Text

Amend 16 to License NPF-86,revising TS to Implement Guidance of GL 90-06 Concerning Safe Plant Operation
	ML20118B048
Person / Time
Site:	Seabrook ; (NPF-86-A-016, NPF-86-A-16)
Issue date:	09/21/1992
From:	Butler W; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20118B049	List: ML20118B048; ML20118B052;
References
	GL-90-06, GL-90-6, NUDOCS 9209290332
	Download: ML20118B048 (15)
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NUCLEAR REGULATORY COMMISSION W ASHING TON, O C. 20555 s

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NORTH ATCANTIC ENERGY SERVICE CORPORATION. ET AL.*

DOCKET NO. 50-443 SEABROOK STATION. UNJT NO. 1 AMENDMENT TO FACIllTY OPERATING LICENSJ Amendment No.16 License No. NPF-86 1.

The Nuclear Regulatory Commission (the Commission or the NRC) has found that:

A.

The application for amendment filed by the North Atlantic Eneroy Service Corporation (the licensee), acting for itself and as agtnt and representative of the 11 other utilities listed below and hereafter referred to as licensees, dated October 16, 1991, compl'ies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules rnd regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the pravisions of the Act, and the rules and reguiations of the Ccmmission; C.

There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Coemission's regu1 Gions set forth in 10 CFR Chapter I; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

North Atlantic Energy Service Corporation is authorized to act as agent for the North Atlantic Energy Corporation, the Canal Electric Company, The Connecticut Light and Power Company, EUA Power Corporatinn, Hudson Light & Power Department, Massachusetts Municipal Wholesale Electric Company, Montaup Electric Company, New England Power Company, New Hampshire Electric Cooperative, Inc., Taunton Municipal Light Plant, The United Illuminating Company, and Vermont Electric Generation and Transmission Cooperative, Inc., and has exclusive responsibility and control over the physical construction, operation and maintenance of the facility.

9209290332 920921 PDR ADOCK 05000443 P

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2.

Accordingly, t.he license is amended by changes to'the T'echnical ^

Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-86 is hereby amended to read as follows:

(2)

Technical Spgcifications

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The Technical Specifications contained in Appendix A,'as revised through Amendment No.16, and the Environmental Protection. Plan contained in Appendix 8 are incorporated into Facility License No.

NPF-86.

NAESCO shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of its date of issuance and~shall be implemented within 60 days of receipt of this letter.

FOR THE NUCLEAR REGULATORY COMMISSION ~

A C< M A Walter R. Butler, Directrr Project Directorate I-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

l

' Changes to the Technical i

Specifications l

Date of Issuance:

September 21, 1992 l

l

ATTACHMENT TO LICENSE AMENDHENT NO. 16 FAtlllT1_0PERATING LICENSE NO. NPF-86 DOCKET NO. 50 4.13 Replace the following pages of the Appendix A Technical Specifications with the attached pages.

The revised pages are identified by Amendm,nt number and ccntain vertical lines indicating the area of change.

Overlap pages are provided for continuity.

Remove Insert 3/4 4-11 3/4 4-11 II 3/4 4-12 3/4 4-12 3/4 4-34 3/4 4-?4 3/4 4-34a 3/4 4-34a 3/4 4-35 3/4 4-35 B 3/4 4-2 B 3/4 4-2 B 3/4 4-2a*

B 3/4 4-15 B 3/4 4-15 m

Denotes new page

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REACTOR COOLANT SYSTEh 3/4.4.4 RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.4 Both power-operated relief valves (PORVs) and their associated biocx valves shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3.

ACTION:

a.

With one or both PORV(s) inoperable, because of excessive weat leak-l age, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the PORV(s) to OPERABLE status or close the associated block valve (s) with power maintained to the block valve (s); otherwise, be in at least HOT STAN0BY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

With one PORV inoperable due to causes other than excessive seat leakage, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either

~^ ore the PORV to OPERABLE status or close the associated block val

,nd remove power from the block valve; restore the PORV to OP L OLE status within the following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c.

Witn both PORVs inoperable due to causes other than excessive seat leakage, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore at least one PORV to OPERABLE status or close each associated block valve and remove power from the i

block valve and be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

(*.

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 l

CPERABLE status within the vxt hour if both block valves are inoperable; restore any remaining inoperable block valve to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

e.

The provisions of Specification 3.0.4 are not applicable.

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l SEABROOK - UNIT 1 3/4 4-11 Anwr.dment No.16

REACTOR COOLANT SYSTEM RELIEF VALVES 1

SURVEILLANCE 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 by:

a.

Performance of a CHANNEL CALIBRATION, and I,

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

4.4.4.2 Each block valve shall be demonstrated OPERABLE at least once per

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92 days by operating the valve through one complete cycle of full travel-unless the block valve is closed with po w remo',ed in order to meet the requirements of ACTION b. or c. in Specificat:

3.4.4.

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't SEABROOK - UNIT 1 3/4 4-12 Amendsent No.16

REACTOR CDvLANT SVSTEM PRESSURE / TEMPERATURE LIMITS PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.9.2 The pressurizer temperature shall be limited to:

a.

A maximum heatup of 100'F in any 1-hour period, b.

A maximum cooldown of 200*F in any 1-hour period, and c.

A maximum spray water temperature differential of 320*F.

APPLICABILITY:

At all times.

ACTION:

With the pressurizer temperaturc limits in excess of any of the above limits, restore the temperature to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the pressurizer; determine that the pressurizer remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the pressurizer pressure to less than 500 psig within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.9.2 The pressurizer temperatures shall be determined to be within the limits at least once per 30 minutes during systes heatup or cooldown.

The i

spray water temperature differential shall be determined to be within the limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during auxiliary spray operation.

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SEABROOK - UNIT 1 3/4 4-33 l

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REACTOR COOLANT SYSTEM PRESSURE / TEMPERATURE-LIMITS OVERPRESSURE PROTECTION SYSTEMS LIMITING C_0NDITION FOR OPERATION 3.4.9.3 The following 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 f; and in MODE 5 and MODE 6 with all Safety Injection pumps inoperable at least one of the following groups of two ove r-pressure 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 (RHR)' suction relief valves each with a setpoint of 450 psig +0, -3 %; or 2)

Two power-operated relief valves (PORVs) with lif t setpoints that vary with rcd temperature which do not exceed the limit i

established in Figure 3.4-4, or 3)

One RHR suction relief valve and one PORV with setpoints as required aoove, b.

In MODE 5 anc MODE 6 with all Safety Injection pumps except one inoperable:

1)

The Reactor Coolant System (RCS) depressurized with an RCS tent area equal to or greater thsn 18 square inches, APPLICABILIH:

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.

ACTION:

a)

In MODE 4 with all Safety Injection pumps inoperable and with one of the two required overpressure protection devices inoper-able, either restore two overpressure protection devices to OPERABLE status within 7 days or within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months (a) depressurize 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 defices inoperable, restore two overpressure protection devices to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months t

(a) depressurize the RCS anu (b) vent the RCS through at least a 1.58-square-inch vent.

i SEABROOK - UNIT 1

.3/4 4-34 Amendment No. J, ),16

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1 REACTOR COOLANT. SYSTEM PRESSURE /Te%PERATURE LIMITS OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPCRATION ACTION:

(Continued) c)

In MODE 4, MODE 5 and MODE 6 with all Safety Injection pumps inoperable and with both o. the two recuired overpressure protection devices inoperable, within the next_8 hours (a) depressurize the RCS and (b) vent the RCS through at least a 1.58-square-inch vent, d)

In the event the-POR'5s, 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 pursuant to Specification 6.8.2 within 30 days.

The report shall describe the circumstances initiating the transient, the

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effect of the PORVs, or the RHR suction relief valves, or RCS vent (s) on the transient, and any corrective action necessary to I

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 S/4 4-34a Amendment No. %,16

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PRESSURE / TEMPERATURE LIMITS OVERPRESSURE PROTECTION SYSTEMS SURVElLLANCE REQUIREMENTS j

l 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE when the PORV(s) are being i

used for overpressure protection by:

e.

Performance of an ANALOG CHANNEL OPERATIONAL TE3T on the PORV actua-l tion channel, but excluding valve operation, at least once per 31 days thereafter when the PORV is required OPERABLE; and i

h.

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

l c.

Verifying the PORV isolation valve is open at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

l i

4.4.9.3.2 Each RHR suction relief valve shall be demonstrated OPERABLE when i

tne RHR suction relief valve (s) are being used for overpressure protection l

as follows:

a.

For RHR suction relief valve RC-V89 by verifying at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months that RHR suction isolation valves RC-V87 and RC-V88 are

open, b.

For RHR suction relief valve RC-V24 by verifying at least 7nce per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months that RHR suction isolation valves RC-V22 and RC.23 are open.

c.

Testing pursuant to Specification 4.0.5.

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

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

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

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3/6.4 REACTOR COOLANT SYSTEM i

BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION i

Ine plant is aesigned tc. operate with all reactor coolant loops in opert-tion and maintain DNBR above 1.30 during all normal operations and anticipated 1

transients.

In MODES 1 and 2 with one reactor coolant loop not in operation, this specification requires that the plant be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

In MODE 3, two reacter coolant loops provide sofficient heat removal capability for removing core decay heat even in the event of a bank withdrawal i

accident; however, a single reactor coolant loop provides sufficient heat removal capacity if a bank withdrawal accident can be prevented, i.e., by opening the Reactor Trip System bre uers.

Single failure considerations require that two loops be OPERABLE at all times.

In A00E 4, and in "0DE 5 with reactor coolant loop; filled, a single I

reactor coolant loop or RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops (either RHR or RCS) be OPERABLE.

In MODE 5 with reactor coolant loops not filled, a single RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a heat removing component, require that at least two RHR loops be OPERABLE.

The operation of one reactor coolant pump (RCP) or one RMR pumD provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System.

The reactivity change rate associated with boron reducticn will, therefore, be within the capability of operator recognition and control.

The restrictions on starting an RCP in MODES 4 and 5 are provided to pre-vent RCS pressure transients, caused by energy additions from the Secondary Coolant System, which could exceed the limits of Appendix G to 10 CFR Part 50.

The RCS will be protecter against overpressure transients and will not exceed the limits of Appendix G 4 restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 50'F above each of.1e RCS cold-leg temperatures.

3/4.4.2 SAFETY VALVES The pressurizer Cude safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig.

Each safety valve is designed to relieve 420,000 lbs per hour of saturated steam at the valve Setpoint.

The relief capacity of a single safety valve is adequate to relieve any overpres-sure condition which could occur during shutdown, In the event that no safety valves are OPEMBLE, en operating RHR loop, connected to the RCS, provides over-pressure relief capability and will prevent RCS overpressurization.

In addi-tion, the Overpressure Protection System provides a diverse means of protecthn against RCS overpressurization at low temperatures, i

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RrCTOR COOLANT 59 TEM BA5f5 3/4.4.2 SAff.TY VAtVES (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 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.

3/4.4.3 PRES $URflER 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.

The maximum water volume also ensures that a steam bubble is formed and thus the RCS u not a hydraulically solid system.

The requirement that a minimum number of pres;urizer heaters be OPERABLE enhances the capability of the plent to control Reactor Coolant System pressure and establish natural circulation.

i 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 Code 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 associated block valves are powered from Class 1E power supply busses.

l The PORVs are equipped with automatic actuation circuitry and manual control capability.

The PORVs are considered OPERABLF in either the automatic or manual mode for the following seasons:

(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 (AC01 or TA00T) exists for verifying automatic operation, (3) The required ACTION for an inoperable PORV(s) (closing the block valve) conflicts with any presumed requirement for automatic actuation.

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

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The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be main-

.ined.

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

Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical i

damage or progressive degradation dut to design, manufacturing errors, or in-service conditions that lead to corrosion.

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

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REACTOR COOLANT SYSTEM l

BASES 3/4,4.9 PRESSURE / TEMPERATURE LIMITS (Continued)

COLD OVERPRESSURE PROTECTION The OPERABILITY of two PORVs, or two RHR suction relief valves, or a com-bination 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 l

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 protect the RCS from overpressurization when the transient is limited to either:

(1) the start of an idle RCP wit

the secondary water temperature of the steam p

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 for 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:

(1) 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 centrifugal charging pump while in MODES 4, 5, and 6 with the reactor vessel head installed and disallow start of an RCP if secondary coolant temperature is more than 50 F above reactor 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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

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 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 dur-ing this time, the failure of the single centrifugal charging pump is not tssumed.

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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

During low pressure, low temperature operation all automatic Safety Injection actuation signats except Containment Pressure - High are blocked.

In normal conditions, a single failure of the ESF actuation circuitry will result in the starting of atmostonetrainofSafetyInjection(onecentrifugalchargingpump,andone Safety Irjection pump).

For temperatures above '25 F, an overpressure event occurring as a result of starting two pumps can te successfully mitigated by SEABROOK - UNIT 1 B 3/4 4-15 Amendment No.

16

REACTOR COOLANT SYSTEM l

BASES 3/4.4.9 PRESSURE / TEMPERATURE LIMITS (Lontinued)

COLD OVERDRE550RE PROTECTION (Continued) operation of both PORVs without exceeding Appendix G limit, A single failure of a PORV is not assumed due to the short duration that this condition is allowed and the low probability of an event occurring during this interval in conjunction with the failure of a PORV to open.

Initiation of both trains of Safety injection during this 4-hour time frame due to operator error or a single failure occurring during testing of a redundant channel are not considered to be credible accidents.

OperationwithallcentrifugalchargingpumpsandbothSafetyInjection pumps OPERABLE is acceptable when RCS temperature is greater than 350 F, a single PORV has sufficient capacity to relieve the combined flow rate of all pumps.

Above 350"F two RCPs and all pressure safety valves are required to be OPERABLE.

Operation of an RCP eliminates the possibility of a 50'F difference existing between indicated and actual RCS temperature as a result of heat trans-port effects.

Considering instrument uncertainties only, an indicated RCS tem-perature of 350*F is sufficiently high to allow full RCS pressurization in ac-cordance with Appendix G limitations.

Should an overprecsure event occur in these conditions, the pressurizer safety valves provide acceptable and redun-dant overpressure protection.

When operating below 200'F in MODE 5 or MODE 6, Technical Specification 3.5.3.2 allows one Safety Injection pump to be made OPERABLE whenever the RCS has a vent area equal to or greater than 18 square inches.

Cold overpressure protection in this configuration is provided by the 18 square inch or greater mechanical opening in the RCS pressure boundary.

This mechanical opening is larger in size than the 1.58 square inch opening required for normal overpres-sure protection and is of sufficient size to ensure that the Appendix G limits are not exceeded when an SI pump is cperating in MODE 5 or MODE 6.

Addition-ally, when operating in a reduced inventory condition, the larger vent area limits RCS pressure during overpressure transients to reduce the possibility of adversely affecting steam generator nozzle dams.

When the reactor has been shut down for at least 7 days, the larger vent area also enhances the ability to provide a gravity feed to the RCS from the Refueling Water Storage Tank in the unlikely event that the CCP and SI pumps were unavailat,le after a loss of RHR.

The Maximum Allowed PORV Setpoint for the Cold Overpressure Mitigation i

l System will be revised on the basis of the results of examinations of reactor vessel material irradiation surveillance specimens performed as required by 10 CFR Part 50, Appendix H.

i SEABROOK - UNIT 1 8 3/4 4-16 Amendment No. 7, 5 M G 1 3 1991

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Person / Time
ML20118B048
Site:	Seabrook (NPF-86-A-016, NPF-86-A-16)
Issue date:	09/21/1992
From:	Butler W Office of Nuclear Reactor Regulation
To:
Shared Package
ML20118B049	List: ML20118B048 ML20118B052
References
GL-90-06, GL-90-6, NUDOCS 9209290332
Download: ML20118B048 (15)
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ML20118B048

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