ML20095L197

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Proposed Tech Specs for Code Safety Valve Setpoint Tolerance Relaxation
ML20095L197
Person / Time
Site: Seabrook NextEra Energy icon.png
Issue date: 05/05/1992
From:
PUBLIC SERVICE CO. OF NEW HAMPSHIRE
To:
Shared Package
ML20095L191 List:
References
NUDOCS 9205060273
Download: ML20095L197 (20)
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Text

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11. Ntarkup of Proposed Chances See attached markup -of proposed changes to Technical Specifications; s

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REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY VALVES SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.2.1 A minimum of one pressurizer Code safety valve shall be OPERABLE with a li f t setting"of 2485 psig i )%,. *

  • 3 APPLICABILITY: MODES 4 and 5.

ACTION:

With no pressurizer Code safety valve OPERABLE, immediately suspend all operations involving positive reactivity changes and place an OPERABLE RHR loop into operation in the shutdown cooling mode.

SURVEILLANCE REQUIREMENTS 4.4.2.1 No additional requirements other than those required by Specifica-tion 4.0.5.

  • The lif t setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure. '

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REACTOR COOLANT SYSTEM SAFETY VALVES OPERATING LIMITING CONDITION FOR OPERATION 3.4.2.2, All pressurizer Code safety valves shall be OPERABLE with a lift settingFof 2485 psig i At.*

  • 3 APPLICABILITY: MODES 1, 2, and 3# .

ACTION:

With one pressurizer Code safety valve inoperable, either restore the inoper-able valve to OPERABLE status within 15 minutes or be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.2.2 No additional requirements other than those required by Specifica-tion 4.0.5.

gt- (dit% b / $ fe Ilosdng pr e. ssaeher Cecle S a$efy yal+< hsfi .

  • The lift setting pressure shall correspond to ambient conditions of the valve

- at nominal operating temperature and pressure.

  1. E ntry into this MODE is permitted for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to perform post-modification or post-maintenance testing to verify OPERABILITY of components.

ACTION requirements shall not apply until OPERABILITY has been verified.

SEABROOK - UNIT 1 3/4 4-9 ,

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TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING FOUR-LOOP OPERATION MAXIMUM NUMBER OF IN0PERABLE MAXIMUM ALLOWABLE POWER RANGE SAFETY VALVES ON ANY NEUTRON FLUX HIGH SETPOINT OPERATING STEAM GENERATOR (PERCENTOFRATEDTHERMALPOWERJ 1 87 2 65 3 43 TABLE 3.7-2 STEAM LINE SAFETY VALVES PER LOOP VALVE NUMBER 3

Loop 1 Loop 2 Loop 3 Loop 4 k.

LIFT SETTING (t M)*M ORIFICE SIZE A

V6 V22 V36 V50 1185 psig 16.0 sq. in.

V7 V23 V37 V51 1203 psig 16.0 sq. in.

V8 V24 V38 VS2 1220 psig 16.0 sq. in.

V9 V25 V39 V53 1238 psig 16.0 sq. in.

V10 V26 V40 VS4 1255 psig 16.0 sq. in.

  • The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

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3/4.7 PLANT SYSTEMS lk BASES

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l t 3/4.7.1 TURBINE CYCLE

-k 3/4.7.1.1 SAFETY VALVES

  • The OPERABILITY of the main steam line Code safety valves ensures that-the Secondary System pressure will be limited to within -110% (1320 psia) of-iY its design pressure of 1200 psia during the most severe anticipated system operational transient. The maximum relieving capacity-is associated with a D

}jTurbinetripfrom100%

condenser heat sink (i.e., no RATED. THERMAL steam bypass =POWERcoincidentwitha to the condenser).

.$l The specified valve lift settings.and. relieving capacities.are-in a.cor-d _ dance with the requirements of Section III of the ASME Boiler and Pressure'

{ " Code,'1071 Editia. The-total relieving-capacity for all valves'~on all of the ,

,'g steam lines is 1.839 x 107 lbs/hr which is-121% of the total secondary steam flow of 1.514 x 107 Ibs/hr at 100% RATED THERMAL POWER. A minimum of two-OPERABLE safety valves per steam generator. ensures that-sufficient relieving k

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capacity

3. is 7-available

/. for the allowable THERMAL POWER restriction in Table o

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/ STARTUP and/or POWER OPERATION is allowable with safety valves inoperable j

/ within the limitations of the ACTION requirements on the basis of the reduction  ;

in Secondary Coolant System steam f'ow and THERMAL POWER required by the- i reduced Reactor trip settings of the Power Range Neutron Flux channels. The Reactor Trip Setpoint-reductions are derived on the following bases:

For four loop operation:

Sp = (X) - (Y)(V) x 109

Where

SP = Reduced Reactor Trip Setpoint in percent'of RATED THERMAL.

POWER,

(

V = Maximum number of inoperable safety valves per steam line,

. 109 = Power Range Neutron Flux-High~ Trip Setpoint for_four loop l operation, X = Total relieving capacity of all safety valves per steam line in lbs/hr.-and~

Y = Maximum relieving capacity of any one safety valve in i lbs/hr I

l-SEABROOK UNIT 1 8 3/4 7-1 y

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III. Hetyne of Pronosed Channes See attached retype of proposed changes to Technical Specifications. The attached retype reflects the currently issued version of Technical Specifications. Pending Technical Specification Changes or Technical Specification changes issued subsequent -

to this submittal are not reflected in the enclosed retype. The enclosed retype should be checked for continuity with Technical Specifications prior to issuance.

Revision bars are provided in the right hand margin to indicate a revision to the. text, No revision bars are utilized when the page is changed solely to" accommodate the shifting of text due to additions or deletions.

5

REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY VALVES SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.2.1 A minimum of one pressurizer Code safety valve shall be OPERABLE with a lif t setting

  • of 2485 psig i 3%."

APPLICABillTY: MODES 4 and S.

ACTION:

With no pressurizer Code safety valve OPERABLE, immediately suspend all operations involving positive reactivity changes and place an GPERABLE RHR loop into operation in the shutdown cooling mode.

SURVEILLANCE REQUIREMENTS 4.4.2.1 No additional requirements other than those required by Specification t 4.0.5.

  • The lift setting pressure shall correspond to ambient _ conditions of the valve >

l at. nominal operating temperature and pressure.

    • Within 11% following pressurizer Code safety valve testing.

SEABROOK - UNIT 1 3/4 4-8 Amendment No.

_.__ _ _ _ _ _ _ _ .. _._. _ -- _ _ . . _ . _ _ _ _ . _ _ _ _ _ _ . _ ~ . _

1 RfACTOR COOLANT SYSTEM l SAFE 1Y VALVEE OPERATING LIMITING CONDITION FOR OPERATION ._ _

3.4.7.2 All pressurizer Code safety valves shall be OPERABLE with a lif t setting' of 2485 psig i 3L" APPLICABillly: MODES 1, 2, and 3#.

ALT 10N:

With one pressurizer Code safety valve inoperable, either restore the inoper.

able valve to OPERABLE status within 15 minutes or be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in at least HOT SHUT 00WN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. l I

SURVEllLAfJCE RE0VIREMENTS ,

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4.4.2.2 No additional requirements other than those required by Specification l 4.0.5.

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  • The liTt setting presture-shall correspond to ambient conditions of the valve-l at nominal operating temperature and pressure.-
    • Within 11% following pressurizer Code safety' valve testing.

-# Entry into this MODE is permitted for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to perform post-modification or post-maintenance testing to verify OPERABILITY of components.

ACTION requirements shall not' apply until OPERABILITY has been verified.

1 SEABROOK - UNIT 1 3/4 4-9 Amendment No.-

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TABLE L71 JiAUMUM AL LOWABLE POWER RANGE NEUTRON ELUX HIOi; SETPOINT WIT 11 i INOPERABLE STEAM LINE, SAFETY VALVE 9 DURING FOUR-LOOP OPERATION l l

MAXIMUM NUMBER OF IN0PERABLE HAXIMUM ALLOWABLE POWER RANGE SAFETY VALVES ON ANY NEUTRON FLUX HIGH SETPOINT OPERATING S1EAM GENERATOR (PERCENT OF RATED THERMAL POWE81 1 87 2 65 3 43 TABLE 3.7-2 STEAM LINE SAFETY VALVES PER LOOP VALVE NUMBER toon 1 Loon 2 Loon 3 Loop 1 OET SETTING * (1 3%1** ORIFICE SIZE V6 V22 V36 V50 1185 psig 16.0 sq. in. '

V7 V23 V37 V51 1203 psig 16.0 sq. in.

. V8 V24 V38 V52 1220 psig 16.0 sq. in.

V9 V25 V39 V53 1238 psig 16.0 sq. in.

V10 V26 V40 V54 1255 psig 16.0 sq. in.

  • The lift setting pressure shall correspond to ambient conditions of the valve

(

at nominal cperating temperature and pressure.

a SEABROOK - UNIT 1 3/4 7-2 Amendment No.

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3/4.7 Pl ANT SYS.1Dili BASES 3/4.7.1 TVRDINE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line Code safety valves ensures that the Secondary System pressure will be limited to within '110% (1320 psia) of its design pressure of 1200 psia during the most severe anticipated system ooerational transient. The maximum relieving capacity is associated with a Turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser).

The specified valve lift settings and relieving capacities are in accordance with the requirements of Section III of the ASME Boiler and Pressure Code, 1974 Edition, including the Summer 1985 Addenda. The total rolleving capacity for all valves on all of the steam lines is:1.839 x 10' lbs/hr which is 121% of the total secondary steam flow of 1.514 x 10' lbs/hr at 100% RATED THERMAL POWER. A minimum of two OPERABLE safety-valves per steam generator ensures that sufficient relieving capacity is available for the allowable ,

THERMAL POWER restriction in Table 3.7 1, l

STARTUP and/or POWER OPERATION is allowable with- safety valves I

inoperable within the limitations of the ACTION requirements on the basis of-the reduction in Secondary Coolant System steam flow and THERMAL POWER required by the reduced Reactor trip settings of the Power Range Neutron Flux channels.

The Reactor Trip Setpoint reductions aie derived on the following bases:

For four loop operation:

SP - (X) - (Y)(U X 109 X

Where:

SP = Reduced Reactor Trip Setpoint in percent of RATED 'iHERMAL POWER, V = Maximum number of inoperable safety valves per steam line, 109 - Power Range Neutron Flux-High Trip Setpoint for four loop operation, X - Total relieving capacity of all safety valves per steam line in 1bs/hr, and Y - Maximum relieving capac'ty of any one safety valve in Ibs/hr i

SEABROOK - UNIT 1 B 3/4 7-1 Amendment No.

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IV. Snfety Auenment of Proposed Chunnen New llampshire Yankee is proposing to revise the Seabrook Station Technical Specifications to allow a relaxation in the Pressurirer Sably Valve (PSV) and Main Steam Safety Valve (MSSV) setpoint tolerances to 2 3% for AShlE Section XI testing i acceptance criteria. The proposed Techcical Specification changes also require that '

the PSV and MSSV setpoints be restored to within 1 1% of their nominal setpoints following testing. New llampshire Yankee is also proposing to revise the llASES for  !'

Technical Specification 3/4.7.1.1 to specify the correct Edition of the AShtE lloiter and Pressure Vessel Code Section ill applicable to the MSSVs. Additionelly, NilY .

is proposing to correct a typographical error in the !!ASUS for Technical Specification  !

3/4.7.1.1. The llASES currently contain an incorrect reference to Technical Specification Tab!c 3.7 2, whereas the correct reference is to Table 3.71.

The Seabrook Station overpressure protection design incorporates three Code safety vaiva on the primary system pressurizer and a total of twenty Code safety valves on the four main steam lines (five per line)-in the secondary system. The pressurizer safety valves (PSVs) were designed and manufactured to meet the 1971 Edition-including the Winter 1972 Addenda of the ASME Code, Section Ill. The main steam -

safety valves (MSSVs) were designed and manufactured to meet. the 1974 Edition including the Summer 1975- Addenda of the- ASME Code, Section 111. An . ASME Code. Section 111 requirement for both the PSVs and the MSSVs is that they be  !

designed to open within 2- 1% of their set pressure. The current Technical Specification Limiting Conditions for Operation (LCO) for the PSVs and MSSVs also impose the tolerince of 1% on their set pressure.

The Technical Specillcation Surveillance Requirements for the PSVs and the MSSVs require that testing be performed in accordance with Section XI of the ASME lloller  ;

and Pressure Vessel Code.and applicable Addenda as required by 10CFR50.55a(g), =

excert where specific written relief has been granted by the Commission. The PSVs and the MSSVs are tested to ' verify that their lift pressures and-seat leakages are acceptable pursuant to the New Hampshire Yankee Inservice Test (IST) Program which complies with the ASME lloller t.nd Pressure Vessel Code,.Section XI,-1983 Edition through the Summer 1983 Addenda. The NRC evaluation of the NilY IST Program is documented in NUREG 0896, Supplement' No. 6, " Safety Evaluation Report Related to the Operation of Seabrook Station, Units 1 and 2", dated October 1986.

The 1983 Edition of ASME Section XI does not specify a tolerance to be applied to safety valve lift pressure verification; therefore the tolerance (2 1%) prescribed-in _

the LCO for the PSVs and MSSVs is utilized as the acceptance criteria for ASME ,

Section XI testing. ASME Section XI Article IWV 3513 requires additional ufety. -

valve testing when a safety- valve " falls to function properly". Currently, a PSV or-MSSV which has a tested lift pressure outside the 2 l ' 70 tolerance specified in the LCO is determined to have failed to function properly, thereby requiring repair or replacement per IWV 3514 and. testing of additional valves in the - system - per IWV 3513. ,

The 1989 Edition of.the ASME Code,Section XI, requires that the PSVs andlMSSVs-be tested pursuant to the ASME/ ANSI OM 1987, Part 1,

  • Requirements for Inservice Performance Testing of Nuclear Power Plant Pressure Relief Devices." This standard allows the tested lift pressure to exceed the stamped set pressure by up to 3% before declaring a test failure, it also provides a guideline for testing additional valves when 6-yMy "4 e?'r-w -

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a valve exceeds the 1 3% tolerance. Therefore, increasing the PSV and htSSV ,

setpoint tolerance to 1 3% for testing acceptance criteria is in compliance with the I later AShtH Code,Section XI requirements. ,,

I The proposed relaxation of the setpoint tolerances for the PSVs and the htSSVs has been det:rmined to be in compliance with the 1989 Edition of the AShth Code, Section lit, Subasticle Nil 7410/NC 7410, which states that *The set pressure of at least one of the pressure relief desices connected to the system not be greater than the Design Press ste of any component within the pressure retaining boundary of the protected sy s t e m, The Reactor Coolant System design pressure is 2485 psig (reference UFSAR Table 5.31) which corresponds to the setpoint of the PSVs. The  :

hiain Steam Supply System design pressure is- 1185 psig (reference UFSAR Section 10.3.2.1) which corresponds to the Group 1 htSSVs which have the lowest opening setpoint.

It is unp 3rtant to note that NilY proposes to utillre the 2 3'?u tolerance for the "as found" acceptance criteria for additional valve testing required by AShtE Section >

XI, Subsection IWV 3513. The proposed Technical Specification revisions require that the PSV and htSSV setpoints be restored to within i 1% of their nominal setpoints following testing.

The impact of the relaxed PSV and htShV setpoint tolerance on the licensing' basis analysis documented in the Scabrook Station Updated Final Safety Analysis Report -

(UFSAR), Chapter 15, has been reviewed by Yankee Atomic Electric Company (YAEC) and documented in topical report YAEC 1847 "Seabrook Station Code Safety Valve Setpoint Tolerance Relaxation". A copy of YAEC 1847 is enclosed in Section Vill herein. YAEC 1847 demonstrates that the licensing basis criteria are still met when the relaxed Code salety valve tolerance of 2 3% is assumed, YAEC 1847 demonstrates the following: ,

1. For events where Departure From Nucleate Boiling Ratio (DNUR) is a concern, there will be no reduction in the calculated minimum DNBR,
2. For events where overpressurization is a concern, the safety limits are not exceeded, and 3, For events where offsite doses are a concern, the safety limits are not exceeded, and 4 For LOCA cvents, the acceptance criteria for Emergency Core Cooling Sysicm (ECCS) performance are not exceeded. -

The proposed Code safety valve setpoint tolerance relaxation does not affect ITSAR Departure From Nucleate Boiling (DNB) evaluations. The UFSAR DNU. evaluations take credit for operation of the pressuriter Power Operated Relief Valves (PORVs) which have a setpeint of 2400 psia. This setpoint is lower than the proposed lower limit of 2425 psia for the PSVs. The UFSAR conservatively assumes PORY operatiou because lower Reactor Coolant System pressures yield more limiting values of DNBR, The YAEC 1847 evaluation of th'e lower limit of the relaxed PSV and htSSV setpoint ,

, tolerance, 3% concludes that there will be no increase- in the frequency : of 7

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. l challenges to either the PSVs or the htSSVs due to the relaxed setpoint tolerance. i A margin of 25 psi remains between the PORY setpoints and the relaxed PSV l setpoints. The proposed lower limit of the PSV setpoint ( 3% tolerance) is 2425 1 psia. This PSV setpoint is evfficiently above the PORY setpoint of 2400 psia to prevent unnecessary challenges to the ISVs. A margin of 24 psi remains between the l Atmospheric Steam Dump Valve- (ASDV) selpoints and the relaxed Group 1 htSSV i setpoints. The Group 1 MSSVs, which have the lowest opening setpoint, have a i proposed lower limit setpoint (- 3% tolerance) of 1164 pela. This is sufficiently above the ASDV opening setpoint of 1140 psia to prevent unnecessary challenges to j the htSSVs. YAEC 1847 also concluded that the proposed lower limit of the PSV  ;

setpoint, 2425 psia, will not affect the automatic reactor trip on high pressu I pressure which occurs at 2400 pala, j i

The YAEC 1847 evaluation of the relaxed Code t.afety <alve setpoint tolerance change

, considered each of the non LOCA transient events documented:in Chapter 15 of the  !

! UFSAR. For most of the UFSAR Chapter 15 non LOCA transient events it is apparent that a relaxation in the IlSV and MSSV setpoint tolerance cannot have an effect on the existing UFSAR analysis as for example when the opening setpoint is ,

not challenged. The-following UFSAR Chapter 15 non LOCA transient = cvents are -

3 evaluated in YAEC 1847:

Reduction in Fcedwater Temperature increase in Fcedwater Flow Excessive increase in Steam Flow -

Inudvertent Opening of a Sicam Generator Relief or Safety Valve Steam Line lireak

, Inadvertent Closure of htSIVs Loss of Condenser Vacuum .

Loss of Off Site Power Loss of Feedwater Flow Feedwater Line lireak Partial and Complete Loss of Reactor Coolant Flow React.:r Coolant Pump Shaft Seizure '

Reuctor Coolant Pump Shaft liteak -

Uncontrolled RCCA Withdrawal From Suberitical Condition Uncontrolled RCCA Withdrawal at Power RCCA hilsoperation Startup of an Inactive Reactor Coolant Pump 11oron Dilution Fuel Loading Error l RCCA Ejection

  • t inadvertent Operation of ECCS During Power Operation  !

CCS htalfunction That increases Reactor Coolant Inventory Decrease in Reactor Coolant Inventory (non LOCA)

The most detailed YAEC 1847 evaluation of the non LOCA transient events was performed for the limiting pressurization transient. the Turbine Trip. This cvent was simulated using the RETRAN02 htODS computer code. _ This computer code has been approved by the NRC for this type of system transient as documented in' NRC Safety i

8

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Evaluation Report dated November 1,1991. Comparisons of the UFSAR Turbine Trip analysis with the RETRAN analysis of the Turbine Trip event were performed by YAEC to demonstrate that the RE1RAN model provides comparable resu'.ts.

Agreement with the UFSAR results was very close, with the RETRAN analysis results slightly overpredicting the peak pressure as compared to the UFSAR. Evaluations with the relaxed PSV and MSSV setpoint tolerances were then performed with the REIRAN model demonstrating that the peak pressure remains well below the Condition 11 limit (Events .J Moderate Frequency) of 110% of design pressure, or 2750 pain for the primary system and 1320 psia for the secondary system.

The impact of the proposed MSSV setpoint tolerance relaxation on the UFSAR design basis LOCA events were evaluated (the PSVs are not challenged in a LOCA transient therefore the proposed PSV setpoint tolerance relaxation doer not affect LOCA analyses). The limiting LOCA analysis for Seabrook Station is a large break LOCA event yiciding a Peak Clad Temperature (PCT) of 2041.2 'F. The limiting large btcak LOCA event is not affected by the proposed MSSV setpoint tolerance relaxation.

because the MSSVs are not challenged in this event as a result of the drawdown of secondary system pressure by the primary system pressure. The proposed relaxation of the MSSV setpoint tolerance does however affect the results of the limiting small break LOCA analysis which does predict a challenge to the MSSVs. __ The UFSAR evaluation of the limiting small break LOCA event -specifies a PCT of 1790.0 'F (reference UFSAR Sec.15.6). The current small break LOCA PCT including margin allocations reported to the NRC by NilY on July 26,1991 (Reference NYN 91120) is 1973.2 'F. The YAEC 1847 evaluation of the Code safety valve setpoint tolerance relaxation specifies an increan in the limiting small break LOCA PCT of about 2.5 "F. YAEC 1847 recomroends that a conservative PCT penalty of 5 'F be applied to the small break LOCA PCT result and be tracked in accordance with 10CFR$0.46 reporting requirements. This increase in PCT is less than 50 *F and therefore is not a "significant" change as defined by 10CFR50.46. The revised small break LOCA PCT value of 1978.2 'F remains below the 2200 *F limit as well as bclow the large break LOCA PCT value of 2041,2 'F. NilY will report the small break LOCA PCT increase resulting from the proposed Code safety valve setpoint tolerance relaxation in. its 10CFR50.46 annual report subsequent to NRC approval of this license amendment request.

The impact of the proposed Code safety valve setpoint tolerance relaxation on the design basis Steam Generator Tubc Rup are- (SGTR) radiological consequences is evaluated in YAEC 1847. An oserall increase in Sted Generator steam mass release is predicted due to the increased MSSY "open" time from 20 seconds to 31.5_ seconds.

The MSSV 'open" time occurs immediately a.'ter the reactor / turbine trip which is early on in the SGTR event and lasting only approximately 30 seconds; YAEC 1847 predicts that un incremental dose inernase of approximately 21 will occur due to the slightly longer MSSV "open* time. The' incremental dose increase is small, to the extent that it is within the round off error of the predicted Exclusion Area Boundary two hour dose or the Low Population Zone eight hour dose. The design basis SGTR radiological consequences are specified in NiiY's letter dated April 16,1991, " A nalysis 1 of a postulated Design Basis Steam ~ Generator Tube Rupture for Seabrook Station" (Reference NYN 91061) 9

_ _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ = _ _ _ _ _ - . . - _ . . _ ._. . . _ . _ _

. l V. J)ktermination of Sluni!Res.nt flasards for License Amendment Request 9111 (1) The proposed changes do not involve a significant increase in the probability or ansequences of an accident previously evaluated.

i New llampshire Yankee is proposing to revise the Seabrook Station Technical Specifications to allow a relaxation in the Pressuriter Safety Valve (PSV) and Main Steam Safety Valve (MSSV) setpoint tolerances to 2 3% for ASME Section XI testing acceptance criteria. The proposed Technical Specification changes also require that the PSV and MSSV setpoints be restored to within

  • 1% of their nominal setpoints '

following testing. New llampshire Yankee is also proposing to revise the BASES for Technical Specification 3/4.7,1,1 to specify se correct Ed. tion of the ASME Boller and Pressure Vessel Code. Section lit applicable to the MSSVs. Additionally, NilY is proposing to correct a typographical error in the BASES for Technical Specification 3/4.7.1.1. The BASES currently contain an incorrect reference to Technical Specification Table 3.7-2. whereas the correct reference is to Table 3.71.

The impact of the relaxed PSV and MSSV setpoint tolerance on the licensing basis analysis documented in the Seabrook Station Updated Final Safety Analysis lleport (UFSAR), Chapter 15, has _ becn reviewed by Yankee Atomic Electric Company (YAEC) and docun.cnted in topical report YAEC 1847 *Scabrook Station Code Safety i

Valve Setpoint Tolerance Relaxation, YAEC 1847 demonstrates that the licensing basis criteria are still met when the relaxed Code safety valve tolerance of 3% is assumed. YAEC 1847 demonstrates the following:

1, For events where Departure From Nucleate Holling Ratio (DNBR)is a concern, there will be no reduction in the calculated minimum DNBR,

2. For events where overpressurization is a concern, the safety limits are not exceeded, and
3. For events where offsitn doses are a concern, the safety limits are not execeded, and
4. For LOCA events, the acceptance criteria for Emergency Core Cooling System performence are not exceeded.

(2) The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

1 The overpressure protection function of the PSVs and MSSVs is not affected by the proposed relaxation of their setpoint tolerance. The nominal setpoint on the PSVs and the MSSVs will not be changed. New llampshire Yani;ee proposes to utilire the 1 3% tolerance for the "as found' acceptance criteria for additional valve testing required by ASME Section XI, Article IWV 3513. The proposed _ Technical Specification revisions require that the PSV and MSSV setpoints be restored to within z 1% of their nominal setpoints following testing. . The evaluation of the proposed relaxation of the PSV and MSSV setpoint tolerances documented in YAEC 1847 demonstrates that new or different kinds of accidents are not created by the proposed changes.

10

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(3) The proposed changes do not result in a significant reduction in the margin of safety.  !

The Seabrook Station overpressure protection design incorporates three Code safety salves on the primary system pressuriter and a total of twenty Code safety valves on  ;

the four main steam lines (five per line) in the secondary system. The proposed j Technical Specification changes involve a relanatior. of the setpoint tolerance for the Code safety valves in the primary and secondary systems. The YAEC 1847 evaluation of the Code safety valve setpoint tolerance relaxation demonstrates that peak pressures in the primary and $ccondary systems during the limiting pressurization transient (Turbine Trip) will remain within the limits. for_ Condition 11_ events (Faults of _

hioderate Frequency). The limiting small break LOCA evenI which involves the opening of the hiSSVs is evaluated in YAEC.1847 which predicis a small increase in i the Peak Clad Temperature (PCT). The small break LOCA PCT increase is not significant as defined by 10CFR50.46 nor' are the PCT limits of 10CFR50.46 exceeded.

YAEC-1847 also demonstrates that the frequency of challenges to the Code safety

, valves will not increase nor will the reactor trip on high pressurizer pressure' be impacted as a result of the lowered setpoint tolerance ( 3%). Additionally, YAEC-1847 demonstrates that the radiological consequences associated with a Steam Generator Tube Rupture (SGTR) event are only minimally increased and remain within the round off error applied to - the SGTR radiological consequences - which were -

delineated in NilY's April 1C,1991 letter to the NRC.

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l VI. Proposed Sch*ble for I.ltense Amendment issunnse and Efferitteness New Hampshire Yankee (NilY) requests NitC review of License Amenduent Request 91-11 and issuance of a license amendment having immediate effectiveness by August  ;

31, 1992. This schedule is proposed in support of Code safety valve testing which is required to be performed pursuant to the Ni1Y inservice Test Program during the second sciueling outage which is scheduled to begin in September 1992.

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4 Vil, 1:nstronmental Imnact Assenment i

New llampshire Yankee (NilY) has reviewed the proposed license amendment against the criteria of 10CFR$1.22 for environmental considerations. The proposed changes ,

, do not involve a significant hatards consideration, nor increase the types and amounts of effluents that may be released offsite, nor significantly increase individual or cumulative occupational radiation exposures. liased on the foregoing, NilY concludes that the proposed change meets the criteria delineated in 10CFR51.22(c)(9) for a l categorical exclusion from the requirements for an Environmental Impact Statement.

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4 Ylli. Other Suppordnu information

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linclosure One:

6 Yankee Atomic Electric Company Topical Report YAEC 1847 'Seabrook Station Code-Safety Valve Setpoint Tolerance I(elaxation', I?cbruary 28, 1992.

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,t New flampshire Yankee  ;

- May 5,1992 ,

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