ML20091D571

Proposed Tech Spec 3/4.4.4 Re Limiting Condition for Operation & Surveillance Requirements for Relief Valves & 3.4.9.3 Re Pressure/Temp Limits for Overpressure Protection Sys

ML20091D571
Person / Time
Site:	Seabrook
Issue date:	10/16/1991
From:	PUBLIC SERVICE CO. OF NEW HAMPSHIRE
To:
Shared Package
ML20091D568	List: ML20091D565 ML20091D569 ML20091D571
References
NUDOCS 9110250082
Download: ML20091D571 (22)

Text

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REACTOR COOLANT SYSTEM

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3/4.4.4' RELIEF VALVES LIMITING 00NOITION FOR OPERATION

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3.4.4 -WM power-operated relief valves 'DORVs) and their associated block 4

valves shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3. u, A wi,. m ,,.m u s A,

+4e f,t,er-h va la cs>

AC1 ION:

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a. With one orsebdl e PORV(s) inoperable, because of excessive seat leakage, within I hour either restore the PORV N to OPERABLE status ,

or close the associated block valve (s)1.otherwise, be in at ! cast

' ' HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in 40ttr SHUT 00WN within the '

following M hours. NOT 4 (a .

b. With one PORV inoperable due to causes other than excessive seat leakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV to OPERABLE' status or close the associated block valve and remove power from the block valve; restore the PORV to OPERABLE status within the following

' 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT STANDBY vithin the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />' and in-6&Le--

SHUTDOWN within the following40 hours. No7-G

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c. WithbothPORVis)inoperableduetocausesother han excessive seat-a * * *p leakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore eech :f th PORV W to OPERABLE status or close%ew associated block valve (piig and reme. power .

from the block valve m and be in-HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and Oste SHUTDOWN within the following M hours.

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d. ' With ne or more lock valve ) inoperable; within 1 hour:  ;

(1) estore th block valve s) to OPERAB status, or cl e the -

b1 ek valve (s and remove ower from'th ~ block valve (s) orcle i e PORV an remove powe from its ass lated solenoi valve; nd (2) apply e ACTION b. r c. above. - appropriate, or the solated dORV(s). J

e. The provisions of Specification 3.0.4 are not applicable. '

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'b Mith one or both block valu s inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> restore the de ,,0y '

bluck valve (s) to OPERABLE status or place its associated PORV(s)

--- i :: 2 :1. Restore at least cne block valve to OPERABLE status within the nut hour if both block valves are inoperable; restere

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~ any remaining inoperable block valve to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; / '

otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in liOT.SHUTOOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, f

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SEABROOK - UNIT 1 3/4 4-11 m DR eswn mou.

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REACTOR COOLANT SYSTEM

,. RELIEF VALVES t

SURVEII. LANCE REQUIREMENTS 4.4.4.1 In addition to the requirements of Specification 4.0.5,-each PORV

-shall be demonstrated OPERABLE at least once per 18 months oy:

a. Performance of a CHANNEL CALIBRATION, and

b. Doerating the valve through one complete cycle of full travel 4.4.4.2 Each block valve shall be demonstrated OPERABLE 6t least once per 92 days by operating the valve thrcuch one complete cycle of full travel unless +

the block valve is closed with powe; removed in order to meet the requirements of ACTION b. or c. in Specification 3.~4.4.

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SEABROOK - UNIT 1 3/4 4-12 M /

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REACTOR COOLANT SYSTEM t

' PRESSURE / TEMPERA (URE LIMITS Va /- c an,L cne Po&  !

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QF OVERPRESSURE PROTECTION SYSTEMS y , ce d. 4 4 g.-

~ LIMITING CONDITION FOR OPERATION 3.4.9.3 The following Overpressure Protection Systems shall ce OPERABLE: I L '

a .' in MODE 4 when the temperature of any RCS cold leg is less than or equal to 329'F; and in MODE ,

pumr at s ino as.etab,ied,sn u'"'cra*^6 dl

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• ' ' MODE,6 N'* *

• with * * * *

• allW Safe y Injectio i

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1) Two residual heat removal (RHR) suc ion relief valves each with a setpoint of 450 psig +0, -3 %; or '

2) Two power-operated relief valves (PORVs) with lift setpoints '

that vary with RCS temperature which do not exceed the limit established in Figure 3.4-4, or '

3) -The-Reseter-Coolant-Sys tem-(RC4-depre s s u r4a ed4th-an-RCS-vent-area of greater than er-equal to 1.58-square-inchen .

b. In H0DE 5 and MODE 6 with all Safety Injection pumps except one inoperable:

1) The Reactor Coolant System (RCS) depressurized with an RC$

vent area equal to or greater than 18 square inches.

s h APPLICABILITY: MODE 4 when the temperature of any RCS cold leg is less than*

, Qt! or equal to 3295r; MODE 5 and MODE 5 with the reactor vessel head on, i,CTION:

4: In MT4-4r-H00E S :nd MODE 6 with-all-Safety-Inject 4en-pumps-4coperaMe+

.7,, m n 4 m v4 a tt Sa fe yt We dio s p.,s s h op << L le- ar.J / pro

,ve,lassorel em" A g) pithonegPORY:nd-one euo suct4cn-re14ef-valve inoperable

""R -" o [=a'" *" I "i

ar tA4*+w.h" A1ther restore two PORVs er te: OPERABLE status within r 'i-*

7 days on%

• yf , ou at least a 1.58-square-inch vent Q essuriz d t the RCS Fc ")(;

{h" '"b codepressuriz M. .  ??% 3 ' $2 *N5o'%Dl. 'd. k.*W.

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inoperableF . . l the RCS through at least a 1.58-squaro-inch vent in 8Jo,u,r,s,.

/)) In the event the P ,

or the RHR suction relief valves, or the RCS vent (s) are used to mitigate an RCS pressure transient, l el d st <4 a Special Report-shall be prepared and submitted to the Cmis sion pursuant to Specification 6.8.2 within 30 days. The ' ' ort %j/rNGresse e ,

shall describe the circumstances initiating ti.. transient, the (g/8 4'y effect of the PORVs, or the RHR suction relis 91ves, or RCS --

vent (s) on the transient, and any correctir 'n la asary to prevent recurrence.

EpsttT SEABROOK - UNIT 1 3/4 4-34 A.nendment No. 3, 5

, R_iACTORCOOLANTSYSTEM PRESSUWEREMPERATURE LIMITS OVERPRESSURE PRO H CTION SYSTEMS LIMITING _ CONDITION FOR OPERATION ACTION: (Continued) e 4.' In MODE 5 and MODE 6 with all Safety Injection pumps except one inoperablep

-it (With the RCS vent area less than 18 square inches, immediately restore all Safety Injection pumps to inoperable status.

2 ~. /u s e y 7 ~ 4 l> . 1~n h c DS 5 a nL G aW n //  % fe fy [n j *Ab n in ope r al/u .o n ,L wiY4 on e e/ H.- Le ll~ pumps h ~'u s

,, e g u ; ~.,L e ae,y re s s ur - proh &<'

j ,, o y c e d k , e e s h <* % o e v e r, pre ssu.ce fr*~/cif*b n p vi a.t s de> of%t A NC s%h c '

ui}/ i s, p.y 4 e a <S 0r w i+ h ( *< fLl n L2 f [ k s ca r s c's > sl? r ' 5 % " r~s'\.<-

-bk e- R c. s a,,L cD venf t < <. Res l 4, ou 5 4 oF /44s1 cc /. ff sguoee f6.4 5 e n *,

x, SEABROOK - UNIT 1 3/4 4-34a Amendment No. :s

RENCTORCOOLANTSYSTEM u le n PA t- Po R V t'n a r e

, j(' PRESSURE /TEMPERATURELIMITS f e ,',3 u u.c/ /o r o v e 7ec tsure

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• OVERPRESSURE PROTECTION SYSTEMS SURVEILLANCE REQUIREMENTS _ _ _ _ _

, f 4.4.9.3.1 Each PORV-shall be demonstrated OPERABLE ,by:

a. Performance of an ANALOG CHANNEL OPERATIONAL TEST on the PORY actuation channel, but excluding valve operation, n 6:

4. +. .... 4 . ,--a4+4m. <a us4,6 +w nnou 4, '_i+ki-

__a_a nenio, e N at least once per 31 days thereafter when thi P0It0'Esle~'ih[d qu OPERABLE; cos'

b. Performance of a CHANNEL CAllBRATION on the PORV actuation channel at least once par 18 months; and

c. Verifying the PORV isolation valv6 is open at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

.6:r. the POPu 5 ge e; g e :a y e r _ , _ , , , _ _ _ _ , , ,, , , ,,, , , ,,,,

4.4.9.3.2 Each RHR suction relief valve shall be demonstrated OPERABLE when the RHR suction relief valves)are being used for eeM overpressure protection as follows:

n .

V ,' a. For RHR suction relief valve RC-V89 by verifying at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> that RHR suction isolation valves RC V87 and RC-V88 are open.

b. For RHR suction relief valve RC-V24 by verifying at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> that RHR suction isolation valves RC-V22 and RC-V23 are open.

c. Testing pursuant to Specification 4.0.5.

4.4.9.3.3 The RCS vent (s) shall be verified to be open at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />

• when the vent (s) is being used for overpressure protection.

• Except when the vent pathway is provided with a valve (s) or devices (s) that is locked, sealed, or.otherwise secured in the open position, then verify this e valve (s) or_ device (s) open at least once per 31 days.

3/4 4-35 Amendment No. 3, 5, SEABROOK - UNIT 1

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" '[R$ ACTOR COOLANT SYSTEM i BASES' gJ.s l

3/4.4.2 SAFETV VALVES (Continued) '

During operation, all pressurizer Code safety valves must be OPERABLE to prevent the RCS from being pressurized above its Safety Limit of 2735 psig. .

The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss of load assuming no Reactur trip until the first Reactor Trip System Trip Setpoint is reache'd (i.e. , no credit '

is taken for a direct Reactor trip on the loss of load) and also assuming no '

operation of the power-operated relief valves or steam dump valves.

Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI i of the ASME Boller and Pressure Code. .

3/4.4.3 PRESSURIZER The limit on the-maximum water volume in the pressurizer assures that the i parameter is maintained within the normal steady-state envelope of operation  ;

-assumed in the SAR. The limit is consistent with the initial SAR assumptions.

The 12-hour periodic surveillance is sufficient to ensure that the parameter is restored to within its limit following expected transient operation. The maximum water volume also enfures that a steam bubble is formed and thus the RCS is not a hydraulically, solid system. The requirement that a minimum-

,w number of pressurizer heaters be OPERABLE enhances the capability of the plant Wfg to control Reactor Coolant System pressure and establish natural circulation.

3/4.4.4 RELIEF VALVES The power-operated relief valves (PORVs) and steam bubble function to re-lieve RCS pressure during all design transients up to and including the design step load decrease with steam dump. Operation of'the PORVs minimizes the undesirable opening of the spring-loaded pressurizer Coda safety valves. Each PORV has a remotely operated block valve to provide a positive shutoff capabil-ity should.a relief valve become inoperable. The PORVs and their arsociated block valves are powered from Class 1E power supply busses. ,

3/s.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes -

ensure that the structural integrity of this portion of the RCS will be main- -

tained. The program for inservice inspection cf steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1. Inservice inspection  !

of steam generator cubing is essential in order to maintain surveillance of the '

conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or in-service conditions that lead to corrosion. Inservice inspection of steam generator tubing 61so provides a means of characterizing the nature and cause of any tube degradation, so that corrective measures can be taken. +

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SEABROOK - UNIT 1 B 3/4 4-2

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~ The PORVs are equipped win automatic actuation circuitry and manual control capability. The PORVs are considered OP11RAtlLE in either the automatic or mhnual mode '

for the following reasons:

(1) No credit is taken in any FSAR accident aralysis for automatic PORV- '

actuation to mitigate the consequences of an accident.

(2) No Surveillance Requirement (ACOT or TADOT) exist for verifying autornatic operation.

(3) -The required ACTION for an inoperable PORV(s) (closing the block valve) conflicts with any presumed requirement for automatic actuation. t-

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. Or 4_. c. ,. b e b d c REACTOR COOLANT SYSTEM '

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3/4.4.9 PRESSURE / TEMPERATURE 11MITS(Continued) ,

COLD OVERPRESSURE DROTECTION The OPERABILITY of two PORVs, or two RHR suction relief valves, or an RCS vent opening of at least 1.58 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of i

to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 329'F. Either PORV or either RHR suction relief valve has ac' equate relieving capability to protect the RCS frem overpressurization when the

-transient is limited to either: (1) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to 50*F above the RCS cold leg temperatures, or (2) the start of a centrifugal charging pump and its injection into a water-solid RCS.

The Maximum Allowed PORY Satpoint for the Cold Overpressure Mitigation System (COMS) is derived by analysis which models the perfomance of the COMS assuming various mass input and heat input transients. Operation with a PORV Setpoint-less than or equal to the r.aximum Setpoint ensures that Appendix G criteria will not be violated with consideration for: (1) a maximum pres-sure overshoot beyone the PORV Setpoint which can occur as a result of time de-lays in_ signal processing and valve opening; (2) a 50*F heat transport effect made possible by the geometrical relationship of the RHR suction line and the W"". RCS wide range temperature indicator used.for COMS; (3) instrument uncertain- t ties; and (4) single failure. To ensure mass and heat input transients more severe than those assumed cannot occur, Technical Specifications require lock-out of both Safety Injection pumps and all but one centrifugal charging pump while in MODES 4, 5, and 6 rlth the reactor vessel head installed and disallow start of an RCP if secondary coolant temperature is more than 50'F above reac-tor coolant temperature. Exceptions to these requirements are acceptable as described below.

Operation above 350'F but less than 375'F with only centrifugal charging pump OPERABLE and no Safety Injection pumps OPERABLE is allowed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. As shown by analysis, LOCAs occurring at low temperature, low pres-sure conditions can be successfully mitigated by the operation of a single centrifugal charging pump and a single RHR pump with no credit for acct.aulator injection. Given the short time duration and the condition of having only o%

centrifugal charging pump OPERAELE cnd the probability of a LOCA occurring our-ing this time, the failure of the single centrifugal charging pump is not assumed.

Operation below 350'F but greater than 325'F with all centrifugal charging and Safety Injection pumps OPERABLE is alleved for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. During low pressure, low temperature operation all automatic Safety Injection actuatic.

signals except Containment Pressure - High are blocked. In normal condftions, ,

a single failure of the EST actuation circuitry will result in the staFting of o atmostonetrainofSafetylujection(onecentrifugalchargingpump,andone l Safety Injection pump). For temperatures above 325 F, an overpressure event occurring as a result of starting two pumps can be successfully sitigated by SEAe.00K - UNIT B 3/4 4-15 1 nnenament No. 3 1 ~'

i l Jul. I 51991

_ _ _ . _ ;_.__m_ __ _ ___- _ _ _ _ __. _ _ _ _ _

. .._ _ _ - _ . . _ _ . _ _ _ _ . - _ _- _. ._ _._m. - . . __.

_ ____.m.._,___.

l.4 b?  ?

j 111. Retyne of Proposed Chnnees_ j 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 ik.e right hand margin to designate a change in the text. '

No revision bars are utilized when the page is changed solely to accommodate the shifting of text due to additions or deletions.  !

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REACTOR COOLANT SYSTEM 3/4.4.4 PEllEF VALVES LIMITING CONDITION FOR OPERATION 3.4.4 Both power-operated relief valves (PORVs) and their associated block

. valves shall be OPERABLE.

APPLICABILITY: MODES 1. 2, and 3.

ACTION:

a. With one or both PORV(s) inoperable, because of excessive seat l leakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV(s) to ODERABLE status or close the associated block valve (s) with power maintained to the block valve (s); otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in 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 />.

b. With one PORV inoperable ldue to causes other than excessive seat leakage, within I hour either restore the PORV to OPERABLE status ,

or close the associated block valve and remove power from the block valve; restore the PORV to OPERABLE status within the following 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in 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 />. l

c. With both PORVs inoperable due to causes other than excessive seat leakage, within I hour either restore at least one PORV to OPERABLE status or close each associated block valve and remove power from the block valve and be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and H0T SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

d. With one or both block valves inoperable, within I hour restore the block _ valve (s) to OPERABLE status or place its associated PORV(s) control switch to "CLOSE". Restore at least one block-valve to OPERABLE status within the n xt hour if both block valves L are inoperable; restore any remaining inoperable block valve ta

operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; otherwise, be at least HOT

. STANDBY.within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the L following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,

e. The provisions of Specification 3.0.4 ere not applicable.

SEABROOK - UNIT 1 3/4 4-11 Amendment No.

I

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. REACTOR C00 TANT SYSTEM-  :

i RftlEF VALVES SVRVEILLANCE RE0VIREMENTS 4.4.4.1 In addition to the requirements of Specification 4.0.5, each PORV shall be demonstrated OPERABLE at-least once per 18 months by:

a. Performance of a CHANNEL CAllBRATION, and

b. Operating the valve through one complete cycle of full travel l during MODES 3 or 4.

4.4.4.2 Each block valve shall be demonstrated OPERABLE at least once' per 92 days by operating the valve through one complete cycle of full- travel unless the block valve-is closed with power removed in order to meet the requirements of ACTION b. or c. in Specification 3.4.4.

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SEABROOK UNIT 1 3/4 4-12 Amendment No.

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, REACTOR COOLANT SYST @

ERESSURE/TEMPERATUREtIMITS OvtRPRESSURE PROTECTION SYSTEMS  ;

LIMITING CONDITION FOR OPERATION ,

3.4.9.3 The foll9 wing Overpressure Protection Systems shall be OPERABLE:

a. In MODE 4 when the temperature of any RCS cold leg is less than or equal to 329'T and in MODE 5 and MODE 6 with all Safety injection pumps inoperabic at least ore of the following groups of two overpressure protection devices shall be OPERABLE when the RCS is not depressurized with an RCS vent area of greater than or equal to 1.58 square inches:

1) Two residual heat removal with a setpoint of 450 psig(RHR) 40, -3 suction

%; or relief valves each

2) Two power opertted relief valves (PORVS) with lift setpoints that vary with RCS temperaturr which do not exceed the limit established in Figure 3.4-4, or

3) One RHR suction relief valve and one PORV with setpoints as required above.

b. In MODE 5 and MODE 6 with all Safety injection pumps except one inoperabic: -

1) The Reactor Coolant System (RCS) depressurized with an RCS vent area equal to or greater than 18 square inches.

APPLICA8JilLY: MODE 4 when the temperature of any RCS cold leg is less than or equal to 329'F; MODE 5 and MODE 6 with the reactor vessel head on.

ACllDft(:

a) In MODE 4 with all Shiety Injection pumps inoperable and with one of the two required overpressure protection devices inoperable. either restore two overpressure protection devices to 0PERABLE status within 7 days or within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (a) d(pressurize the RCS and (b) vent the RCS through at least a 1.58 square-inch vent.

b) In MODE 5 and MODE 6 with all Safety injection pumps inoperable and with one of the two required overpressure protection devices inoperable restore two overpressure protectiondevicestoOPERABLEstatuswithin24hoursor within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (a) depressurize the RCS and (b) vent the RCS through at least a 1.58 square-inch vent.

SEABROOK UNIT 1 3/4 4-34 Amendment No. 4, 4, l

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REACTOR C00lAN1 SYSTEM PRESSURE / TEMPERATURE LIMITS OVERPRESSURE PROTECTION SYSTEMS

' LIMITING CONDITION FOR OPERATION ACTION:. (Continued) c) In MODE 4, MODE 5 and MODE 6 with all Safety Injection pumps inoperable and with both of the two required overpressure protection devices inoperable, within the next'8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (a) depressurize the RCS and (b) vent the RCS through at least a 1.58-square-inch vent.

d) In the event- the PORVS,- or the RHR suction relief valves, or-the RCS vent (s) are used to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursucnt to Specification 6.8.2 within 30 days. The. report shall describe the circumstances initiating the transient, the effect of the PORVS, or the RHR suction relief valves, or RCS vent (s) on the transient, and any corrective action necessary to prevent recurrence.

e) In MODE 5 and MODE 6 with all Safety injection pumps except one inoperable and with the RCS-vent area less than 18 square inches, immediately restore all Safety Injection -

pumps to inoperable status.

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- SEABROOK UNIT 1 3/4 4-34a Amendment No. 6 i-

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. 1 1, REACTOR C00LRNT SYSTEM PRESSURE /TEMPERATUREllMilE  !

OVERPRESSURE PROTECTION SYSTEME l SURVEILLANCE REQUIREMENTS _ ,

s 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE when the PORV(s) are being used for overpressure protection by:

a. Performance of an ANALOG CHANNEL OPERATIONAL TEST on the PORV actuation channel,-but excluding valve operation, at least once l per 31 days when the PORV is required OPERABLE; and j

b. Performance of a CHANNEL CALIBRATION on the PORV actuation channel at least once por 18 months; and

c. Verifying the PORV isolation valve is open at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

4.4.9.3.2 Each RHR suction relief valve shall be demonstrated OPERABLE when L the RHR suction relief valve (s) are being used for overpressure, protection as

. follows:

- a. For RHR suction relief valve RC-V89 by verifying at least once per -

72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> that RHR suction isolation valves RC-V87 and RC-V88 are open,

b. For RHR suction relief valve RC-V24 by verifying at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> that RHR suction isolation valves RC-V22 and RC-V23 are i open. ,

. c. . Testing pursuant to Specification 4.0.5.

4.4.9.3.3 The RCS vent (s) shall be verified to be open at least once per 12

' hours

• when-the vent (s) is being used for overpressure protection.

• Except when the vent pathway is provided with a valve (s) or devices (s) that is locked, sealed, or otherwise secured in the open position, then verify this >

l valve (s) or device (s) open at least once per 31 days.

SEABROOK - UNIT 1 3/4 4-35 Amendment No. 4, 6,

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,' REACTOR COOLANT SYSTER BASES 3/4.4.2 SAFETY VALVES (Continued)

During operation, all pressurizer Code safety valves must be OPERABLE to prevent the RCS from being pressurized above its Safety Limit of 2735 psig.

The combined relief capacity of all of these valves is greater than the maximum surge rate resultliig from a complete loss of load assuming 2.0 Reactor trip until the first Reactor Trip System Trip Setpoint is reached (i.e., no credit is taken for a direct Reactor trip on the loss of load) and also assuming no operation of the power-operated relief valves or steam dump valves.

Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code.

.]/4.4.3 PRESSVRIZER The limit on the maximum water volume in the pressurizer assures that the parameter is maintained within the normal steady-state envelope of operation assumed in the SAR. The limit is consistent with the initial SAR assumptions. The 12-hour periodic surveillance is sufficient to ensure that -

the parameter is restored to within its limit following expected transient operation. Th maximum water volume also ensures that a steam bubble is formed and thut the RCS is not a hydraulically solid system. The requirement that a minimum number of pressurizer heaters be OPERABLE enhances the capability of the plant to control Reactor Coolant System pressure and establish natural circulation.

3/4.4.4 REllEF VALVES The power-operated relief valves (PORVS) and steam bubble function to relieve RCS pressure during all design transients up to pnd including the design rtop load decrease with steam dump. Operation of the PORVs minimizes the undesirable upening of the spring-loaded pressurizer Code safety valves.

Each PORV has a remotely operated block valve to previde-a positive shutoff capability should a relief valve become inoperable. The PORVs and their associated block valves are powered from Class IE-power supply busses. ,

The' PORVs are equipped with automatic actuation circuitry and manual control capability. The PORVs are considered OPERABLE in either the automatic or manual mode for the following reasons:

(1) No credit is taken in any FSAR accident analysis for automatic PORV actuation to mitigate the consequences of an accident.

(2) No Surveillance Requirement (ACOT or TAD 0T) exist for verifying automatic operation.

(3) The required ACTION for an inoperable PORV(s) (closing the block valve) conflicts with any presumed requirement for autcmatic actuation.

SEABROOK - UNIT 1 B 3/4 4-2 Amendment No.

PEACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained. The program for inservice inspection of steam generator tubes is based an a .acdification of Regulatory Guide 1.83, Revision 1. Inservice in.pection of Men generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion.

Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause af any tube degradation, so that corrective measures can be taken.

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

REALTOR COOLANT SYSTEM BASES

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3/4.4.9 PRESSURF/ TEMPERATURE LIMITS (Continued)

.CSLD OVERPRESSURE PROTECTION The OPERABILITY of two PORVs, or two RHR suction relief valves, or a combination of a PORV and RHR suction relief valve, or an RCS vent opening of

- at least 1.58-square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 329'F.

Either PORV or either RHR suction relief valve has adequate relieving

- capability-to 3rotect the RCS from overpressurization when the transient is

- limited to eitler: (1) the start of an idle RCP with the secondary water temperature of the steam generator.less than or equal to 50'F. above the RCS cold -leg temperatures, or (2) the start of a centrifugal charging pump and its injection into a water-solid RCS.

The Maximum Allowed PORV Setpoint fer the Cold Overpressure Mitigation System (COMS) is derived by. analysis which models the performance of the COMS assuming various mass input and heat' input transients. Operation with a PORV Setpoint less than.or equal to the maximum Setpoint ensures that Appendix G criteria will,not be. violated with consideration for:-(l) a maximum pressure overshoot beyond the PORV Setpoint which can occur as a result of time delays in signal processing and valve opening; (2) a 50*F heat transport effect made possible by the geometrical relationship of the RHR suction line and the RCS wide range temperature indicator. used for COMS; (3) instrument uncertainties; and;(4) single failure. To ensure mass and heat input transients more severe

~

than those assumed cannot occur, Technical Specifications require. lockout of

- both Safety Injection pumps and all but one centriWal charging pump while in

- MODES- 4, 5. and 6 with the reactor vessel head inst iled and disallow start of an RCP if secondary coolant temperature is more- than 50*F above reactor '

coolant temperature. Exceptions i.o these requirements are acceptable as described below.

Operation above 350'F but less than 375'F with only centrifugal charging pump _0PERABLE-and no Safety Injection puropF OPERABLE is allowed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. As-shown by analysis, LOCAs occurring at low temperature, low pressure conditions can be successfully mitigated by the operation of a single centrifugal charging pump and a single RHR pump with no credit for accumulator injection. Given the short time duration and the condition of having only one centrifugal charging pump OPERABLE and the probability of a LOCA occurring during this time, the failure of the single centrifugal charging pump is not-assumed.

Operation below 350'F but greater than 325'F with all centrifugal charging and Safety Injection pumps OPERABLE is allowed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

During low pressure, low temperature operation all automatic Safety Injection actuation signals except Containment Pressure - High are blocked. In normal conditions, a single failure of the ESF actuation circuitry will result in the starting of at most one train of Safety injection (one centrifugal charging pump, and one Safety injection pump). For temperatures above 325'F, an overpressure event occurring as a result of starting two pumps can be succersfully mitigated by I SEABROOK - UNIT 1 B 3/4 4-15 Amendment No. 4 i

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. ~ _

-IV, Safety Evaluation of License Amendment Reouest 91-08 Pronosed Chances The proposed Technical Specifications changes are consistent with USNRC Generic Letter (OL) 90 06. OL 90 06 requests changes to Seabrook Station Technical Specification 3/4.4.4 " Relief Valves" and- its associated Bases . and Technical Specification 3/4.4.9.3

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" Overpressure Protection Systems" and its associated Bases. The proposed revisions to L Technical Specification 3.4.4 ACTION a. require that power be maintained to block valves

! which are - closed to isolate PORVs. which 'are exhibiting excessive seat leakage. By maintaining power to closed block valves when the PORVs are exhibiting excessive seat leakage the block valves can be readily opened to afford use of the PORVs in mitigating RCS ' pressure transients. If the block valves are inoperable, the proposed' revisions to

. ACTION d. [rovide adequate measures to assure that a PORY will not.become stuck open when a block valve is inoperable yet maintains the ability to use the PORVs for transient _ mitigation. The _ changes to Technical Specification 3/4.4.4. are intended to enhance the availability of the PORVs for mitigation of RCS pressure transients.

The _ changes to Technical Specification 3/4.4.9.3 provide enhanced operational flexibility L through the use 'of a PORV in combination with a RI-IR suction relief valve for low L .teinperature overpresserc protection. Each of these relief valves, alone is capable of mitigating a design basis mass or heat addition transient as stated in the Bases for Technical Specification 3/4.4.9.3. The proposed revisions to ACTION b. reduce the allowed outage time for one of the two required overpressure protection devices from '7 days to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when in MODE 5_or 6, because the NRC has determined based on its i review 'of low temperature overpressure transients that the potential for-en overpressure transient is highest in these MODES; The proposed changes to Technical Specifications 3/4.4.4. and 3/4.4.9.3 do not affect the functions of the PORVs in MODE 1, 2 or 3 or the Overpressure Protection System functions required in MODE 4 below 329 F MODE 5 and MODE 6; There is no change g proposed to the PORV actuation circuitry or to the PORV or block valve power supply l configuration. There is no reducti_on in surveillance testing of the PORVs or Overpressure Protection Systems.

The proposed changes will result in an improvement in the availability of the PORVs and i' Overpressure Protection Systems-to mitigate RCS pressure transients and will therefore j; enhance safe operation.

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W V, Determination of Sienificant Ilarards for Iicense Amendment Reonest 91-08 Proonsed Chances

- (1) The - proposed changes Tdo~ not: involve a significant increase in the - probability or

consequences of an accident previously evaluated.

The proposed changes to Technical Specification 3/4.4.4 are intended to increase the availability of the PORVs to mitigate RCS pressure transients. The proposed changes will not increase the probability of occurrence of. an inadvertent opening of a l pressurizer safety or relief valve which ^ are analyzed in the Updated Final Safety Analysis Report-_(FSAR) Section 15.6.1. There is no change proposed to_ the PORV actuation circuitry or to the PORV or block valve power supply configuration.

Tlic proposed' changes to Technical Specification 3/4.4.9.3 are intended to increase the availability of equipment which is utilized to mitigate low temperature overpressure transients,.by reducing the allcwed outage time for such equipment in MODE 5 and 6,-

Design 13 asis low temperature overpressure . transients are initiated by dvertent mass

- additions (e.g. a- charging pump or Safety injection (SI) pump) with no letdown, or by heat*

additions caused by the starting of an idle RCP with the secondary side more than

!50 F wurmer than the primary.- The proposed : changes to Technical Specification 3/4.4.9.3, will' have no affact _on the probability of occurrence of low temperature overpressure transient. The changes will improve the availability of equipment utilized to mitigate such a tiansient.

There is 'no increase- in the consequences of an accident previously evaluated in the FSAR. The only accident analysis which take credit for the PORVs in mitigating the accident and its consequences is the Steam Generator Tube Rupture (SGTR) Accident.

The FSAR description of this accident has been ~ superseded by the SGTR analysis submitted to the NRC on April 16, 1991, in this analysis the' PORVs are assumed to be utilized by the _ operators as required by Emergency Response Procedure E-3 to reduce RCS pressure acid thus terminate flow to the faulted Steam Generator. The proposed- changes to Technical Specification 3/4.4.4 which improve the availability.of the PORVs, will not Lincrease the consequences of an accident previously evaluated (i.e._ the SGTR accident).

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

'The proposed changes to Technical bpecifications 3/4.4.4 and 3/4.4.9.3 do not affect the i lL functions of the PORVs in MODE 1, 2 or 3 or the Overpressure Protection System functions required in MODE 4 below 129 F, MODE 5 and MODE 6. There is no change proposed to the PORV actuation circuitry or to the PORV or block valve power supply configuration _ - There is no reduction in surveillance testing of the PORVs or Overpressure Protection Systems, p l-L The proposed changes will result in an improvement in the availability of the PORVs and Overpressure Protection Systems to mitigate RCS pressure transients.

8 I

l-I i -

_ -_ ~ _ _ _

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

The margin of safety associated with the PORVs or the Overpressure Protection Systems is defined in the Bases for their corresponding Technical Specifications.

The Bases for Technical Specification 3/4.4.4 reads as follows:

3/4.4.4. REllEF VALVES The power-operated relief valves (PORVs) and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump. Operation of the PORVs minimizes the undesirable opening of the spring loaded pressurizer Code safety valves. Each PORV has a remotely operated block valve to provide a positive shutoff capability should a relief valve become inoperable. The PORVs and their associated block valves are powered from Class 1E power supply busses.

The proposed changes to Technical Specification 3/4.4.4 do not reduce the margin of safety defined in its Bases. The function of the PORVs and their block valves is not changed. The Bates for Technical Specification 3/4.4.4 is proposed to be clarified by specifying that automatie operation of the PORVs is not credited in any MODE 1, 2 or 3 transient and therefore the PORVs can be considered operable in either the automatic or manual mode.

The Bases for Technical Specification 3/4.4.9 in pertinent part reads as follows:

3/4.4,9 PRESSURE / TEMPER ATUR E 1.!MITS COLD OVERPRESSURE PROTECTION The OPERABILITY of two PORVs, or two RilR suction relief valves, or an RCS vent opening of at least 1.58 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix 0 to 10, CFR Part 50 when one or more of the RCS cold legs are less than or equal to 329 F. Either PORV or either RilR suction relief valve has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either: (1) the start with the secondary water temperature of the steam generator less than of an idle or equal to RCllF 50 above the RCS cold leg temperature, or (2) the start of a centrifugal charging pump and its injection into a water-solid RCS.

The proposed changes to Technical Specification 3/4.4.9.3 do not reduce the margin of safety defined in its Bases. The changes will enhance the availability of the Overpressure Protection System devices by reducing the current allowed outage time in MODE 5 or 6 from 7 days to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, thereby providing a greater level of safety.

The Bases for Technical Specification 3/4.4.9.3 will be clarified by stating that a PORV in combination with a RHR suction relief valve is an acceptable configuration for cold overprer,sure protection. This combination is acceptable because, as indicated in the current Bases, "Either PORV or either RilR suction relief valve has adequate relieving capability to protect the RCS from overpressurization..

• In view of the preceding, NilY has determined that the Technical Specification changes proposed in License Amendment Request 91-08 do not involve a significant hazards consideration.

9

VI. Proposed Schedule for License Amendment luuance and Effectiveness l New Hampshire Yankee requests NRC review of License Amendment Request 91-08 and issnance of a license amendment having immediate effectiveness by March 31, 1992.

As specified in Section 1, the Technical Specification changes proposed herein were developed by a group of seven utilities comprising eight Westinghouse PWRs which utilire the PORVs or RilR suction relief valves for low temperature overpressure protection.

NHY recommends that its License Amendment Request 91-08 be reviewed in conjunction with the reviews of the aforementioned group of licensecs.

The Technical Specification changes proposed herein will enhance safe operation by improving the availability of safety related equipment, the Power-Operated Relief Valves and Overpressure Protection Systems, which may be utilized in the mitigation of transients.

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Vll.Other Sunnortine Documentation i

NIIY Letter NYN 90217, " Response to Generic Letter 90 06" dated December 21, 1990 to USNRC.

11

Nev[ Hampshire Y

Ted C. Feigenbovm (h Preddent and Chief Execu6e Officer a

NYN 90217 Decemoer 21, 1990 United States Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Document Control Desk

References:

(a) Facility Operating License No. NPF-86, Docket No. 50-443 (b) USNRC Generic Letter 90 06, dated June 25,1990, " Resolution of Ger eric Issue 70, ' Power operated Relief Valve and Block Valve Reliability', and Generic Issue 94, ' Additional Low-Temperature Overpressure Protection For Light-Water Reactors', Pursuant to 10 CFR 50.54(f)'

S ubject: Response to Generic Letter 90 06 Gentlemen:

New Hampshire Yanket, provides the following information in response to Generic Letter 90 06 regarding the pressurizer power-operated relief valves (PORVs) and block valves installed in Seabrook Station. This information addresses desigt., programmatic and technical specification requirements applicable to the PORVs and block valves.

The Seabrook Station PORVs are solenoid valve controlled, pressure actuated, poppet-type relief valves which do not depend upon control air systems for their operation. The block valves are flexible wedge, rising stem, motor-operated gate valvesJ The Seabrook Station PORVs, block valves and associtted components were designed to meet safety grade requirements. Mechanically, the PORVs and block valves are Safety Class I components. Electrically, the PORV solenoids and the block valve motor-operators are designated Class 1E. The PORV and block valve control switches, selector switches and position ind! cation are also designated Class 1E. Other electrical components in the PORV and block valve circuitry including cables, connectors and splices that are located in harsh environments are environmentally qualified. The PORV circuitry includes contacts from relays driven by non-safety-related signals associated with automatic control functions. These relays are not qualified but are similar to Class IE relays. An analysis ha', been performed to demonstrate that no credible failure of these relays can create a condition which degrades the function of the Class 1E portion of the PORV circuitry.

n.M j [7  :

l l New Hampshire Yankee Division of Public Service Company of New Hampshire P.CLAqs300

• Sabreek. NH 03874

• Tcacham M031L4769121 ,

t United States Nuclear Regulatory Commission December 21, 1990 Attention: . Document Control Desk Page two 1

The PORVs and block valves at Seabrook Station are included within the scope of the NHY Operational Quality Assurance Program and are therefore listed in FSAR Table 3.22. The following programmatic requirements applicable to the PORVs and bleck valves have been irople:nented. Approved procedures implement maint ' nance re,uiremet.ts for the PORVs and block valves in accordance with manufacturer's recommendations. Maintenance is performed by trained maintenance personnel. The PORVs and block valves are included within the scope of the Inservice Testing Program for valves. The block valves are included within the scope of the Motor-Operated Valve Testing Program being developed-in response to NRC Generic Letter 89-10. Complete replacement PORVs or block valves would be -

purchased, if needed, in accordance with the requirements of the original construction specification including applicable amendments. Replacement parts for the PORVs or block valves would be purchased, if needed, by reference to the original manufac:urer or supplier's part number. Additionally, purchasing requirements have been pre established for specific valve parts based upon a review of the safety function of each part with input from the original rnanufacturer and/or supplier. These requirements are imposed by the parts purchase order.

New Hampshire Yankee intends to propose plant-spccific technical specification changes regarding PORV operability. The proposed changes will be based upon the model technical specifications provided as Attachment A-1 of Enclosure A to Generic Letter 90-

06. Additionally, NHY intends to propose plant-specific technical specification changes relating to low temperature overpressure protection. These proposed changes will be based upon the model technical specifications provided as Attachment B 1 to Enclosure B of Generic Letter 90 06, but will reflect the fact that either the residual heat removal system (RHRS) suction relief valves or the PORVs can provide the required low temperature overpressure protection, The proposed technical specification changes described above will be submitted prior to startup following the first refueling outage.

New Hampshire Yankee is working with six other utilities in developing a common approach to the proposed technical specification changes associated with Generic Letter 90-

06. The plants involved in this effort are: Wolf Creek, Vogtle, Comanche Peak, Millstone 3, Seabrook, Byron, Braidwood, and Callaway. A joint effort is facilitated by the similarity of plant types and technical specifications. All the plants in the group are Westinghouse PWRs which utilize the PORVs and RHRS suction relief valves for low temperature everpressure protection.

If you have any questions on this matter, please contact Mr. Geoffrey Kingston at (603)474-9521, Extension 3371.

Very truly yours, gh # --

Ted C. Feigenbaum TCF:GK/ssi