible Branch: Licensing Branch No. 2 Responsible Project Manager: M. Haughey Review Status: Continuing The enclosed Safety Evaluation Report (SER) input was prepared by DE:C&SE, Equipment Qualification Branch and covers the following items in the subject report:

(1) Seismic and Dynamic Qualification of Seirmic Category 1 Mechanical and Elcetrical Equipment.

(2) Purp and Valve Operability Assurance Program The Equipment Qualification Branch provides in the enclosure the current review status of each of the items presently assigned to them.

The adecuacy of the overall qualification program will be determined based upon an on-site plant audit. The seismic and dynamic qualification program audit (item 1) will be conducted by the Seismic Qualification Review Team (SQRT) and the pump and valve operability assurance program audit (item 2) will be conducted by the Purp and Valve Operability Revicw Team (PVORT). Wo IgqWP/abbp ploj e-g m

T. Novak NOV 2 6 1984.

plan to conduct the PVORT audit concurrent with the SQRT audit. These audits will be conducted when at least 85% of the safety related equipment is

! qualified, documented in an auditable manner, and installed in the plant.

I Original signed EY:

l Jnnes P. Fa'.Ct James P. Knight, Assistant Director Components & Structures Engineering

Division of Engineering

Contact:

SQRT N. Romney, X28115 PVORT R. Wright, X28209

Enclosure:

As stated cci V. Neonan A. Schwencer i

M. Haughey G. Bagchi J. Jackson i N. Remney R. Wright B. Miller, BNL 2 C. Hofmayer, BNL l

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Status of Review for Nine Mile Point 2 Oceket No. 50 410 l Seismic and Dynamic Leads Section

! Equipment Qualification Branch l October, 1984

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J (1) Seismic and Dynamic Qualification of Seismic Category ! Mechanical and j Electrical Equipment: SER input is attached. Review is continuing.

(2) Puep and Valve Operability Assurance Program: SER input is attached.

'leview is continuing.

) (3) ContainmentIsolationDependability(PartofTM!ActionItem!!.E.4.2):

] Applicant to provide detailed information on operability of containeent purge and vent valves. Review is open.

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] (4) Performance Testing of Relief and Safety Valves (Part of TM! Action Item

)  !!.0.1): Applicant's response to plant specific request for additional infor-j mation has been received ard will be reviewed. Review is continuing.

l l (5) Verify Qualification of Accumulators for Automatic Depressurizaticn System Valves (Part of TMI Action Item !!.K.3.28): FSAR material under staff review, Review is continuing.

(6) Long Term Operability of Deep Draft Pumps (!E Bulletin ?9-15): Under staff review, Review is continuing.

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1 Safety Evaluation Report Input Nine liile Point 2 Docket No. 50-410 Equipment Qualification Branch 3.10 Seismic and Dynamic Oualification of Seismic cateeory I !!ec~n anical and Electrical Ecuipment 3.10.1 Seismic and Dynamic Qualification The staff's evaluation of the adequacy of the applicant's progran for quali-fication of electrical and rechanical equipment important to safety (cr seismic and dynamic loads consists of (1) a determination of the acceptsbility of the procedures usedi standards follcwedi and the corpleteness and adecuacy of the inplemantation of the entire seismic and dynamic qualification progrsn.

The staff has reviewed the nethodology and procedures of the seismic and dynamic qualification program contained in the pertinent FSAR Sections 3.9.2, 3.9.3, and 3.10. The review of the seismic cualification program description included an assessment of th3 test and analysis methods and general criteria and guidelines used for testing and selection of representative er typical equipment. The staff finds that the equiprer,t qualification prcgram, except at described belew, meets the intent of the acceptance criteria specified in SRP Saction 3.10. To cceply with SRP Section 3.10 the applicant's qualification pregran mest rent the requirerents and recorrendations of IEEE 344-1975 (cr fer category 2 plants, identify and justify the use of other criteria), and the regulatory positions of RGs 1.61, 1.59, 1.92, and 1.100 to provide adequate assurance that such equipment will function properly under all icposed design and service loads including the leadings by the safe shutdcwn earthquake, postulated accidents, and less-of-ccclant accidents.

Cerplian e with these regulatcry guides was accceplished through an alternate approach which is ret fully described in the FSAR. Also becau:e this is a category 2 plant (CP application deckets befnre October 27, 1972), the Gineral E1cetric equipment has been qualified to earlier standard and guidelines that h4 1 -- .

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4. require supplementary support to sathfy NRC requirements. The applicant must
  • " clarify in an amendment to the FSAR tnis additional support and provide the information requested below for both the nuclear steam supply system and some y

of the balance of plant equipment as applicable.

(1) Applicant's Tables 3.10A-1 and 3.10.B-1 of the FSAR which are to be provided by the applicant in the third quarter of 1984 will be reviewed for completeness. These tables shou'd also reference all the important applicable standards which are met in the qualification process particularly those referenced in SRP Section 3.10.

(2) While it is recognized that qualification to IEEE 382-1972 was used, since NMP 2 in a category 2 plant, the applicant should clarify in the FSAR what additional guidelines are used to upgrade the cualification to reflect the considerations stated in SRP Section 3.10. Indicate what operability testing is included simultaneous with thermal aging and discuss fatigue censiderations for number of actuation cycles, since these are not addressed in the 1972 -

j standard.

(3) In the staff's r$ view of the HPCS diesel generate , it is not clear that the check valve upstream of the air receiver tank is qualified against well j established leakage criteria following a seismic event. The applicant sheuld amend the FSAR to clarify how sufficient are capacity is assured to start the HPCS diesel generator imediately after a ssismic event. )

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(4) While mounted components are described as qualified to acceleration levels consistent with those transmitted by their supporting structures, the applicant is to clarify in the FSAR how the equipment interactions with the mounting are addressed censidering actual deflections.

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(5) Although the applicant has comitted to follow the requirements and recomendations of IEEE 344-1975 and RG 1.100 for balance :f ,alant equipment, the methods of handling agir,g effect on seismic capabil'.ty of both electrical I

i and mechanical equipment should be clarified in an amendment to the FSAR. In addition, tne applicant should clarify how the General Electric generically qualified equipment to IEEE Std. 344 1971, without full use of RG 1.100 which is applicable to NKP-2. The applicant should also describe what additional l measures were taken to ensure adequate consideration was given to aging, sequential testing and upgrading of analytical methods as appropriate.

i (6) The applicant should comit within the FSAR to establish a maintenance and surveillance program to maintain equipment in a qualified status throughout I

plant life.

(7) The applicant should amend the FSAR dercription of its seismic qualification program and clarify the seismic margin for the required response spectra employed with respect to safety-related mechanical equipment. Also while qualification tests were perforned on cabinets, vertical bnards and benchboards for electrical equiprent, the applicant needs to clarify the methodology used to qualify multicabinet assemblies, particularly those too large to test.

(8) The applicant's description of testing of equipment to frequencies up to 33 Ha needs to be extended in the FSAR to address higher frequer..ies for the BCP equiprent. The applicant has cemitted within the FSAR to qualify the St.C.

RCIC and ECCS pump asse.blies, the ADS and main steam SRV accurulater systen to hydrodynanic loads. Hewever, the FSAR should be seended to cla.ify what additional safety-related rechanical equipment may see high frequencies from hydrodynamic loads. Describe in Section 3.10 the tests and analyses conducted en all such equierent to preperly include the envelope of input motion produced by hydredynamic loads.

The staff will follow the applicant's effort closely, and will ccnfirm its implementatien during the onsite audit. During the plant site at.dit, the staff will review in detail the applicant's implementation cf the qualification pregram to confirm that all applicable leads and combinations of 1Mds have

been defined, operability has been verified through appropriate tests and analyses, assemblies rather than individual components have been verified operable, and that for all safety related equipment, op?rability can be assured through.the plant life. Where static coefficent analysis was used to demonstrate seismic adequacy the staff will confirm compliance with R.G.1.100.

Approximately 85% of the equipment must be qualified, documented in er -

auditable manner, and installed on site before an ensite audit by the staff can be performed. When the applicant indicates that his work is a least GS%

complete, the staff will then conduct an onsite audit shortly thereafter. The staff will report the results of its audit and the followup and resolution of its concerns described above in a future supplement to the SER.

3.10.2 Pump and Yalve Operability Assurance The staff evaluation of the adequacy of the pump and valve operability assurance program consists of two parts. First a determination is made of the l completeness of the program with regard to the standards and guides used and )

the procedures used for progran implementation. This determination is based I upon the sufficiency of information in the FSAR and its supporting documents  !

which gives evidence that the applicant is following a disciplined and thorough I progran for pump and valva eperability assurance. Upon a satisfautory l evaluation of the FSAR information an entite audit of selected equipment is perforred. This secor.d part, the audit, is to develop the basis for the staff's judgment regarding the adequacy and ccrpleteness of irplementation of the entire program on pump and valve cperability assurance.

The pump and valve operability review team (PVORT) has reviewed the scope, methodelegy, and procedures of the pump and valve operability assurance progran described in FSAR Section 3.9.3 and selected parts of FSAR Sections 3.2 and i 3.9.2. l t

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h l' The information in the FSAR suggests cenpliance with the general intent o' the ,

I acceptance criteria as specified in the SRP Section 3.10. Based upcn the 1 comitments in the FSAR, the applicant's qualification program for balance of i plant equipment does meet the requirements and recomendations of IEEE Standard 323 1974 and IEEE Standard 344-1975. However, the qualification program for ,

the nuclear steam supply system equipmeht complies with the earlier standard; IEEE Standard 323 1971 and IEEE Standard 344-1971: clarification is needed on

_ some of the additional measures taken to upgrade the icvel of qualification because NRC requirements in the areas of aging and dynamic test methods exceed 5 those of the earlier ctandards. The staff requires that the FSAR describe further the use made of applicable references in SRP Section 3.10 and other .

[ guidelines to ensure that the equipment is qualified and will operate properly '

c under all irposed design aad service conditions, including the loadings irpo',td

[ by the safe shutdown earthquake, postulated accidents and loss of ecolant acsidents. The folicwing areas require clarification or reselution:

(I) The extent to which the complete draft standards ANSI /AS!!E QNPE 1

(N551.1), GNPE-2 (n551.2) QNPE-3 (NS51.3), QNPE-4 (N55.4) and N41.6 and issued standard ANSI /ASME B.16.41 are used needs to te clearly stated in the FSAR.

r (2) The applicant has stated cempliance with Regulatory Guide 1.148. Hewever, L the discussien of testing in the operational condition is limited with regard

_ to several cceponents. Assessment of degraded conditions and how testing was tailored to rett the requirements of SRP Sectien 3.10, paragraph !!.1.a(2) should be addressed in the FSAR. Also, a cemitment to Regulatory Guide 1.148 fer replacerent cceponents should be clearly stated in the FSAR.

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[ (3) The applicant should anond the existing tables of purps and valves in the

FSAR to include the standards used for qualification. As an alternate, a separate table may be provided which includes the above inferration correlated to tables 3.9A 1 and 3.98 4 E

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(4) In many cases the motor of an a.sembly was independen*1y qualified and the pump separately qualified for operation, using the inputs at the mounting.

Further justification is needed in the FSAR to describe how an acceptable qualification of the asseebly was arrived at, considering simultaneous dynamic interactions between the pump, motor and pedestel/nounting structure.

(5) Aging and the sequence of environmental conditions on the qualification process is only briefly addressed in the FSAR. Clarify how these findings will be reflected in the maintenance and surveillance program. The FSAR should include the criteria for the maintenance program as it relatea +6 equipment qualification test and analysis results.

(6) The criteria used to determine what auxiliary, active, safety-related er,uipment is included in the FSAR tables of active safety-related equipment should be described 'n an amendment to the FSAR. For example, while the HPCS diesel generator is described in Section 8.3.1.1.2, active valves and other ..

auxiliary safety related equipment related to the air starting system should be' included in Table 3.9A 9.

(7) The FSAR should be amended to include the ger.eric testing criteria for qualifying check valves for service conditions. The FSAR should address considerations ef load conditions (end 1 cads, vibrations, seismic and reverse flew) and environmental conditions (thermal and radiation aging of sensitive materials) and their irpact on valve function and valve leakage.

The applicant should submit FSAR Amendments to re.o., the identified FSAR deficiencies. In addition, the PYORT will follew the applicant's effort closely, and will confirm its implementation during the en-site plant audit.

During the on-site plant audit the staff will review in detail the implerentation of the applicant's program to cenfirm that all applicable leads and cortinations of leads have been defined, operability has been verified through apprcpriate tests and analyses, assemblies rather than individual

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+ components have been verified operable, and that for all safety-related equipment operability can be assured through the plant life. At least 85% of (1

the safety-related equipment must be qualified, documented in an auditable m' a nner, .and installed before an on-site planc audit by the PVORT can be performed. When the applicant indicates that his work has reached this status the PVORT will schedule an on-site audit. The staff will report the results of the audit and the followup and resolution of the concerns described above in a future supplement to the SER.

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DISTRIBUTION i DOCKET NO. 50-410 l l SDOCKET FILE l

MEMORANDUM FOR: Elinor Adensam, Director PSB R/F l BWR Project Directorate No. 3 F. Witt

! Division of BWR Licensing J. Kudrick D. Vassallo FROM: Gus C. Lainas, Assistant Director G. Lainas Division of BWR Licensing bl'T

SUBJECT:

NINE MILE POINT UNIT 2 TECHNICAL SPECIFICATION The Plant Systems Branch has reviewed the applicable sections of the applicant's submittal, dated August 6,1986, regarding additional Technical J Specification changes. We have completed our review for all the requested

. changes except one. The MSIY related change has required significant amounts -

of added information which we have recently received. This issue should be resolved prior to exceeding 5% power since the HSIV's have been leak tested '

but, by a test method which has not been approved by the staff. The fact that I tests have been performed provided reasonable assurance that some degree of leak tightness has been demonstrated, even if further testing becomes necessary.

For the remaining changes, we find them acceptable with minor modifications (the marked up pages are enclosed). Additionally, changes related to 10 CFR l 50, Appendix J, required a supporting SER which is also enclosed. Our SALP l for this effort is attached. i l

ORIGINAL SIGNED BY l D. VASSALLO FOR Gus C. Lainas, Assistant Director l Division of BWR Licensing c

Enclosures:

As stated cc w/ enclosures:

R. W. Houston T. Speis D. Crutchfield E. Rossi C. Schulten M. Haughey

Contact:

F. Witt (X29440) ho o104 e60027 Dock 0300043g CF 5520 Document Name: 9 MILE POINT TECH. SPEC. '

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DATE :8/n/86 :8AV 6 :8/s E86 :8/11/86 :8/9/86  :  : ,

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Subject:

Justification for Technica.1 Specification Change to Main Steam Isolation Valve Leak Rate The current Nine Mile Point Unit 2 Technical Specification Table 3.6.1.2-1 allows six standard cubic foot per hour (sefh) of leak rate per Main Steam Isolation Valve (MSIV). This leak rate is based on potential bypass l

analytical limit of 6 scfh of leakage through the valve under l

Loss-of-Coolant-Accident (LOCA) condition. To ensure that the HS!V leak rate l is within the Technical Specification limit, the MSIV ball valve is leak tested through a test connection such that the volume between the valve's two seats is pressurized to test condition. The flow resistance under this test cor.dition (two seats in parallel) is less than the flow resistance that would be encountered under the LOCA condition (two seats in series). Thus, the leak rate when testing the valves between the seats could exceed 6 scfh but still satisfy the LOCA potential bypass analytical limit for leakage through the valve. Calculations show that a leak rate under field test condition of 14.86 scfh (valve seats in parallel) is equivalent to the LOCA bypass analytical limit of 6 scfh with the valve seats in series. Niagara Mohawk, therefore, requests changes to the Technical Specification allowable leakage rete for the MS!Vs to reflect the actual test configuration. The requested change to the Technical Specifiction Table 3.6.1.2-1 is attached.

CHANGE REQUESTED TO SUPPORT OPERATIONAL FLEXIBILITY f

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Justification'for change to Technical Specification Bases 3/4.6.3, i "Primary Containment Isolation Valves" The requested change is enclosed. This change will clarlfy the relationship between isolation system instrumentation response time and Isolation valve closing time.

CHANGE REQUESTED FOR CLARIFICATION .

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Subject:

Justification for changes to Technical Specification Tables

3. 3. 7.10-1 and 4. 3. 7.10-1 in the area of radioactive liquid effluent monitoring instrumentation The current Technical Specification Section 3.3.7.10 requires tne Liquid Radwaste Monitor to be OPERABLE at all times, whether radwaste discharge is occurring or not. System design provides three valves to prevent inadvertent discharge. These vaivas must be specifically lined up in the course of making a discharge. Inherent in this design is the isolation of the small section of I discharge line from and to which the liquid Radwaste monitor's sample pump takes supply and return. When in continuous use, the sample pump produces more heat than can be dissipated in the small volume of water contained in this section of pipe. Therefore, it is requested to revise Technical Specification Tables 3.3. 7.10-1 and 4.3.7.10-1 to provide:
1. The Liquid Waste Monitor must be OPERABLE at all times during discharge of liquid waste.
2. The CHANNEL CHECK and SOURCE CHECK are to be perfomed P (prior to discharge).

Thei.101 renuested changen to ff changes thnical 8pecification Tables ,3 7)inal

]01 aafety and Analysis 4.L are enclosed. These also af fect the r Report and the Safety Evaluation R oort. Changes to the appropriate pages of these reports are also enclosed.

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e INSTRUMENTATION HONITORING INSTRUMENTATION RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION l l

l LlHITING CONDITIONS FOR OPERATION . l 3.3.7.10 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3.7.10 1 shall be OPERABLE with their Alare/ Trip 5etpoints set to ensure that the Ilmits of Specification 3.11.1.1 are not exceeded. The Alarm / Trip Setpoints of these channels shall be determined and adjusted in accordance with the hethodology and parameters in the OFFSITE DOSE CALCULATION ,

MANUAL (CDCM). ,

APPLICABILITY- * * ' ' ' ' - - - [ MMM 7

  1. l ACTION:
a. With a radioactive liquid effluent monitoring instrumentatio'n channel Alarm / Trip 5etpoint less conservative than required by the above specifi-cation, imediately suspend the release of radioactive liquid effluents monitored b;' the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative.
b. With the number of channels OPERABLE 1ess than the Minimum Channels OPER-ABLE requirement, take the ACTION shown in Table 3.3.7.10 1. Restore ,

the instruments to CPERABLE status within 30 days and, if unsuccessful, I explain in the next Semiannual Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner.

c. The provisions of Specifications 3.0.3 and 3.0.4 are not app'.icable.

SURVEILLANCE REQUIREMENTS 4.3.7.10 Each radioactive if cuid ef fluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOUPrE

  • CHFCK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST at the frequencies shown in 1rable 4.3.7.101.

1 NINE MILE POINT UNIT 2 3/4 3 98 JWi 151986 4

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TABLE 3.3.7.10-1 RADIOACTIVE LIQUID EFFLUENT HONITORING INSTRUMENTATION HINIMUM CHANNELS

, INSTRUMENT OPERABLE ACTION

1. Radioactivity Monitors Providing Alarm and Automatic Termination of Release .

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Liquid Radwaste Effluent Line 1 128 l

2. Radioactivity Monitors Providing Alarm But Not Providing Automatic Termination of Release
a. Service Water Effluent Line A 1 130
b. Service Water Effluent Line B 1 130
c. Cooling Tower Blowdown Line 1 130
3. Flow Rate Measurement Devices
a. Liquid Radwaste Effluent Line 1 131
b. Service Water Effluent Line A 1 131
c. Servi.e Water Effluent Line B 1 131 l
d. Cooling Tcwer Blowdown Line 1 131 4 Tank Level Indicating Devices
  • 1 132 I

a Tanks included in this specification are those outdoor tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and

  • do not have tank overflows and surrounding area drains connected to the liquid radweste treatment system, such as temporary tanks.

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1. Radioactivity Monitors Providing Alarm E and Automatic Termination of Release P OC

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Radioactivity Monitors Providing Alarm But Not 2.

Providing Automatic Termination of Release

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a. Service Water Effluent Line A R D M R(c) SA(b)

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a. Liquid Radwaste Effluent Line l HA R Q
b. Service Water Effluent Line A 0(d)

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  • Tanks included in this specification are those outdoor tanks that are not surrounded by lin l

h capable of holding the tank contents and do not have tank overflows and surrounding area drains connected w

the liquid radwaste treatment system, such as temporary tanks.

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ge to Technical Definition 1.38. "Secondary

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The requested change is enclosed. The change reflects the Nine Alle Point Unit 2 design. This change is made in order to make definition 1.38 consistent with 4.6.5.1.b.2 on page 3/4 6-37. 1 CHANGE REQUESTED FOR CERTIFICATION WC ,

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DEFINITIONS REAC10R PROTECTION SYSTEM RESPONSE TIME 1.35 (Continued) until deenergization of the scram pilot valve solenoids. The response time rnay be measured by any series of sequential, overlapping, or total steps so that the entire response time is measured.

REPORTABLE EVENT 1.36 A REPORTABLE EVENT shall be any of those conditions specified in  !

10 CFR 50.73.

ROD DENSITY 1.37 R00 DENSITY shall be the number of control rod notches inserted as a fraction of the total number of control rod notches. 'll rods fully inserted is equivalent to 10% R00 DENSITY, .

SECONDARY CONTAINMENT INTEGRITY l

1.38 SECONDARY CONTAINMENT INTEGRITY shall exist when:

a. All reactor building and auxiliary bay penetrations required to be closed during accident conditions are either:
1. Capable of being closed by an OPERABLE reactor building automatic isolation system, or
2. . Closed by at least one manual valve, blind flange, or deactivated auto-matic damper secured in its closed position, except as provided in l Table 3.6.5.2-1 of Specification 3.6.5.2.
b. All auxiliary bay hatches are closed and sealed,
c. The standby gas treatment system is in compliance with the requirements of l Specification 3.6.5.3. .
d. At least one doo in tach access to the reactor building and auxiliary bays is closeds-  % M g d p. 1
e. The sealing mechanism associated with each reactor building and auxiliary l bay penetration (e.g. , welds, bellows, or 0-rings) is OPERABLE. l
f. The pressure within the reactor building and auxiliary bays is less than or equal to the value required by Specification 4.6.5.1.a.

, SHUT 00W KARGIN 1.39 SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor is suberitical or would be suberitical assuming all control rods are fully in-serted except for the single control red of highest reactivity worth which is HINE HILE PolNT - UNIT 2 1-7 JUN 2 51986

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NIAQ ARA WoH AWK POWER CoAPoRAftoMI30C [R-( EOg,[vA AO W($t $YRACy${ N 4 13IO2 *I.{ I*C',( 7'! .

  • 4 l July 3. 1936 (fiMPEL 0770) l Hs. Elinor G. Adeasam Otr?: tor BWR Project Olrertorate No 3 i 1

U.S. N- ear Regt latory Comn t ssion 7920 Nv folk Aver.je Washington, DC 20'55 4 I

Dear Ms. Acensam-Re: Nine Mlle Point Unit 2 Docket No. 50 a10 Niagara Mohawk requests changes to Nine Mlle Point Unit 2 Technical Specifications in the area of Fire Protection Program.

In response to our discussion with Mr. C. Shulten and Mr. D. Kubickt on June 3,1986, the proposed changes to the Technical Specifications, as well as j justification for these changes are attached.

Very truly yours. .

C. V. Mangan Senior Vice President LL:ja Od 1778G ,- ' /.t < 4 'J,3

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Attachments s '.' 5. . ,',7 ./ ,

f ic: R. A. Gra53, NRC Resident inspector ~, n.

Project File (2) ". t ' ' ' ^~ "7 r

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UNITED STATES OF AMERICA NUCLEAR REGULAt0RY COHN!SS!ON l

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Niagara Mohawk Power Corperation ) Docket No. 50 41^ l l

(11108 H)lt P4101 4n11 il 1 t v Aff4@(j_{

_ C. V. Mancan , being duly tworn. Statet that he 11 Senior Vico President of Niagara Mohawk Power Corporation; that he 11 authortred on the {

j part of said Corporation to sign and file with the Nuclear Regulatory Comission the documents attached hereto; and that all such documents are true i and correct to the best of his knowledge, information and belief. l l

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O I M A__M> h Subscribedandsworntobeforeme,aNotary(PublicinandfortheStateofNew York and County of OAMddoa , t h i s d' _, day of Atlu , 1986.

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PLANT SYSTEMS

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' IRE SUPPRESSION SYSTDis HALON SYST(MS V

4 LIMITING CONDITIONS FOR OPERATION ,

I 3.7.7.4 The following Halon systems shall be CPERA8tE with the storage tanks I having at least 95% of full charge weight or levet'and 9(M of full enarge pre s sure.

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ZONE Ho. BUILDING / ELEVATION .

353 SG Control /284' 6' I 354 SG Control /288' 6' 362 SG Control /288' 6' 357 XG Control /288' 6' 358 XG Control /288' l' 374 SG Control /3OJ' 0' 375 SG Control /306' 0*

361 SG Control /306' 0' '

376 XG Control /306' 0' I

APPLICA81LITY: Vbenever equipment protected by the Kalon systess is required to be OPERABLC.

ACTION:

a. With one or nors of the above required Malon systees inoperable, within ,

I hour estabitsh a continuous fire watch with baciup fire suppression equipeent for those areas in which redund!nt systems or components could be damaged; for c?her areas, establish an hourly fire watch patrol.

b. The provisions of Specifications 3.0.3 and 3.0.4 are 'not applicable. ,

SURVEILJMcE REQUIRENENTS 4.7.7.4 Each of the above required Halon systans shall be demonstrated OPERMLE:

a. At least once per 31 days by verifying that each valve-sanual, power-operated, or autcoatic in the flow path is in its correct position.
b. At least once per 6 months by verifying Kalon storage tank weighte

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MIME MILE Po!NT - UNIT 2 3/4 7 30 .

APR 0 41986

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7**,PLANTSYSTEMS

" FIRE SUPPRESSION SYSTEX5 FIRE SUPP9*SSION WATER SYSTEM 4

SURVEILLANCE REQUIREMENTS 4.7.7.1.1 (Continued)

c. At least once per 6 months by performarv.e of a systes flush,

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d. At least once per 12 months by cycling each testable valve in the flow path through at least one complete cycle of full travel.

- e. At least once per 18 months by performing a systas functional test I which includes simulated autoestic actuation of the systes throughout its opereting seovence, and:

1. Verifying that each automatic valve in the flow path attuates to its correct position,
2. Verifying that each fire suppression pump develops at least 2500 gpa at a not discharge head of 113 psig, .
3. Cycling each valve in the flow path that is not testable during plan

.g.t operation through at least one complete cycle of full travel,

4. Verifying that each fire suppresskn pump starts and maintains the fire s@pression water syntes pressure of 125 psig or more, At least once per 3 years by oerforming a flow tant of the systas in

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accordance with Chapter 6, Section la, of the Fire Protection Handbook.

15th Edition, p211shed by the National Fire Protection Association.

4.7.7.1.2 The diesel driven fire suppression pump shall be demonstrated OPERA 8LE:

a. At least once per 31 days by:
1. Verify 1'ng the fuel day tank contains at least 350 gallons of fuel.
2. Starting the diesel driven pump free ambient conditions and operating for greater than or equal to 30 minutes on recirculation flow,
b. At least on a per 92 days by verifying that a sample of diesel fuel from the fuel storage tank, obtained in accordance with ASTWD4057-81, is within the acceptable lietts sp+cified in Table 1 of ASTM 0975-81 when checkad for viscosity, water, and sediment.
c. At least once per la month:! d:M CO Qby subjecting the diesel to an inspection in accordance w th proceduMs prepared in conjunction with its manufacturer's recommendations for the class of service.

NINE MILE PolKT - UNIT 2 3/4 7-25 APR 0 41

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io SINE MILE POINT. UNIT SECTION 4.7.7.4(b) RALOS SYSTEM 3 EXISTING At least once per 6 months be verifying Halen storage t a r. .. 'e i ? h t u.

pressure.

PROPOSED ,

At least once per 6 monthseby verifying Halon storage tank weigh er level and pressure.

DISCUSSION Section 3.7.7.4(LCO) provides the option of using level to determine the current capacity of the storage tanks. In addition, Bases Section 3/4.7.7 identifies that level measurements are made by either a 'JL listee or FM approved method. The change identified above would be consistent with these ref erences and enable the surveillance to be perfor=ed without physically disconnecting the storage tank f rom the discharge manifold. !c satisfy an NRC concern, a footnote will be added where ever reference is made to level measurement for this system which will read as follows:

"Level determination for the purpose of verif ying Halon System operability shall conform to NRC accepted CL or FM test procedures and/or equipment." "-

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l SISE P.It,E POIST. USIT 2 ?i h!

SECTION t. 7. 7.1.2 (c) FIRE SUPPRESSIOS ETER SYSTI.M I

EXISTING h1!

At least once per 16 months, during shutdown, by subjecting the .

dievi to an inspection in accordance with procedures prepared in )

tonjunction with its .anuf acturer's recomendations for the class of service.* .

. PROPOSED At least once per 18 months by ' subjecting the diesel ........

DISCUSS 105 Based on industry operating experience, fires are more likely to '

l occur during an outage when construction activity at the site is elevated.

Therefore, the basis for limiting this activity during shutdown is undesireabla. It is our intention to have available for service the diesel engine driven fire pu=p during the outage. if possible. Any .

necessary scheduled maintenance work would be performed either prior to.

during, specified. or af ter the outage in accordance with the surveillance interval I vill beDuring this period of maintenance, the provisions of Section 3.7.7.1 maintained.

for service to supply UnitNo 100% back-up fire punps could be readily available l

underground distribution system between the two units.2. if required. by cross connecti * '

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PtW(T SYSTEMS

  • ' M'.. l FIRE SUPPREs5 ION SYSTENs I'

HALOM SYS70tS LIMITING CON 0!TIONS FOR OPERATION l

3.7.7.4 The following Halon systems shall be OPERA 8LE with the storage tanks I I

having at least 95% of full charge weight or leveY'and 9C% of full charge pre s sure:

ZONE NO. BUILDING / ELEVATION t

353 M Control /288' 6' i 3 54 % Control /288' 6' 362 M Conter1 '288' 6' 357 XG Control /288' 6' 358 XG Control /288' 6' 374 % Control /306' 0*

375 % Control /306' 0' 381 M Control /306' 0' .

376 XG Control /304' 0* .

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APPLIC.A8ILITY: Whenever equipment protected by the Halon systees is required to t>e OPEiL6[E.

ACTION:

a. With one or more of the above required Kalon systems inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> establish a continuous fire watch with backte fire suppression equipeent for those areas in which redundant systems or components could be damaged; for other areas, establish an hourly fire watch patrol,
b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. ,,

SURVEILLAMcE REQUIRD4 EXT 5 I

4.7.7.4 Each of the above required Halon systans shall be demonstrated CPERABLE:

e. At least once per 31 days by verifying that each valve-sanual, power-operated, or auteettic in the flow path is in its correct position.
b. At least once per 6 months by verifying Kalon storage tank weight.ead==

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NINE MILE Po!KT = UNIT 2 ,

3/4 7 30 -

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Subject:

Change to Technical Specification Definition 1.42, "Source Check" The requested change and justification for the change were submitted to you in a letter dated July 24,1986. That letter is enclosed for your infomation.

O CHANGE REQUESTED FOR CERTIFICATION

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OEFINITIONS RW DWT SHUT 00VN MARGIN 1.39 (Continued)

I assumed to be fully withdrawn and the reactor is in the shutdown condition, cold (i.e., 68'F), and xenon free, t

SITE BOUNDARY 1.40 The SITE b0UNDARY shall be that line around the Nine Mlle Point Nuclear Station beyond which the land is not owned, leased, or otherwise controlled by -

the Niagara Mohawk Fower Corporation or the New York State Power Authority.

SOLIDIFICATION 1.41 SOLIDIFICATION shall be the conversion of wet wastes into a form that meets shipping and burial ground requirerents.

SOURCE CHECK 1.42 A SOURCE CHECK shall be the qualitative assessment of channel response ttr-

.irifj eier end/er trip f.n;thn; :nd ;Mard 1;ihre tript when tt.e channel Af rNr sensor is exposed to a source of increased activity. 'itt.

1. 6 U/C STAGGERED TEST BASIS # mW.

1.43 A STACGERED TEST BA5!$ shall consist of: *

a. A test schedule for n systems, subsystems, trains, or other designated Whg components obtained by dividing the specified test interval into n equal subintervsts, ,
b. The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval.

THERNAL POWER 1.44 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant. .

TURB!NE BYPA55 SYSTEM RESPONSE TIME 1.45 The TUR8INE BYPASS SYSTEM RESPONSE TIME consists of two tie < intervals:

a. Time from initial movement of the sain turbine stop valve or control valve untti 80% of turbine bypass capacity is estaclished, and

. b. the time froa initial movement of the sain turbine stop valve or control valve until initial movement of the turbine bypass valve.

Either response time may be measured by any series of sequential, overlap-ping, or total steps, so that both entire response time components are seasured. -

NINE MILE POINT UNIT 2 18 1

JUN 15 B06

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Subject:

Justification for changes to Technical Specification Table 3.6.3-1, "Primary Containment Isolation Yalves" The requested changes are enclosed. The changes are consistent to our letter dated July 3,1986 which requested thres re'.kf valves to be tested under reverse flow condition and 13 relief v0ves to be exem;:t f-om Type C testing.

Our letter dated July 3,1986 is als2 inclosed for you- infomation.

Subsequent discussic' with Mr. J. Aveick and MS, M. F i ughty, of your staff, resolved their revinc ~ rn%

  • E RT I F I C,.. t:1 CHANGE REQUIRED 3[M #

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