ML030300440

Request for Amendment to TS Emergency Core Cooling System Requirements During Shutdown Conditions

ML030300440
Person / Time
Site:	Pilgrim
Issue date:	01/23/2003
From:	Bellamy R Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LER 2001-003-00, LER 2001-006-00, LER 2001-007-00
Download: ML030300440 (31)

Text

AE I 'En terqy Entergy Nuclear Operations, Inc.

Pilgnm Nuclear Power Station 600 Rocky Hill Road Plymouth, MA 02360 Mike Bellamy Site Vice President January 23, 2003 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555-0001

SUBJECT:

Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station Docket No. 50-293 License No. DPR-35 Request for Amendment to the Technical Specifications Emergency Core Cooling System Requirements During Shutdown Conditions

REFERENCE:

1. NUREG 1433, Standard Technical Specifications for General Electric Plants, BWR/4.

2. Entergy letter dated June 20, 2001, LER 2001-003-00, ESF Actuations Due to Invalid Water Level Indications.

3. Entergy letter dated October 12, 2001, LER 2001-006-00, Automatic Scram During Surveillance Test and Subsequent Reactor Water Level Anomalies.

4. Entergy letter dated February 25, 2002, LER 2001-007-00, Automatic Scram During Transient Caused by Failure of Calibrating Unit.

LETTER NUMBER: 2.03.004

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Entergy Nuclear Operations Inc. (Entergy) proposes to amend the Pilgrim Station Facility Operating License, DPR-35. This proposed license amendment will modify the requirements for the Core Spray and Low Pressure Coolant Injection Systems during shutdown conditions. Consistent with Standard Technical Specifications (Reference 1) both more restrictive and less restrictive requirements for the systems' Operability are proposed.

Pilgrim has reviewed the proposed amendment in accordance with 10 CFR 50.92 and concludes it does not involve a significant hazards consideration.

203004

Entergy Nuclear Operations, Inc. Letter Number: 2.03.004 Pilgrim Nuclear Power Station Page 2 References 2, 3, and 4 described reactor water level indication anomalies that have occurred at Pilgrim Station. Entergy has been aggressively pursuing a long-term resolution to the identified issues. Entergy has developed a design change to be installed during the upcoming refueling outage to resolve the issues. While developing the installation details for the modification it has been determined that a Technical Specification change is needed to allow the modification to be installed without an unnecessary extension in the refueling outage duration.

Entergy requests approval of the proposed amendment by April 19, 2003 to support Pilgrim's upcoming refueling outage (scheduled to commence on April 19, 2003). Once approved, the amendment will be implemented within 60 days.

If you have any questions or require additional information, please contact Bryan Ford at (508) 830-8403.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 23rd day of January 2003.

Sincerely, Robert M. Bellamy

Enclosure:

Evaluation of the Proposed Changes - 10 pages Attachments: 1. Proposed Technical Specification and Bases Changes (mark-up) - 15 pages

2. List of Regulatory Commitments- 1 page 203004

Entergy Nuclear Operations, Inc. Letter Number: 2.03.004 Pilgrim Nuclear Power Station Page 3 cc: Mr. Travis Tate, Project Manager Office of Nuclear Reactor Regulation Mail Stop: 0-8B-1 U.S. Nuclear Regulatory Commission 1 White Flint North 11555 Rockville Pike Rockville, MD 20852 U.S. Nuclear Regulatory Commission Region 1 475 Allendale Road King of Prussia, PA 19406 Senior Resident Inspector Pilgrim Nuclear Power Station Mr. Steve McGrail, Director Mass. Emergency Management Agency 400 Worcester Road P.O. Box 1496 Framingham, MA 01702 Mr. Robert Walker Radiation Control Program Commonwealth of Massachusetts Exec Offices of Health & Human Services 174 Portland Street Boston, MA 02114 203004

Letter 2.03.004 Enclosure Page 1 of 10 ENCLOSURE Evaluation Of The Proposed Changes

Subject:

Emergency Core Cooling System Requirements During Shutdown Conditions

1. DESCRIPTION

2. PROPOSED CHANGES

3. BACKGROUND

4. TECHNICAL ANALYSIS

5. REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration

6. ENVIRONMENTAL CONSIDERATION

7. COORDINATION WITH OTHER PENDING TS CHANGES

8. REFERENCES

Letter 2.03.004 Enclosure Page 2 of 10 Emergency Core Cooling System Requirements During Shutdown Conditions

1. Description This letter is a request to amend Operating License DPR-35 for Pilgrim Nuclear Power Station. This proposed license amendment will modify the requirements for the Core Spray and Low Pressure Coolant Injection (LPCI) Systems during shutdown conditions.

Consistent with Standard Technical Specifications (Reference 1) both more restrictive and less restrictive requirements for the systems' Operability are proposed. Other administrative changes are also made for consistency with the proposed shutdown requirements.

References 2, 3, and 4 described reactor water level indication anomalies that have occurred at Pilgrim Station. Entergy has been aggressively pursuing a long-term resolution to the identified issues. Entergy has developed a design change to be installed during the upcoming refueling outage to resolve the issues. While developing the installation details for the modification it has been determined that a Technical Specification change is needed to allow the modification to be installed without an unnecessary extension in the refueling outage duration due to instrumentation operability requirements during refueling operations.

Entergy has reviewed the proposed amendment in accordance with 10 CFR 50.92 and concludes it does not involve a significant hazards consideration. Entergy requests approval of the proposed amendment by April 19, 2003 to support Pilgrim's upcoming refueling outage (scheduled to commence on April 19, 2003).

2. Proposed Change The following changes are proposed:

A. High Drywell Pressure Instrumentation Add a Note (7) to Table 3.2.B identifying that the High Drywell Pressure Initiation Instrumentation is only required to be Operable during the Run, Startup and Hot Shutdown Modes.

B. Table 3.2.B Note (1)

Reword the Note to clarify that the supported feature(s) must be restored to operable within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or immediately be declared inoperable so that the appropriate actions will be taken.

C. Core Spray System Requirements in Run, Startup, and Hot Shutdown Modes Reformat and reword Specification 3.5.A.1 and Specification 3.5.A.2 to only apply to the Core Spray Systems during the Run, Startup and Hot Shutdown Modes and incorporate the requirements of Specification 3.5.A.5.

Letter 2.03.004 Enclosure Page 3 of 10 D. LPCI System Requirements in Run, Startup, and Hot Shutdown Modes Reformat and reword Specification 3.5.A.3 and Specification 3.5.A.4 to only apply to the LPCI System during the Run, Startup and Hot Shutdown Modes and incorporate the requirements of Specification 3.5.A.5.

E. Existing Specifications 3.5.A.5 and 3.5.F.2 Delete the existing Specifications .3.5..5 and 3.5.F.2. The applicable requirements are included in Specifications 3.5.A.2, 3.5.A.4, and 3.5.A.6.

F. Low Pressure Injection/Spray Requirements in Cold Shutdown and Refueling Modes when not Flooded Up Delete the current allowance in Specification 3.5.F.3 for no low pressure injection/spray systems to be Operable during Cold Shutdown. Add a requirement as Specification 3.5.A.5 that 2 low pressure injection/spray subsystems shall be Operable during the Cold Shutdown and Refuel Modes unless the reactor head is removed, the spent fuel pool gates are removed, and water level is at greater than or equal to elevation 114 foot. Also, add the applicable required Actions to take for low pressure injection/spray subsystem inoperability as Specification 3.5.A.6.

G. Low Pressure Injection/Spray Requirements in Cold Shutdown and Refueling Modes when Flooded Up Delete the current requirement in Specification 3.5.F.4 for a low pressure injection/spray system to be Operable during the Refuel Mode when the reactor head is removed, the spent fuel pool gates are removed, and water level is at greater than or equal to elevation 114 foot.

H. Specification 3.7.A.1 .n Delete the cross reference to Specification 3.5.F.3 from Specification 3.7.A.1 .n.

Also included in Attachment 1 are the associated Bases changes that will be made as part of implementation of the proposed changes.

3. Background

The Emergency Core Cooling System (ECCS) is designed, in conjunction with the primary containment and secondary containment, to limit the release of radioactive materials to the environment following a loss of coolant accident (LOCA). The ECCS uses two independent methods (flooding and spraying) to cool the core during a LOCA.

The ECCS network consists of the High Pressure Coolant Injection (HPCI) System, the Core Spray (CS) System, the Low Pressure Coolant Injection (LPCI) System, and the Automatic Depressurization System (ADS). The suppression pool provides the required source of water for the ECCS. Although no credit is taken in the safety analyses for the condensate storage tank (CST), it is capable of providing a source of water for the HPCI and CS systems.

Letter 2.03.004 Enclosure Page 4 of 10 During power operation on receipt of an initiation signal, ECCS pumps automatically start; simultaneously, the system aligns and the pumps inject water, taken either from the CST or suppression pool, into the Reactor Coolant System (RCS) as RCS pressure is overcome by the discharge pressure of the ECCS pumps. Although the system is initiated, ADS action is delayed, allowing the operator to interrupt the timed sequence if the system is not needed. The HPCI pump discharge pressure almost immediately exceeds that of the RCS, and the pump injects coolant into the vessel to cool the core.

If the break is smalE, the HPCI System will maintain coolant inventory as well as vessel level while the RCS is still pressurized. If HPCI fails, it is backed up by ADS in combination with LPCI and CS. In this event, the ADS timed sequence would be allowed to time out and open the selected safety/relief valves (S/RVs) depressurizing the RCS, thus allowing the LPCI and CS to overcome RCS pressure and inject coolant into the vessel. If the break is large, RCS pressure initially drops rapidly and the LPCI and CS cool the core.

Water from the break returns to the suppression pool where it is used again and again.

Water in the suppression pool is circulated through a heat exchanger cooled by the Reactor Building Closed Cooling Water System (RBCCW). Depending on the location and size of the break, portions of the ECCS may be ineffective; however, the overall design is effective in cooling the core regardless of the size or location of the piping break. Although no credit is taken in the safety analysis for the RCIC System, it performs a similar function as HPCI, but has reduced makeup capability. Nevertheless, it will maintain inventory and cool the core while the RCS is still pressurized following a reactor pressure vessel (RPV) isolation.

All ECCS subsystems are designed to ensure that no single active component failure will prevent automatic initiation and successful operation of the minimum required ECCS equipment.

The CS System is composed of two independent subsystems. Each subsystem consists of a motor driven pump, a spray sparger above the core, and piping and valves to transfer water from the suppression pool to the sparger. The CS System is designed to provide cooling to the reactor core when reactor pressure is low. Upon receipt of an initiation signal, the CS pumps in both subsystems are automatically started when AC power is available. When the RPV pressure drops sufficiently, CS System flow to the RPV begins.

LPCI is an independent operating mode of the RHR System. There are four motor driven LPCI pumps with piping and valves to transfer water from the suppression pool to the RPV via a recirculation loop. The LPCI subsystems are designed to provide core cooling at low RPV pressure. Upon receipt of an initiation signal, all four LPCI pumps are automatically started and RHR System valves in the LPCI flow path are automatically positioned to ensure the proper flow path for water from the suppression pool to inject into the recirculation loops. When the RPV pressure drops sufficiently, the LPCI flow to the RPV, via a recirculation loop, begins. The water then enters the reactor through the jet pumps.

The ECCS performance is evaluated for the entire spectrum of break sizes for a postulated loss of coolant accident (LOCA). The long term cooling analysis following a

Letter 2.03.004 Enclosure Page 5 of 10 design basis LOCA demonstrates that only one low pressure ECCS injection/spray subsystem is required, post LOCA, to maintain adequate reactor vessel water level in the event of an inadvertent vessel draindown. Therefore, it is reasonable to assume, based on engineering judgement, that while in Cold Shutdown or Refuel Mode, one low pressure ECCS injection/spray subsystem can maintain adequate reactor vessel water level. During the Cold Shutdown and Refuel Modes when the reactor head is removed, the spent fuel pool gates are removed, and water level is at greater than or equal to elevation 114 foot in excess of 300,000 gallons of water is available to provide core cooling in the event of leakage from the reactor vessel without ECCS operation.

During the upcoming refueling outage Entergy plans to install modifications on the reference and variable legs of the instrument racks that support the ECCS and Feedwater level instrumentation. The purpose of the modifications are to resolve reactor water level indication issues described in References 2, 3, and 4. Installing these modifications will take about 9 days. During this time the associated level instrumentation will be inoperable.

Entergy plans to install the modifications with the unit in the Refuel Mode while the reactor head is removed, the spent fuel pool gates are removed, and water level is at greater than or equal to elevation 114 foot. The current Pilgrim Technical Specifications prevents this modification from being installed while the unit is in the Refuel Mode by requiring a CS or LPCI subsystem and its associated instrumentation to be Operable for refueling activities to continue. Compliance with these requirements would have extensive impact on the refueling outage's duration.

4. Technical Analysis The following is an analysis of the proposed changes:

A. High Drywell Pressure Instrumentation The High Drywell Pressure actuation instrumentation provides a redundant and diverse initiation signal to the Reactor Water Level actuation instrumentation.

The High Drywell Pressure signal is actuated by the increase in the Drywell (a.k.a, Primary Containment) pressure that results from high energy fluid escaping from the reactor coolant system into the Drywell during a LOCA; thereby, pressurizing the Drywell. When the unit is in the Cold Shutdown or in the Refueling Mode the Drywell is not required to be Operable due to the lower energy contained in the reactor coolant system. In these modes, the Drywell may be opened (e.g. equipment hatch removed) and consequentially, a LOCA will not result in significant pressurization. As a result, this instrument cannot provide a redundant signal to the Reactor Water Level actuation instrumentation.

The Reactor Water Level actuation instrumentation will still be required to be Operable when the associated systems are required to be Operable.

The proposed change adds Note 7 to Table 3.2.B identifying that the High Drywell Pressure Initiation Instrumentation is only required to be Operable during the Run, Startup and Hot Shutdown Modes. The proposed change is consistent with the function of the High Drywell Pressure instrument to provide an initiation signal as a result of a LOCA that results in Drywell pressurization.

Letter 2.03.004 Enclosure Page 6 of 10 B. Table 3.2.B Note (1)

Specification 3.2.B requires the instrumentation functions contained in Table 3.2.B to be Operable whenever the supported systems are Operable. If a required instrument function is inoperable Note (1) requires that the unit be placed in Cold Shutdown. An acceptable alternative to the Note (1) action in the current Technical Specifications is to exit the applicability of the Specification (e.g., declare the supported system inoperable).

The systems associated with these instruments are in some cases required to be Operable in the Cold Shutdown or Refuel Mode. When in these Modes actions other than placing the unit in Cold Shutdown (e.g., halting operations with potential for draining the reactor vessel) may be necessary and identified in the supported system Specification if the supported system cannot perform its function because of an inoperable Table 3.2.B instrument. Therefore, the appropriate way to implement this Specification is to exit the applicability of the Specification if a required function is inoperable and take the actions of the associated affected system. The current wording of the Note provides inadequate compensatory requirements for an inoperable instrument function by directing the unit to be placed in Cold Shutdown at the end of the allowed out of service time. In fact, the appropriate action is to exit the applicability of the Specification (e.g., declare the supported system inoperable).

The proposed change rewords the Note to clarify that the supported feature(s) must be declared inoperable so that the appropriate actions will be taken. This wording is consistent with Reference 1.

C. Core Spray System Requirements in Run, Startup, and Hot Shutdown Modes Proposed Specification 3.5.A.1 and Specification 3.5.A.2 provide the requirements for the Core Spray Systems during the Run, Startup and Hot Shutdown Modes. The proposed Specifications contain the same requirements for these Modes as previously contained in the current Specifications 3.5.A.1, 3.5.A.2, 3.5.A.5 and 3.5.F.2.

D. LPCI System Requirements in Run, Startup, and Hot Shutdown Modes Proposed Specification 3.5.A.3 and Specification 3.5.A.4 provide the requirements for the LPCI during the Run, Startup and Hot Shutdown Modes.

The proposed Specifications contain the same requirements for these Modes as previously contained in the current Specifications 3.5.A.3, 3.5.A.4, 3.5.A.5 and 3.5.F.2.

E. Existing Specifications 3.5.A.5 and 3.5.F.2 The existing Specifications 3.5.A.5 and 3.5.F.2 are deleted by the proposed change. The applicable requirements are included in Specifications 3.5.A.2, 3.5.A.4, and 3.5.A.6. Any changes in requirements are discussed in sections 4.F and 4.G.

Letter 2.03.004 Enclosure Page 7 of 10 F. Low Pressure Injection/Spray Requirements in Cold Shutdown and Refueling Modes when not Flooded Up The current allowance in Specification 3.5.F.3 for no low pressure injection/spray systems to be Operable during Cold Shutdown is deleted by the proposed change.

When the unit is in the Cold Shutdown or Refuel Mode and at low reactor vessel water level (i.e., unless the reactor head is removed, the spent fuel pool gates are removed, and water level is at greater than or equal to elevation 114 foot) a drain down of the vessel could occur quickly. In this case, an ECCS system may be needed to ensure adequate coolant inventory and sufficient heat removal capability for the irradiated fuel in the core.

The long term cooling analysis following a design basis LOCA demonstrates that only one low pressure ECCS injection/spray subsystem is required, post LOCA, to maintain adequate reactor vessel water level. It is reasonable to assume, based on engineering judgement, that while in the Cold Shutdown or Refuel Modes, one low pressure ECCS injection/spray subsystem can maintain adequate reactor vessel water level. Appropriate actions are also proposed for system inoperability consistent with Reference 1.

In general, the times during a refueling outage when the unit is at a low reactor vessel water level are the highest risk times of the outage. Therefore, it is prudent, consistent with Reference 1, to require a low pressure injection or spray system to be Operable during this time. To provide redundancy, a minimum of two low pressure injection/spray subsystems are proposed to be required to be Operable in this condition.

Each low pressure injection/spray subsystem required to be Operable will consist of one pump and the associated required piping and valves of the LPCI or Core Spray Systems. Consistent with current requirements and Reference 1, one LPCI subsystem may be aligned for decay heat removal and considered Operable, if it can be manually realigned (remote or local) and is not otherwise inoperable. This will be allowed because the low pressure and low temperature conditions allow sufficient time to manually align and initiate LPCI subsystem operation to provide core cooling prior to postulated fuel uncovery.

G. Low Pressure Injection/Spray Requirements in Cold Shutdown and Refueling Modes when Flooded Up Current Specification 3.5.F.4 requires that a low pressure injection or spray system be Operable during refueling operations. Most activities that could be considered refueling operations (e.g., fuel movement) occur when the reactor vessel is flooded up (i.e., the reactor head is removed, the spent fuel pool gates are removed, and water level is at greater than or equal to elevation 114 foot).

The condition of the reactor head being removed, the spent fuel pool gates removed, and water level at greater than or equal to elevation 114 foot provides

Letter 2.03.004 Enclosure Page 8 of 10 greater than 300,000 gallons of water over the fuel seated in the reactor vessel.

This provides sufficient coolant inventory to allow operator action to terminate the inventory loss 6r establish make up capability prior to fuel uncovery in case of an inadvertent drain down. ECCS capability does not mitigate other types of refueling operation events that are not associated with an inadvertent loss of inventory (e.g. fuel handling accident).

Therefore, a low pressure injection/spray system should not be required to be Operable during the Cold Shutdown or Refuel Modes when the reactor head is removed, the spent fuel pool gates are removed, and water level is at greater than or equal to elevation 114 foot.

H. Specification 3.7.A.1.n Current Specification 3.7.A.1 .n contains a cross reference to both Specifications 3.5.F.3 and 3.5.F.5. Specification 3.5.F.5 provides the necessary requirements for the condition discussed in Specification 3.7.A.1.n. Therefore, the cross reference to Specification 3.5.F.3 is deleted.

5. Regulatory Safety Analysis 5.1 No Significant Hazards Consideration Entergy Nuclear Operations, Inc. (Entergy) is proposing to modify the Technical Specifications for Emergency Core Cooling Systems during shutdown conditions for Pilgrim Nuclear Power Station. The proposed license amendment will modify the requirements for the Core Spray and Low Pressure Coolant Injection (LPCI)

Systems during shutdown conditions. Consistent with Standard Technical Specifications both more restrictive and less restrictive requirements for the systems' Operability are proposed. Other changes are also made for consistency with the proposed shutdown requirements.

Entergy has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed changes that are not administrative in nature either increase the requirements for system operability or remove requirements that are unnecessary.

Additional requirements are proposed for the Cold Shutdown or Refuel Modes when the availability of an Emergency Core Cooling System (ECCS) is most likely to be needed. The additional requirements provided reduce the probability or consequences of potential accidents. The requirements being removed were unnecessary due plant conditions.

Letter 2.03.004 Enclosure Page 9 of 10 Therefore, the proposed change does not inVolve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The propo3ad change does not involve any physical alteration of plant equipment and does not change the method by which any safety-related system performs its function. As such, no new or different types of equipment will be installed, and the basic operation of installed equipment is unchanged. The methods governing plant operation and testing remain consistent with current safety analysis assumptions. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The proposed changes that are not administrative in nature either increase the requirements for system operability or remove requirements that are unnecessary.

Additional requirements are proposed for the Cold Shutdown or Refuel Modes when the availability of an Emergency Core Cooling System (ECCS) is most likely to be needed. The additional requirements provided reduce the probability or consequences of potential accidents. The requirements being removed were unnecessary due plant conditions.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, Entergy concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

6. Environmental Consideration As defined in 10 CFR 20, a review of this TS change determined that the proposed amendment would change a requirement in respect to installation or use of a facility component located within the restricted area, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with the proposed amendment.

Letter 2.03.004 Enclosure Page 10 of 10

7. Coordination With Other Pending TS Changes The mark-ups of TS submitted by Reference 5 added a Note (6) on TS page 3/4.2-17; therefore, the Note added by the proposed TS changed is identified as Note (7).

8. References

1. NUREG 1433, Standard Technical Specifications for General Electric Plants, BWR/4.

2. Entergy letter dated June 20, 2001, LER 2001-003-00, ESF Actuations Due to Invalid Water Level Indications.

3. Entergy letter dated October 12, 2001, LER 2001-006-00, Automatic Scram During Surveillance Test and Subsequent Reactor Water Level Anomalies.

4. Entergy letter dated February 25, 2002, LER 2001-007-00, Automatic Scram During Transient Caused by Failure of Calibrating Unit.

5. Entergy letter dated October 10, 2002, Request for Technical Specification Change Concerning Change of Trip Level Setting, Calibration Frequencies, and Editorial Changes, Revision 1.

ATTACHMENT 1 PROPOSED TECHNICAL SPECIFICATIONS AND BASES CHANGES (MARK-UP)

Marked TS pages: 3/4.2-11 3/4.2-17, with INSERTS 1 and 2 for page 3/4.2-17 3/4.5-1 3/4.5-2, with INSERT for pages 3/4.5-1 and 2 3/4.5-10 3/4.7-3 Marked TS Bases pages: B3/4.5-1, with INSERT 1 B3/4.5-2, with INSERTS 1 and 2 B3/4.5-22

PNPS TABLE 3.2.B (Cont)

INSTRUMENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINMENT COOLING SYSTEMS Minimum # of Operable Instrument Channels Per Trip System (1) Trip Function Trip Level Setting Remarks High Drywell Pressure <2.22 psig 1. Initiates Core Spray; LPCI; HPCI.

2. In conjunction with Low-Low Reactor Water Level, 94.4-115.6 second time delay and LPCI or Core Spray pump running, initiates Auto Blowdown (ADS)

3. Initiates starting of Diesel Generators

4. In conjunction with Reactor Low Pressure initiates closure of HPCI vacuum breaker containment isolation valves.

1 Reactor Low Pressure 400 psig +/- 5 Permissive for opening Core Spray and LPCI Admission valves.

1 Reactor Low Pressure <76 psig In conjunction with PCIS signal permits closure of RHR (LPCI) injection valves.

1 Reactor Low Pressure 400 psig +/- 5 In conjunction with Low-Low Reactor Water Level initiates Core Spray and LPCI.

2 Reactor Low Pressure 900 psig +/- 5 Prevents actuation of LPCI break detection circuit.

2 Reactor Low Pressure 80 psig +/- 5 Isolates HPCI and in conjunction with High Drywell Pressure initiates closure of HPCI vacuum breaker containment isolation valves.

Revision 180 42 113

-1 4 8 7- 1 5 1 , 162 3/4.2-11 Amendment No. 7-

NOTES FOR TABLE 3.2.B

1. Whenever any CSCS subsystem is required by Section 3.5 to be operable, there shall be two (Note 5) operable trip systems. If the first column cannot be met for one of the trip systems, that system shall be repaired or the reactor sa be aced in heoi ds td nrn hours ate-F-r-r s stem is ade or found to be inoperable -

2. Close isolation valves in RCIC subsystem.

3. Close isolation valves in HPCI subsystem.

4. Instrument set point corresponds to 79.96 inches above top of active fuel.

5. RCIC has only one trip system for these sensors.

Revision 177 Amendment No. 198;-148;-151 3/4.2-17

INSERT 1 FOR PAGE 3/4.2-17 restored to Operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Otherwise, immediately declare the associated supported feature(s) inoperable.

INSERT 2 FOR PAGE 3/4.2-i7

7. Only required to be Operable in Run, Startup, and Hot Shutdown Modes.

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5 CORE AND CONTAINMENT COOLING 4.5 CORE AND CONTAINMENT COOLING SYSTEMS SYSTEMS Applicability Applicability Applies to the operational status Applies to the Surveillance of the core and suppression pool Requirements of the core and cooling systems. suppression pool cooling systems which are required when the corresponding Limiting Condition for operation is in effect.

Obiective Obiective To assure the operability of the To verify the operability of the core and suppression pool cooling core and suppression pool cooling systems under all conditions for systems under all conditions for which this cooling capability is which this cooling capability is an essential response to station an essential response to station abnormalities. abnormalities.

Specification Specification.

A. Core Spray and LPCI Systems A. Core Spray and LPCI Systems "1. Both core spray systems shall 1. Core Spray System Testing.

be operable whenever irradiated' fuel is in the vessel and prior Item Frequency to reactor startup from a Cold Condition, except as specified in a. Simulated Once/

".5.A.2below. Automatic Operating Actuation Test. Cycle

2. Froý',ýd after the date that one of the cores!ray systems is made or b. Pump When tested found to be-i4 operable for any Operability. as specified in reason, continueRd reactor 3.13 verify that operation is perrristible during the each core succeeding seven days, provided spray pump that during such seven;"ays all delivers at active components of the ot er least core spray system and active 3300 GPM components of the LPCI system against a and the diesel generators are system head o erable corresponding to a reactor vessel pressure of 104 psig.

Amendment No. 176 3/4.5-1 I

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5 CORE AND CONTAINMENT COOLING 4.5 CORE AND CONTAINMENT COOLING SYSTEMS SYSTEMS' S... A. Core Spray and LPCI Systems (Cont) A. Core Spray and LPCI Systems (Cont)

1. c. Motor As Specified Operated in 3.13 Valve Operability

d. Core Spray Header Ap Instrumentation Check Once/day Calibrate Once/3 months Test Step Once/3 months

2. This section intentionally left blank 3.Te LPCI system shall be opeal 3. LPCI system testing shall be as I \ whenever irradiated fuel is in the follows:

/ \ reactor vessel, and prior to reactor startup from a Cold Condition, a. Simulated Once/

except as specified in 3.5A.4 and -Automatic Operating Actuation Test. Cycle

4. F and after the date that the b. Pump When tested LPC system is made or found to be Operability. as specified inope ble for any reason, continued in 3.13, verify reactor peration is permissible only that each during th succeeding seven days LPCI pump unless it is ooner made operable, delivers provided that\during such seven 4800 GPM at days the activecomponents of both a head across core spray syst# s and the diesel the pump of generators requir d for operation of at least 380 such components, ý no external ft.

source of power wer vailable, shall be operable. c. Motor As Specified Operated in 3.13

5. Ifthe requirements of 3.5. nnot Valve be met, an orderly shutdown the Operability reactor shall be initiated and th reactor shall be in the Cold Shutdown Condition within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Amendment No. 176 314.5-2

INSERT FOR PAGE 3/4.5-1 and 2

1. Both core spray systems shall be Operable during Run, Startup, and Hot Shutdown Modes and prior to reactor startup from Cold Shutdown, except as specified in 3.5.A.2.

2. During Run, Startup, and Hot Shutdown Modes:

a. With one of the core spray systems inoperable, restore the inoperable core spray system to Operable status within 7 days and maintain all active components of the LPCI system and the diesel generators Operable. Otherwise, be in at least Cold Shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. With both of the core spray systems inoperable be in at least Cold Shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3. The LPCI system shall be Operable during Run, Startup, and Hot Shutdown Modes and prior to reactor startup from Cold Shutdown, except as specified in 3.5.A.4.

4. During Run, Startup, and Hot Shutdown Modes with the LPCI system inoperable, restore the LPCI system to Operable status within 7 days and maintain both core spray systems and the diesel generators Operable. Otherwise, be in at least Cold Shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

5. Two low pressure injection/spray subsystems shall be Operable during Cold Shutdown and Refuel Modes unless the reactor head is removed, the spent fuel pool gates are removed, and water level is at greater than or equal to elevation 114 foot, except as specified in 3.5.A.6.

6. During Cold Shutdown and Refuel Modes unless the reactor head is removed, the spent fuel pool gates are

removed, and water level is at greater than or equal to elevation 114 foot:

a. With one of th'e required low pressure injection/spray subsystems inoperable, restore the inoperable required low pressure injection/spray subsystem to Operable status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Otherwise, take immediate action to suspend activities with potential for draining the reactor vessel.

b. With both of the required low pressure injection/spray subsystems inoperable, take immediate action to suspend activities with potential for draining the reactor vessel and restore 1 low pressure injection/spray subsystem to Operable status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Otherwise, take immediate action to restore secondary containment and one standby gas treatment system to Operable status and to restore isolation capability in each required secondary containment penetration flow path not isolated.

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.5 CORE AND CONTAINMENT 4.5 CORE AND CONTAINMENT COOLING SYSTEMS (Cont) COOLING SYSTEMS (Cont)

F. Minimum Low Pressure Cooling and F. Minimum Low Pressure Cooling and Diesel Generator Availability Diesel Generator Availability

1. During any period when one 1. When It Is determined that one emergency diesel generator EDG is inoperable, within 24 I (EDG) is Inoperable, continued hours, determine that the reactor operation is permissible operable EDG Is not inoperable I only during the succeeding 72 due to a common cause failure, hours unless such EDG is sooner made operable, provided OR that all of the low pressure core perform surveillance 4.9.A.1.a and containment cooling systems shall be operable, and for the operable EDG, the remaining EDG shall be AND operable in accordance with 4.5.F.1. If this requirement within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and once every 8 cannot b.9 met, an orderly hours thereafter, verify correct shutdown shall be Initiated and breaker alignment and Indicated the reactor shall be placed in power availability for each the Cold Shutdown Condition offslte circuit.

within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

2. Confirm the Station Black Out The 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> LCO can be Diesel Generator (SBO-DG) has extended to 14 days provided, In been demonstrated operable addition to the above within the preceding 7 days requirements, the Station Black Out Diesel Generator Is verified OR operable In accordance with 4".5Q.F.2 within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of declaring an EDG Inoperable, perform a

2. combination of Inoperable surveillance to demonstrate that co ents In the core and tha SBO-DG is operable, containm cooling systems shall not defea t-tlcapability of the remaining opera m. AND components to fulfill the coIIag within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of demonstrating the SBO-DG operability as specified above and once every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter, verify normal breaker configuration.

Amendment No. 179 314.5-10

LIMITING CONDITIONS FOR OPERATION SURVEIIJACE REQUIREMENTS 3.7 CONTAINMENT SYSTEMS (Cont) 4.7 CONTAINMENT SYSTEMS (Cont)

A. Primary Containment (Cont)

k. The differential pressure may be reduced to less than 1.17 psid for a maximum of four (4) hours for maintenance activities on the differential pressure control system and during required operability testing of the HPCI system, the relief valves, the RCIC system and the drywell-suppression chamber vacuum breakers.

1. If the specifications of Item i, above, cannot be met, and the differential pressure cannot be restored within the subsequent (6) hour period, an orderly shutdown shall be initiated and the reactor shall be in a cold shutdown condition in twenty-four (24) hours.

m. Suppression chamber water level shall be maintained I between -6 to -1 inches on I torus level instrument which corresponds to a downcomer submergence of 3 feet to 3 feet 5 inches.

n. The suppression chamber.can be drained if the conditions as specified in Sectiotl

-3.5.F.5 of this Technical Specification are adhered to.

Revision 183 Amendment No. 177-1l3, 163 3/4.7-3

B 3/4.5.A Core Spray and LPCI System BASES:

BACKGROUND Each Core Spray system consists of one pump and associated piping and valves with all active components required to be operable.

The LPCI system consists of four LPCI pumps and associated piping

ý1 .ý and valves with all active components required to be operable.

The LPCI system is not considered inoperable when the RHR System is operating in the shutdown cooling mode.

APPLICABLE Based on the loss of coolant analysis performed by General Electric in SAFETY ANALYSES accordance with Section 50.46 and Appendix K of 10CFR50, the Pilgrim I Emergency Core Cooling Systems are adequate to provide sufficient cooling to the core to dissipate the energy associated with the entire spectrum of break sizes for a postulated loss of coolant accident, to limit calculated fuel clad temperature to less than 2200°F, to limit calculated local metal water reaction to less than or equal to 17%,

and to limit calculated core wide metal water reaction to less than or equal to 1% to maintain the core in a coolable geometry, and to provide adecquate long term cooling. The detailed bases is described in Section 6.5 of the PNPS FSAR.

The analyses described in Section 6.5 of the PNPS FSAR calculated a peak clad fuel temperature of less than 2200'F with a Core Spray pump I

flow of 3200 gallons per minute (gpm). A flow rate of 3300 gpm ensures adequate flow for events involving degraded voltage.

Core spray distribution has been shown, in full-scale tests of systems similar in design to that of Pilgrim, to exceed the minimum requirements by at least 25%. In addition, cooling effectiveness has been demonstrated at less than half the rated flow in simulated fuel assemblies with heater rods to duplicate the decay heat characteristics of irradiated fuel. The accident analysis takes credit for core spray flow into the core at vessel pressure below 270 psig. However, the analysis is conservative in that no credit is taken for spray cooling heat transfer in the hottest fuel bundle until the pressure at rated flow for the core spray (104 psig vessel pressure) is reached.

The LPCI system is designed to provide emergency cooling to the core by flooding in the event of a loss-of-coolant accident. This system functions in combination with the core spray system to prevent excessive fuel clad temperature. The LPCI system and the core spray system provide adequate cooling for break areas of approximately 0.2 square feet up to and including the double-ended recirculation line break without assistance from the high pressure emergency core cooling systems. The analyses in described in Section 6.5 of the PNPS FSAR calculated a peak clad fuel temperature of less than 2200°F with LPCI pump flows of 4550 gpm, 4033 gpm, and 3450 gpm for two, three, and four pump combinations feeding into a single loop. A single pump flow rate at 4800 gpm ensures sufficient flow to meet or exceed the analyses' assumptions.

(continued)

Revision 219 B13/4.5-1 Amendment No. 4-7-6

INSERT 1 PAGE B3/4.5-1 Each low pressure injection/spray subsystem during the Cold Shutdown or Refuel Mode consists of any one pump and the required associated piping and valves with all active components required to be operable of the LPCI or Core Spray Systems.

Core Spray and LPCI System B 314.5.A BASES APPLICABLE The analyses of LOCA for PNPS demonstrated the combination SAFETY ANALYSES of LPCS/LPCI systems are sufficient to provide core cooling (continued) even with a single failure of either an active or passive safety-related component.

The analyses determined there were four significant single failures that challenge the Emergency Core Coolant Systems' capability to prevent fuel damage during the postulated LOCA. They are:

1) Battery Failure - Loss of a single battery train could leave only one LPCS pump, two LPCI pumps, and ADS to mitigate the LOCA. This is the most limiting single failure for all but the largest postulated recirculation line breaks and for all postulated non-recirculation line breaks.

2) LPCI Injection Valve Failure - Loss of the injection valve selected by LPCI Loop Selection Logic for the pathway for all LPCI pumps' flow leaves two core spray pumps, HPCI, and ADS for LOCA mitigation. This becomes the limiting single failure for the largest postulated recirculation line breaks.

3) Loss of one emergency diesel generator - This leaves one LPCS pump, two LPCI pumps, and ADS for LOCA mitigation.

4) HPCI Failure - This leaves all other ECCS resources available. It is a significant failure primarily for small line breaks.

In all cases above, the remaining ECCS resources are sufficient to

- l prevent PCT from exceeding 22001F and other criteria provided in I "Ic A- 1j *.Section 50.46 and Appendix K of 10CFRS0.

-ouldone Core Spray system become inoperable. the remaining Ore Spray and the LPCI system are available should the need for core ssrd-crn~udgrae otfltrhlgreliability of the oe Eremaining systems (i.e., the Core Spray and*EP-C-a-sen.

n-dý paid eriod was obtained.

SURVEILLANCE The testing interval for the core and containment REQUIREMENTS cooling systems is based on industry practice, quantitative reliability analysis, judgment and practicality. The core cooling systems have not been desioned to be fully testable during operation.. To increase the availability of the core and containment cooling systems, the components which make up the system; i.e., instrumentation, pumps, valves, etc., are tested frequently. The pumps and motor operated valves are tested in accordance with ASME B&PV Code,Section XI (IWP and IWV, except where specific relief is granted) to assure their operability.

The frequency and methods of testing are described in the PNPS 1ST (continued)

Amendment No. 176 83/4.5-2

INSERT 1 PAGE B3/4.5-2 During unit operation the ECCS performance is evaluated for the entire spectrum of break sizes for a postulated loss of coolant accident (LOCA). The long term cooling analysis following a design basis LOCA demonstrates that only on- l6w pressure ECCS injection/spray subsystem is required, post LOCA, to maintain adequate reactor vessel water level in the event of an inadvertent vessel draindown.

Therefore, it is reasonable to assume, based on engineering judgement, that while in Cold Shutdown or Refuel Mode, one low pressure ECCS injection/spray subsystem can maintain adequate reactor vessel water level. During the Cold Shutdown and Refuel Modes when the reactor head is removed, the spent fuel pool gates are removed, and water level is at greater than or equal to elevation 114 foot in excess of 300,000 gallons of water is available to provide core cooling in the event of leakage from the reactor vessel without ECCS operation. These conditions provide sufficient coolant inventory to allow operator action to terminate the inventory loss prior to fuel uncovery in case of an inadvertent draindown; therefore, no ECCS is required to be operable.

INSERT 2 PAGE B3/4.5.2 3.5.A.2 During the Run, Startup, and Hot Shutdown Modes should one Core Spray system become inoperable, the remaining Core Spray and the LPCI system are available should the need for core cooling arise.

Based on judgment of the reliability of the remaining systems a seven-day repair period is justified. If the inoperable Core Spray system is not restored to Operable status within the 7 days or one of the remaining systems becomes inoperable the unit must be in at least Cold Shutdown within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3.5.A.4 During the Run, Startup, and Hot Shutdown Modes should the LPCI system become inoperable, the remaining Core Spray systems are available should the need for core cooling arise. Based on judgment of the reliability of the remaining systems a seven-day repair period is justified. If the inoperable Core Spray system is not restored to Operable status within the 7 days or one of the remaining systems becomes inoperable the unit must be in at least Cold Shutdown within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3.5.A.6.a During Cold Shutdown and Refuel Modes unless the reactor head is removed, the spent fuel pool gates are removed, and water level is at greater than of equal to elevation 114 foot if any one required low pressure injection/spray subsystem is inoperable, the inoperable subsystem must be restored to Operable status in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. In this Condition, the remaining Operable subsystem can provide sufficient vessel flooding capability to recover from an inadvertent vessel draindown. However, overall system reliability is reduced because a single failure in the remaining Operable subsystem concurrent with a vessel draindown could result in the ECCS not being able to perform its intended function. The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> completion time for restoring the required low pressure injection/spray subsystem to Operable status is based on engineering judgment that considered the remaining available subsystem and the low probability of a vessel draindown event.

With the inoperable subsystem not restored to Operable status in the required completion time, action must be immediately initiated to suspend operations with a potential for draining the reactor vessel (OPDRVs) to minimize the probability of a vessel draindown and the subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.

3.5.A.6.b During Cold Shutdown and Refuel Modes unless the reactor head is removed, the spent fuel pool gates are removed, and water level is at greater than of equal to elevation 114 foot with both of the required ECCS injection/spray subsystems inoperable, all coolant inventory makeup capability may be unavailable. Therefore, actions must immediately be initiated to suspend OPDRVs to minimize the probability of a vessel draindown and the subsequent potential for fission product release. Actions must continue until OPDRVs are suspended. One ECCS injection/spray subsystem must also be restored to Operable status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

If at least one low pressure ECCS injection/spray subsystem is not restored to Operable status within the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Completion Time, additional actions are required to minimize any potential fission product release to the environment. This includes ensuring secondary containment is Operable; one standby gas treatment subsystem is Operable; and secondary containment isolation capability (i.e., one isolation valve and associated instrumentation are Operable or other acceptable administrative controls to assure isolation capability) in each associated penetration flow path not isolated that is assumed to be isolated to mitigate radioactivity releases. Operability may be verified by an administrative check, or by examining logs or other information, to determine whether the components are out of service for maintenance or other reasons. It is not necessary to perform the Surveillances needed to demonstrate the Operability of the components. If, however, any required component is inoperable, then it must be restored to Operability status. In this case, the Surveillance may need to be performed to restore the component to Operable status. Actions must continue until all required components are Operable.

The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> completion time to restore at least one low pressure injection/spray subsystem to Operable status ensures that prompt action will be taken to provide the required cooling capacity or to initiate actions to place the plant in a condition that minimizes any potential fission product release to the environment.

Minimum Low Pressure Cooling and Diesel Generator Availability B 3/4.5CORE AND CONTAINMENT COOLING SYSTEMS 3/4.5.F 1314.5.F. Minimum Low Pressure Cooling and Diesel Generator Availability

) BASES I

BACKGROUND The purpose of Specification 3/4.5.F is to assure that adequate core cooling equipment is available at all times. If, for example, one core spray were out of service and the diesel which powered the opposite core spray were out of service, only 2 LCPI pumps would be available.

t s unng refueling outages--V=V'a5or maintenance is performed

- -dp*nsuch time that all low pressure core cooling systems may beanout of ser-if is-specffica1~onq..ovdes that should this occur, no work will be performed on the pnmarsys s- vwhoih.codId lead to draining the vessel. This work would include work on certain conmTro-rod-driv components and recirculation system.

Specification 3.4.F.5 allows removal of one CRD mechanism while the torus is In a drained condition-without compromising core cooling

- capability. The available core cooling capability for a potential draining of the reactor vessel while this work is performed is based on an estimated drain rate of 300 gpm If the control rod blade seal is unseated. Flooding the refuel cavity and dryer/separator pool to elevation 114'-02 corresponds to approximately 305,000 gallons of water and will provide core cooling capability In the event leakage from the control rod drive does occur. A potential draining of the reactor vessel (via control rod blade leakage) would allow this water to enter I

Into the torus and after approximately 243,000 gallons have accumulated (needed to meet minimum NPSH requirements for the LPCI and/or core spray pumps), the torus would be able to serve as a common suction header. This would allow a closed loop operation of the LPCI system and the core spray system (once re-aligned) to the torus. In addition, the other core spray system is lined up to the condensate storage tanks which can supplement the refuel cavity and dryer/separator pool water to provide core flooding, If required.

ACTION The maximum allowed out-of-service (OOS) time-for one EDG Is 14 days, provided that one EDG and the SBO-DG are operable, In addition to all of the low pressure core and containment cooling systems as specified In 3.5.F.1. If the SBO-DG Is determined to be Inoperable, the maximum allowed OCS time for one EDG is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. A 24-hour LCO will control the plant for cold shutdown If the SBC-DG becomes inoperable anytime after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during a 14-day EDG LCO.

SURVEILLANCE T'he SBO-DG shall be determined to be operable as defined below for extending the 3 days QOS time to 14 days for an EDG. The SBQ-DG is operable if a surveillance was completed within the last seven days before extending to a 14-day 0OS; otherwise, a surveillance must be completed to demonstrate that the SBO-DG is operable. The 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> period allows the operators to complete the required SBO-DG surveillance using the 23Kv offsite power source and to notify Commonwealth Electric of the needed use of the 23Kv line In the testing configuration. The SBO-DG is operable if it is capable of PNPS 83/4 .S-2.2 Amendment No. 179

ATTACHMENT 2 LIST OF REGULATORY COMMITMENTS

List of Regulatory Commitments The following table identifies tho6e actions committed to by Pilgrim in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.

REGULATORY COMMITMENT DUE DATE Implementation of amendment within 60 60 days after the approval.

days of approval