LR-N17-0121, License Amendment Request to Relocate the Reactor Coolant System Pressure Isolation Valve Table from the Technical Specifications to the Technical Requirements Manual

License Amendment Request to Relocate the Reactor Coolant System Pressure Isolation Valve Table from the Technical Specifications to the Technical Requirements Manual

ML17270A076
Person / Time
Site:	Salem
Issue date:	09/27/2017
From:	Mcfeaters C Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LAR S17-04, LR-N17-0121
Download: ML17270A076 (22)

Text

PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New ,Jersey 08038-0236 PSEG NuclearLLC

" SEP 2 '1 2017 10 CFR 50.90 LR-N17-0121 LAR S17-04 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Salem Generating Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-70 and DPR-75 NRC Docket Nos. 50-272 and 50-311

Subject:

License Amendment Request to Relocate the Reactor Coolant System Pressure Isolation Valve Table from the Technical Specifications to the Technical Requirements Manual In accordance with the provisions of 10 CFR 50.90, PSEG Nuclear LLC (PSEG) is submitting a request for an amendment to the Technical Specifications {TS) for Salem Generating Station (Salem) Units 1 and 2.

The proposed amendment revises Unit 1 TS 3/4.4.6.3, Primary Coolant System Pressure Isolation Valves, and Unit 2 TS 3/4.4. 7.2, Operational Leakage. Specifically, this change relocates the reactor coolant system (RCS) pressure isolation valve (PIV) lists, Unit 1 TS Table 4.4-3 and Unit 2 TS Table 3.4-1, from the TS to the Salem Technical Requirements Manual

{TRM). In addition, the references to the TS Tables are being removed from the TS Limiting Condition for Operation (LCO) 3.4.6.3 (Unit 1) and 3.4. 7.2.f (Unit 2), TS Action 3.4.6.3.a (Unit 1) and Surveillance Requirements (SR) 4.4.6.3 (Unit 1) and 4.4.7.2.2 (Unit 2). The Unit 1 PIV leakage acceptance criteria contained in Unit 1 TS Table 4.4-3 is relocated to Unit 1 SR 4.4.6.3.

Attachment 1 provides an evaluation supporting the proposed changes. Attachment 2 provides the existing TS pages marked up to show the proposed changes. Attachment 3 provides existing TS Bases pages marked up to show the proposed changes and are being provided for information only.

PSEG requests approval of this license amendment request (LAR) in accordance with standard NRC approval process and schedule. Once approved, the amendment will be implemented within 60 days from the date of issuance.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated State of New Jersey Official.

There are no regulatory commitments contained in this letter.

LR-N17-0121 10 CFR 50.90 Page 2 If you have any questions or require additional information, please contact Mr. Brian Thomas at 856-339-2022.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on __ 9_/_Z_l_IJ..7.L..

(Date)

Respectfully, Charles V. McFeaters Site Vice President Salem Generating Station Attachments:

1. Evaluation of Proposed Changes

2. Mark-up of Proposed Technical Specification Pages

3. Mark-up of Proposed Technical Specifications Bases Pages cc: Mr. D. Dorman, Administrator, Region I, NRC Mr. R. Ennis, Project Manager, NRC NRC Senior Resident Inspector, Salem Mr. P. Mulligan, Chief, NJBNE PSEG Corporate Commitment Tracking Coordinator Salem Commitment Tracking Coordinator

LR-N17-0121 LAR 817-04 Attachment 1 Evaluation of Proposed Changes Table of Contents 1 .0 DESCRIPTION . . . . . . . . ........... . .. . . . . . . . ........ . . . . . . . . ......... . . . . . . . ...... . . . . .. . ......... . . . . . . .. . . . . ...... .. . . ....... 1

2.0 PROPOSED CHANGE

. . . . .. . . .. . . . .. . . ...... . . . . . . . . . . . . ...... . . .. . ........ . . . . . . . .......... . .. . . . . . . ........ . .. . . ....... 1

3.0 BACKGROUND

.............................................................................................................. 2

4.0 TECHNICAL ANALYSIS

. . . . . . .... ....... . . . . . . . . . .. . ...... . .. . . . . ....... . . . . . . . . ..... . . . . . . . . . ............ . . . . . .......... 3

5.0 REGULATORY ANALYSIS

..... . .. . . . . . .. . .. . . . . . . . . . . .. . . ...... . . . . . ...... . . . . .. . . . . . . . . . . . . . . .. . . . . .. . . ...... . . . . . . .. 4

5. 1 No Significant Hazards Consideration .. . . . . . . ........ . . . ........ . . . . . . . ............ .. . . .. .. . . . . . . . . . . . . .. .. 4 5.2 Applicable Regulatory Requirements/Criteria ... . . . . ...... . . .. .. . . . . ...... . . . . . .. .. . . . ......... . . ...... 6

6.0 ENVIRONMENTAL CONSIDERATION

. . . . ...... . ... . . . . . . . . .. .. . . . . . . ........ . . . . . . . . . . . ........ . . . . . . . .. .. . . . . . 8

7.0 REFERENCES

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LR-N17-0121 LAR S17-04 Attachment 1

1.0 DESCRIPTION

The proposed amendment revises Unit 1 TS 3/4.4.6.3, Primary Coolant System Pressure Isolation Valves, and Unit 2 TS 3/4.4.7.2, Operational Leakage. Specifically, this change relocates the reactor coolant system (RCS) pressure isolation valve (PIV) lists, Unit 1 TS Table 4.4-3 and Unit 2 TS Table 3.4-1 , from the TS to the Salem Technical Requirements Manual (TRM). In addition, the references to the TS Tables are being removed from the TS Limiting Condition for Operation (LCO) 3.4.6.3 (Unit 1 ) and 3.4.7.2.f (Unit 2), TS Action 3.4.6.3.a (Unit 1 )

and Surveillance Requirements (SR) 4.4.6.3 (Unit 1 ) and 4.4.7.2.2 (Unit 2). The Unit 1 PIV leakage acceptance criteria contained in Unit 1 TS Table 4.4-3 is relocated to Unit 1 SR 4.4.6.3.

The proposed change is consistent with Generic Letter 91 -08, "Removal of Component Lists from Technical Specifications," which provides guidance to remove component lists from the TS. This request meets all conditions outlined in the Generic Letter. The proposed change for removal of the PIV Table is also consistent with the NUREG-1 43 1 , Revision 4, "Standard Technical Specifications - Westinghouse Plants."

2.0 PROPOSED CHANGE

The proposed changes to the Salem Unit 1 and 2 TS are described below and are indicated on the marked up TS pages provided in Attachment 2 of this submittal.

Salem Unit 1 :

• TS LCO 3.4.6.3 will be revised to delete reference to Table 4.4-3.

• TS Action 3.4.6.3.a. will be revised to delete reference to Table 4.4-3

• TS SR 4.4.6.3 will be revised to delete reference to Table 4.4-3 and add the PIV leakage limits contained in Table 4.4-3 and Notes (a) and (b) of Table 4.4-3.

• TS Table 4.4-3 will be deleted. The list of valves will be relocated to the Salem TRM and the PIV leakage limits will be relocated to SR 4.4.6.3.

Salem Unit 2:

• TS LCO 3.4.7.2.f will be revised to delete reference to Table 3.4-1 .

• SR 4.4.7.2.2 will be revised to delete reference to Table 3.4-1

• TS Table 3.4-1 will be deleted and will be relocated in its entirety to the Salem TRM.

Proposed changes to the Salem Unit 1 and 2 TS Bases are provided in Attachment 3 for information only; changes to the TS Bases pages will be incorporated in accordance with Unit 1 TS 6. 1 7 and Unit 2 TS 6.1 6, "Technical Specifications (TS) Bases Control Program."

No changes are being made to the current RCS PIV leakage limits, actions for inoperable PIVs, or surveillance frequencies contained in the Salem Unit 1 and 2 TS.

1

LR*N17-0121 LAR 517-04

3.0 BACKGROUND

The function of RCS PIVs is to separate the high pressure RCS from an attached low pressure system. Although PIV TS provides a limit on allowable PIV leakage rate, its main purpose is to prevent overpressure failure of the low pressure portions of connecting systems. The leakage limit provides indication that the PIVs between the RCS and the connecting systems are degraded or degrading. PIV leakage could lead to overpressure of the low pressure piping or components.

On May 6, 1 99 1 , Generic Letter (GL) 9 1 -08, "Removal of Component Lists from Technical Specifications," was issued to provide guidance to remove component lists from the Technical Specifications. The guidance stipulates that the TS requirements are stated in general terms that describe the types of components to which the requirements apply, and that the removal of component lists does not alter existing TS requirements or those components to which they apply. In addition, the removed lists must be included in a plant procedure that is subject to the change control provisions for plant procedures in the Administrative Controls section of TS.

Generic Letter 91 -08 provides guidance for preparing a request for a license amendment to remove component lists from technical specifications (TS). The nuclear industry and the U.S.

Nuclear Regulatory Commission (NRC) identified this line-item TS improvement during investigations of TS problems.

The removal of component lists from TS permits administrative control of changes to these lists without processing a license amendment. Any change to component lists contained in plant procedures is subject to the requirements specified in the Administrative Controls section of the TS on changes to plant procedures. Therefore, the change control provisions of the TS provide an adequate means to control changes to these component lists, when they have been incorporated into plant procedures, without including them in TS.

An Enclosure to the Generic Letter provided additional guidance for changing individual TS sections. At the time of issuance in 1 99 1 , the Enclosure to GL 91-08 specifically addressed the issue of PIVs stating:

Guidance on removing from the TS the list of reactor coolant system pressure isolation valves is pending the NRC staff's resolution of generic concerns with existing lists for these valves. In the interim, licensees should not submit proposals to remove this list from the TS.

The NRC has since resolved the Generic Safety Issue referenced in the GL Enclosure. On July 1 , 1 993, NUREG-1 463, "Regulatory Analysis for the Resolution of Generic Safety Issue 1 05:

Interfacing System Loss-of-Coolant Accident in Light-Water Reactors" was issued. The NUREG addressed the outstanding Generic Safety Issue (GSI) 1 05 regarding Interfacing Systems Loss of-Coolant Accident (ISLOCA) and PIVs. Additionally, the NRC has since approved NUREG-1 43 1 , "Standard Technical Specifications - Westinghouse Plants," which does not include PIV Tables. In addition, the NRC has approved specific LARs for relocation of PIV Tables from TS (see References 7.4, 7.5, 7.6, 7.7).

2

LR-N17-0121 LAR 517-04

4.0 TECHNICAL ANALYSIS

PSEG proposes to relocate the RCS PIV component list to the Salem TRM. The TRM is a PSEG controlled document that has been developed to contain requirements relocated from the TS. The TRM is described in Salem Updated Final Safety Analysis Report (UFSAR) section 1 3.5.4 and is controlled in a m anner consistent with procedures fully or partially described in the UFSAR. Revisions to the TRM are reviewed pursuant to 1 0 CFR 50.59.

Relocating the PIV component list from the TS will eliminate the burden of processing license amendments when changes are made to the PIV Table and will facilitate the more effective utilization of NRC and PSEG resources.

GL 9 1 -08, relating to the issue of removing component lists from the TS, states in part:

This guidance includes the incorporation of lists into plant procedures that are subject to the change control provisions for plant procedures In the Administrative Controls Section of the TS. The removal of component lists from TS permits administrative control of changes to these lists without processing a license amendment, as is required to update TS component lists. Any change to component lists contained in plant procedures is subject to the requirements specified in the Administrative Controls Section of the TS on changes to plant procedures. Therefore, the change control provisions of the TS provide an adequate means to control changes to these component lists, when they have been incorporated into plant procedures, without including them in TS.

Specific items identified in Enclosure 1 to GL 91 -08 to be addressed with a request to remove component lists from the TS include:

1 . Each TS should include an appropriate description of the scope of the components to which the TS requirements apply. Components that are defined by regulatory requirements or guidance need not be clarified further. However, the Bases section of the TS should reference the applicable requirements or guidance.

2. If the removal of a component list results in the loss of notes that modify or provide an exception to the TS requirements, the specification should be revised to incorporate that modification or exception. The modification or exception should be stated in terms that identify any group of components by function rather than by plant identification number, if practical.

3. Licensees should confirm that the lists of components removed from the TS are located in appropriately controlled plant procedures. The list of components may be included In the next update of the FSAR. The Bases section of individual specifications also may reference the plant procedures or other documents that identify each component list.

With regard to item (1 ) above, PIVs are described in NUREG- 1 431 as any two normally closed valves in series within the reactor coolant pressure boundary which separate the high pressure RCS from an attached low pressure system. The TS requirements for LCO, Actions, and SR relating to PIVs remain applicable. Therefore, removal of the RCS PIV component list does not affect the scope of components to which the TS requirements apply. Per the proposed changes 3

LR-N17-0121 LAR 817-04 Attachment 1 in Attachment 3, the TS Bases now describe PIVs, which is consistent with the NUREG-1431 Bases.

With regard to item (2) above, the S alem Unit 1 RCS PIV leakage acceptance criteria and associated notes are being relocated from Table 4.4-3 to SR 4.4.6.3. For Salem Unit 2 there are no notes, exceptions, or modifications listed directly in Table 3.4-1 .

With regard to item (3), PSEG will relocate the list of PIVs to the TRM, which is an appropriately controlled plant procedure, during the implementation of this LAR .

GL 9 1 -08 provided the guidance for changing individual TS sections. The guidance written in the Generic Letter was written prior to the resolution of GSI 1 05, which discusses Interfacing Systems Loss of Coolant Accidents. The enclosure to GL 91 -08 specifically addresses the issue of PIVs and this GSI stating:

Guidance on removing from the TS the list of reactor coolant system pressure isolation valves is pending the NRC staff's resolution of generic concerns with existing lists for these valves. In the interim, licensees should not submit proposals to remove this list from the TS.

Explicit guidance on removal of lists of PIVs from the TS has not been issued by the NRC.

However, in S eptember 1 992, the NRC issued NUREG-1 431 , Rev 0, "Standard Technical Specifications, Westinghouse Plants," NUREG- 1 43 1 TS Section 3.4.1 4, "RCS Pressure Isolation Valve (PIV) Leakage," does not contain a list of PIVs.

PSEG concludes that the proposed change to relocate the list of PIVs from the TS to the TRM is administrative in that it merely relocates the component list. No changes are being made to the current RCS PIV leakage limits, actions for inoperablePIVs, or surveillance frequencies contained in the Salem Unit 1 and 2 TS.

PSEG determined that the relocation of PIV component list does not eliminate the requirements for the licensee to ensure that the RCS pressure isolation valves are capable of performing their safety function. Although the PIV component list is relocated from the TSs to the TRM, the information being relocated will be controlled and further revisions to the TRM Table will be subject to 1 0 CFR 50.59.

5.0 REGULATORY ANALYSIS

5. 1 No Significant Hazards Consideration PSEG requests an amendment to the Salem Unit 1 and 2 Operating Licenses. The proposed amendment revises Unit 1 TS 3/4.4.6.3, Primary Coolant System Pressure Isolation Valves, and Unit 2 TS 3/4.4. 7.2, Operational Leakage. Specifically, this change relocates the reactor coolant system (RCS) pressure isolation valve (PIV) lists, Unit 1 TS Table 4.4-3 and Unit 2 TS Table 3.4-1 , from the TS to the Salem Technical Requirements Manual (TRM). In addition, the references to the TS Tables are being removed from the TS Limiting Condition for Operation (LCO) 3.4.6.3 (Unit 1 ) and 3.4.7.2.f (Unit 2), TS Action 3.4.6.3.a (Unit 1 ) and Surveillance 4

L R-N17-0121 LAR 817-04 Requirements (SR) 4.4.6.3 (Unit 1 ) and 4.4. 7.2.2 (Unit 2). The Unit 1 PIV leakage acceptance criteria contained in Unit 1 TS Table 4.4-3 is relocated to Unit 1 SR 4.4.6.3.

PSEG has evaluated the proposed changes to the TS using the criteria in 1 0 CFR 50.92, and determined that the proposed changes do not involve a significant hazards consideration. The following information is provided to support a finding of no significant hazards:

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

Response: No The proposed changes to the TS will not alter the way any structure, system, or component (SSC) functions, and will not alter the manner in which the plant is operated. The proposed changes do not alter the design of any SSC. The relocation of the RCS PIV valve lists from the TS to the TRM is an administrative change. Future revisions to the TRM are subject to 1 0 CFR 50.59. Therefore the probability of an accident previously evaluated is not significantly increased.

The proposed changes do not alter the RCS PIV leakage limits contained in the TS nor do they alter the frequency for testing of the RCS PIV. Therefore, the consequences of an accident previously evaluated are not increased.

Therefore, these proposed changes do not represent 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 proposed changes do not involve a modification to the physical configuration of the plant or changes in the methods governing normal plant operation. The proposed changes will not impose any new or different requirement or introduce a new accident initiator, accident precursor, or malfunction mechanism. The proposed changes are administrative in nature.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Do the proposed changes involve a significant reduction in a margin of safety?

Response: No The proposed changes to the RCS PIV TS are administrative in nature. The proposed changes do not alter the RCS PIV leakage limits contained in the TS nor do they alter the frequency for testing of the RCS PIV. The proposed changes will not result in changes to system design or setpoints that are intended to ensure timely identification of plant 5

LR-N17-0121 LAR 517-04 Attachment 1 conditions that could be precursors to accidents or potential degradation of accident mitigation systems.

The proposed amendment will not result in a design basis or safety limit being exceeded or altered. Therefore, since the proposed changes do not impact the response of the plant to a design basis accident, the proposed changes do not involve a significant reduction in a margin of safety.

Based upon the above, PSEG concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 1 0 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

5.2 Applicable Regulatory Requirements/Criteria 1 0 CFR 50, Appendix A, General Design Criteria (GDC)

Salem was designed and constructed in accordance with Atomic Energy Commission (AEC) proposed General Design Criteria published in July 1 967. The applicable AEC proposed criteria, as document in Salem UFSAR Section 3. 1 , were compared to 1 0 CFR 50 Appendix A General Design Criteria (GDC) as discussed below. The applicable GDC criteria are GDC 1 4, 54, and 55.

Criterion 14-Reactor coolant pressure boundary. The reactor coolant pressure boundary shall be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture.

GDC Criterion 1 4 is similar to AEC Criterion 9.

Criterion 54-Piping systems penetrating containment. Piping systems penetrating primary reactor containment shall be provided with leak detection, isolation, and containment capabilities having redundancy, reliability, and performance capabilities which reflect the importance to safety of isolating these piping systems. Such piping systems shall be designed with a capability to test periodically the operability of the isolation valves and associated apparatus and to determine if valve leakage is within acceptable limits.

GDC Criterion 54 is similar to AEC Criterion 51 and 57.

Criterion 55-Reactor coolant pressure boundary penetrating containment. Each line that is part of the reactor coolant pressure boundary and that penetrates primary reactor containment shall be provided with containment isolation valves as follows, unless it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable on some other defined basis:

( 1 ) One locked closed isolation valve inside and one locked closed isolation valve outside containment; or (2) One automatic isolation valve inside and one locked closed isolation valve outside containment; or (3) One locked closed isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment; or 6

LR-N17-0121 LAR 517*04 Attachment 1 (4) One automatic isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment.

Isolation valves outside containment shall be located as close to containment as practical and upon loss of actuating power, automatic isolation valves shall be designed to take the position that provides greater safety.

Other appropriate requirements to minimize the probability or consequences of an accidental rupture of these lines or of lines connected to them shall be provided as necessary to assure adequate safety. Determination of the appropriateness of these requirements, such as higher quality in design, fabrication, and testing, additional provisions for inservice inspection, protection against more severe natural phenomena, and additional isolation valves and containment, shall include consideration of the population density, use characteristics, and physical characteristics of the site environs.

Salem performed a comparison to GDC Criterion 55 and stated in UFSAR Section 3.1 .3 that valve arrangements that do not comply are discussion in UFSAR Section 6.2.4. USFAR section 6.2.4.1 states in part:

• . . the two barriers may consist of: (a) two closed piping systems or vessels, one inside and one outside the containment, (b) two automatic isolation valves, one inside and one outside containment, (c) an automatic isolation valve inside the containment and a closed system outside the containment, (d) an automatic isolation valve outside the containment and a closed system inside the containment, or (e) an automatic isolation valve outside containment and a closed system outside the containment.

• A check valve on an incoming line or a normally closed valve is considered an automatic valve.

1 0 CFR 50.55a (c) Reactor coolant pressure boundary. Systems and components of boiling and pressurized water-cooled nuclear power reactors must meet the requirements of the ASME BPV Code as specified in this paragraph.

(f) lnservice testing requirements. Systems and components of boiling and pressurized water cooled nuclear power reactors must meet the requirements of the ASME BPV Code and ASME Code for Operation and Maintenance of Nuclear Power Plants as specified in this paragraph.

The administrative change to relocate the RCS PIV component list to the TRM was generically approved by the NRC in NUREG-1 431, '1Standard Technical Specifications - Westinghouse Plants," which is consistent with the NRC Final Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, and 1 0 CFR 50.36.

Generic Letter 91-08, "Removal of Component Lists from Technical Specifications, " provides guidance to remove component lists from the Technical Specifications.

Therefore, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, 7

LR-N17-0121 LAR 517-04 (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

6.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 1 0 CFR 20, 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 1 0 CFR 51 .22(c)(9). Therefore, pursuant to 1 0 CFR 5 1 .22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

7.0 REFERENCES

1. 1 0 CFR 50.36, "Technical Specifications"

2. NUREG1431 , Revision 4.0, Standard Technical Specifications Westinghouse Plants, April 201 2

3. Generic Letter 91 -08, " Removal of Component Lists from Technical Specifications"

4. NRC Safety Evaluation Related to Amendment No. 44 , Seabrook Station, dated November 28, 1 995

5. NRC Safety Evaluation Related to Amendment No. 76, River Bend Station, dated March 8, 1 995

6. NRC Safety Evaluation Related to Amendment Nos 1 82 and 1 44, Limerick Generating Station Unit Nos. 1 and 2, dated February 1 7, 2006.

7. NRC Safety Evaluation Related to Amendment No. 206, Nine Mile Point Unit 1 , dated July 26, 201 0.

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LR-N17-0121 LAR 517-04 Attachment 2 Mark-up of Proposed Technical Specification Pages The following Technical Specifications pages for Renewed Facility Operating License DPR-70 are affected by this change request:

Technical Specification 3/4.4.6.3, Primary Coolant System Pressure Isolation Valves 3/4 4-1 6a, 16b, 1 6c The following Technical Specifications pages for Renewed Facility Operating License DPR-75 are affected by this change request:

Technical Specification 3/4.4. 7.2, Operational Leakage 3/4 4-1 7, 1 8, 1 9 1

R EACTOR COOLANT S YS TEM PRIMARY COOLANT SYSTEM PR ES S U RE ISOLATION VALVES

(

LIMITING CONDITION FOR OPERATION 3.4.6.3 R eactor Coolant System Pressure Isolation Valves Jii.Gifi-r;-t.a9* 72tall be OPERABLE.

APPLICABLITY: MODES 1, 2, 3, and 4.

ACTION:

a. With any Reactor Coolant S y st em Pressure Isolation Valve leakage greater than the specified limit i*R-"]:a.sl+e.tisolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least tiNa c losed manual or deactivated a utom atic valves, or be in at least HOT S TANDBY 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 COLD SHUTDOWN within the fol l owing 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.6.3 Each Reactor Coolant System Pressure Isolation Valve e-eei#ed-imc:rb'lall demonstrated OPERABLE pursuant to the INSERVICE TESTING PROGRAM, except that in be lieu of any leakage testing required by the INSERVICE TEST I NG PROGRAM, each valve shall 1

fd.)Cb) be demonstrated OPERABLE by verifying le a ka g e to bew.iittio..L"ti...l*i*fflit.;.:!: r;, 0 p -Por e.e.'ll.t1 vc.lve  :

a. ln accordance with the Surveillance Frequency Control P ro g ram .

b. Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or more and if leakage testing has not been performed In the previous 9 months.

c. Prior to returning the valve to service following maintenance repair or replacement work on the valve.

d. For the Residual Heat Removal and Safety Injection Systems hot and cold leg injection valves and accumulator va l ves lie--ifl4"-Er19*1-el1e testing will be done within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following valve actuation due to automatic or manual action or flow through the v a l v e. For all other systems tes ting will be done once per refueling.

The provisions of specification 4.0.4 are not applicable for entry into MODE 3 or 4.

SALEM - UNIT 1 3/4 4-16a Amendment No. 319

REACTOR COOLANT Maxiumla) lb)

System Valve No. Allowable Leakage Low Pressure Safety Injection Gl?M Loop ll, cold. leg llSJ56 GPM 11SJ43 GPM Loop 12, cold. leg l2SJS6 Gl?M l2SJ43  ::;;

Loop 13, cold. leg l3SJ56 l3SJ43 s Loop 13, hot leg l3SJ156  ::;

l3RH27  :;;;

Loop 14, cold. lag 14SJ56 5. 0 l4SJ43  ::;; 5. 0 Loop 14 , hot leg 14SJ156 5.0 14RH27 s 5. 0 Intermediate Pressure Safaty Injection  ::;; s.o valve Loop 11, cold leg llSJ144  :;; 5.0 Loop ll, hot leg llSJl56 s 5.0 llSJl39 s 5.0 Loop 12, cold leg l2SJl44 s 5. 0 Loop 12, hot leg l2SJlS6 s s.o l2SJ139 s 5.0 Loop 13, cold leg l3SJ144 s 5.0 Loop 13, hot leg l3SJ156 s s.o l3SJl39 s 5.0 Loop 14, cold. leg l4SJ144  ::; 5.0 Loop 14, hot leg 14SJl56 s s.o l4SJ139 s 5.0 Safety Injection Accumulators to cold leg loop 11, cold leg llSJSS loop 12, cold leg 12SJ55 loop 13, cold leg 13SJ55

--- - - - :r:o p 14,co: a la*g; -------------- -----141JJs*s-- ------- - $; 5-:---

o

-- - - - --- -- --------- ------- - r * - --

Safety Injection Boron Injection to cold legs loop 11, cold leg llSJ l7  ::::; Gl?M each valve loop 12, cold leg l2SJl7 Gl?M each valve loop 13, oold leg l3SJl7 Gl?M each valve loop 14, cold leg l4SJl7 GPM each valve lSJ150 GPM each valve RHR Suction loop 11 lRHl s.o GPM each valve loop ll lRH2 s.o GPM each valve

- ---jl--

SALEM - UNIT l 3/4 4-16b A.mendmen t No. 225

• (a) l. Leakage rates less than or equal to 1.0 gpm are acceptable. However, for initial teats, or tests following valve repai r or replacement, leakage rates less than or equal to 5.0 gpm are considered acceptable.

2. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are c.onsidered acceptable if the latest measured rate has not exce eded the rate detened by the previous test by an amount that reduces the margin between measured le akage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.

3, Leakage rates gr e ater than 1.0 gpm but less than or equal to 5.0 gpm are considered unacceptable if the latest measured rate exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by SO% or greater.

4. Leakage rates greater than 5.0 gpm are considered unacceptable.

(b) Minimum differential test pressure shall not be less than 150 ps iot

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'P"lf-SALEM UNIT 1 Amendment No. 225

OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATL 3.4.7.2 Reactor Coolant System leakage shall be limited to:

(

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 GPM UNIDENTIFIED LEAKAGE,

c. 150 gallons per day primary-to-secondary leakage through any one steam generator, and

d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and

e. NOT USED

f. 1 GPM leakage at a Reactor Coolant System pressure of 2230 +/- 20 psig from )-

any Reactor Coolant System Pressure Isolation Valve i#e'd-in-"fa.

APPLICABILITY: MODES 1, 2, 3 and 4 ACTION:

a. With any PRESSURE BOUNDARY LEAKAGE, or primary;.to-secondary leakage not within limit, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY L EAKAGE and leakage from Reactor

(

Coolant System Pressure Isolation Valves, and primary-to-secondary leakage, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within the n ext 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least two closed manual or deactivated automatic valves, or 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 COLD SHU TDOW N within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.4. 7. 2 1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:

a. Monitoring the containment atmosphere particulate radioactivity monitor in accordance with the Surveillance Frequency Control Program.

b. Monitoring the containment sump inventory in accordance with the Surveillance Frequency Control Program.

SALEM

• UNIT 2 3/4 4-17 Amendment No. 282

SURVE;;ILLANCE REQUIREMENTS (Continued) c*. Verifying primary-to-secondary leakage is 5 150 gallons per day through any one steam generator in accordance with the Surveillance Freq uen c y Control Program during stead y state operation, d*. Performance of a Reactor Co olant System water inventory balance** in accordance with the Surveillance Frequency Control Program. The water inventory balance shall be performed with the plant at steady state conditions.

The provisions of specification 4.0.4 are not applicable for entry into Mode 4, and

e. Monitoring the reactor head flange leakoff system in accordance with the Surveillance Frequency Control Program.

4.4.7.2.2 Each Reactor Coolant S y stem Pressure Is olation Valve 5'petif!ed""in"'FIe :3.4 -a ll be d e mon s trated OPERABLE pursuant to the INSERVICE TESTING PROGRAM, except that in lieu of any leakage testing re quired by the IN SERVICE TESTING PROG RAM each valve shall be demonstrated OPERABLE by verifying leak age to be within its limit:

a. In accord ance with the Surveillance Frequency Control Program.

b. Prior to ente ring MODE 2 wh enever the plant has been in C O LD SHUTDOWN for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or more and if leakage t es ting has not been performed in the previous 9 months.

c. Prior to returning the valve t o service following maintenance repai r or replacement work on the valve .

d. for the Re s idual Heat Removal and Safety Injectlon Systems hot and cold leg injection valves and accumulator valves Jrhe testing will be done with in 24 hou rs following valve actuation due to automatic or manual action or flow through the valve. For all oth er systems testing will be done once per refueling.

The p ro vis ions of specification 4.0.4 are not applicable for entry into M ODE 3 or 4.

Not required to be completed u nti l 12 h ours after establishment of steady state o peration .

Not applicable to primary-to-secondary leakage.

SALEM -UNIT 2 3/4 4-18 Amendment No. 300

REACTOR C004A!T SYSTIM

• TULI 3.4-1

(

REACTOR QOLANT SYSTEM tRESSUll tSOLATlOK VAVES VA!IVI NO F.YlfCTIOM 21SJ43 Safety Injection (L.P. troa RHR Puap1 to C o l d Legs) 22SJ43 Safety Injection (L.P. troa RHR Pps to Cold Legs) 23SJ43 Safety Injection (L.P. troa RBR Puap1 to Cold Legs) 24SJ43 Safety Injection (L.P. froa RBR Puap1 to Cold Legs) 21SJ55 Safety Injection {Accuaulator Discharge to Cold Legs) 22SJSS Safety Injection (Aceuaulator D i s cha rge to Cold Ltgsl 23SJSS S afe t y Injection (Aecuaulator Diaebar;t to Cold Leqsl 24SJ55 safety Injection {Aeeuaulator Diteharqe to Cold Legs) 21SJ56 Safety Injection (Aecuaulator Diaehar;t to Cold Les) 22SJ56 Safety Injection (Aecuaulator Discharge to Cold Legs) 23SJSE safety Injection (Aceuaulator Discharge to Cold Legal 24SJS6 Safety Injection (Aeeuaulator Diaehar g e to Cold Legs) 21SJ17 Safety Injection (Boron Xnjtetion to Cold Ltqs) 22SJ17 Safety Injection (Boron I n j ection to cold Ltgt) 23SJ17 Safety Injection (Beron Injection to Cold Legs) 24SJ17 Safety Injection (Boron Injection to Cold Legs) 2SJ150 Safety Injection (Boron lnjtetion to Cold Ltgl) 21SJ139 sahty Injection *(1!.?. troa SI Puaps to Bot Le;sl 22SJ139 23SJ139 Safety Injection ( B. P fro* SI Puapa to Bot Leg a }

Safety Injection (H.P. ftol SI Puapt to Bot Legs)

Safety Injection (B.P. fro* SI Puaps to Bot Legs) 2.SJ139 21SJ156 Safety Injection (B.P. fro* SI Pumpa to Rot Ltgs) (

22SJ1S6 safoty Injteion (B.P. fro* SI Puaps to Bot Legs) 23SJ1S6 Safety Injection (B.. fro* SI Puaps to Hot Leg*)

2UJ156 Safety Injection (B.P. troa Sl Puap1 to Bot Legs) 21SJU4 Safety Injection (B.P. fro* SI Pusp1 to Cold Logs) 1 22SJ1U Safety Injection (B.P. fro* SI Puap1 to Cold Ltgs}

23SJU4 Sdety Injection (!.P. tro* SI PUIIJII to Cold Legs) 24SJ1U Safety Injection (B.P. fro* SI Puap1 to Cold Legs) 2RB1 RBl suction froa Sot Leg No. 21 2Rll2 RBl Suetion fro* Rot Leg Mo. 21 23RH27 RBl D isc ha rge to Bet Leg No. 23 2UR27 RBl Di1cbarge to Hot Leg No. 24 R.doc4 +1> ..f.k_ T.u/,;ni CA-l f<t,.._,;.c.s "'

lhw -p 'tv*Jvthll'M<tlj IJ-t bi.,.,k, 3/4 4 .. 19 .

AaendJlent Mo 106

LR-N17-0121 LAR S17-04 Attachment 3 Mark-up of Proposed Technical Specification Bases Pages The following Technical Specifications pages for Renewed Facility Operating License DPR-70 are affected by this change request:

Technical Specification 3/4.4.6.3, Primary Coolant System Pressure Isolation Valves (new) B 3/4 4-4b The following Technical Specifications pages for Renewed Facility Operating License DPR-75 are affected by this change request:

Technical Specification Bases 3/4.4.7.2, Operational Leakage B 3/4 4-4 1

LR-N17-0121 LAR 817-04 BASES Insert 1 :

3/4.4.6.3 REACTOR COOLANT SYSTEM (RCS) PRESSURE ISOLATION VALVES (PIV)

The function of the RCS PIVs is to separate the high pressure RCS from the attached low pressure systems. The PIV leakage limit applies to each individual valve listed in the Technical Requirements Manual. Leakage through both series PIVs in a line must be included as part of the IDENTIFIED LEAKAGE, governed by LCO 3.4.6.2, "Operational Leakage." This is true during operation only when the loss of RCS mass through two series valves is determined by a water inventory balance (SR 4.4.6.2.d). A known component of the IDENTIFIED LEAKAGE before operation begins is the least of the two individual leak rates determined for leaking series PIVs during the required surveillance testing; leakage measured through one PIV in a line is not RCS operational leakage if the other is leaktight.

Although this specification provides a limit on allowable PIV leakage rate, its main purpose is to prevent overpressure failure of the low pressure portions of connecting systems. The leakage limit is an indication that the PIVs between the RCS and the connecting systems are degraded or degrading. PIV leakage could lead to overpressure of the low pressure piping or components. Failure consequences could be a loss of coolant accident (LOCA) outside of containment, an unanalyzed accident, that could degrade the ability for low pressure injection.

Bases Insert 2:

The function of the RCS PIVs is to separate the high pressure RCS from the attached low pressure systems. The PIV leakage limit applies to each individual valve listed in the Technical Requirements Manual. Leakage through both series PIVs in a line must be included as part of the IDENTIFIED LEAKAGE, governed by LCO 3.4.7.2, "Operational Leakage." This is true during operation only when the loss of RCS mass through two series valves is determined by a water inventory balance (SR 4.4.7.2.1 .d). A known component of the IDENTIFIED LEAKAGE before operation begins is the least of the two individual leak rates determined for leaking series PIVs during the required surveillance testing; leakage measured through one PIV in a line is not RCS operational leakage if the other is leaktight.

Although this specification provides a limit on allowable PIV leakage rate, its main purpose is to prevent overpressure failure of the low pressure portions of connecting systems. The leakage limit is an indication that the PIVs between the RCS and the connecting systems are degraded or degrading. PIV leakage could lead to overpressure of the low pressure piping or components. Failure consequences could be a loss of coolant accident (LOCA) outside of containment, an unanalyzed accident, that could degrade the ability for low pressure injection.

2

REACTOR COOLANT SYSTEM BJI_SES 3/4.4.6.2 OPERP.TIONAL LEAKAGE (Continued) its potential consequences. It should be noted that leakage past seals and gaskets is not pressure boundary leakage. The reactor must be brought to HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within 36 hours. This action reduces the leakage and also reduces the factors that tend to degrade the pressure boundary. The action times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. In COLD SHUTDOWN, the pressure stresses acting on the RCPB are much lower, and further deterioration is much less likely.

Surveillances Verifying RCS leakage to be within the LCO limits ensures the integrity of the Reactor Coolant Pressure Boundary is maintained. Pressure boundary leakage would at first appear as unidentified leakage and can only be positively identified by inspection. It should be noted that leakage past seals and gaskets is not pressure boundary leakage. Unidentified leakage and identified leakage are determined by performance of an RCS water inventory balance. The RCS water inventory must*be met with the reactor at steady state conditions.

The surveillance is modified by a Note that the surveillance is not required to be performed until 12 hours after establishing steady state operation. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowance provides sufficient time to collect and process all necessary data after stable plant conditions are established. Steady state operation is required to perform a proper inventory balance since calculations during maneuvering are not useful. For RCS operational leakage determination by water inventory balance, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and Reactor Coolant Pump seal injection and return flows. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

Mode ascension to MODE 1-3 is acceptable without a current RCS Inventory Balance, provided the asterisked note of "Not required to be completed until 12 hours after establishment of steady state operations", is complied with.

Satisfying the primary-to-secondary leakage limit ensures that the operational leakage performance criterion in the Steam Generator Program is met. If SR 4.4.6.2.c is not met, compliance with LCG 3.4.5, "Steam Generator Tube Integrity," should be evaluated. The 150 gallons per day limit is measured at room temperature (in accordance with EPRI PWR Primary-to-Secondary Leak Guidelines). If it is not practical to assign the leakage to an individual steam generator, all the primary-to-secondary leakage should be conservatively assumed to be from one Steam Generator. The Surveillance is modified by a Note which states that the surveillance is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation. For RCS primary-to-

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-----------*--se*c -reaK.a e--det rmlnc:reic:m*;-st-e-a-cty--s *at - ---:t-s--d - *::i:n -d-a:s--s-t-ab-l-ondary g e t e ef e e-Re-s pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and Reactor Coolant Pump seal injection and return flows. The Surveillance Frequency is based on operating experience, equipment reliability,

[ J and plant risk and is controlled under the Surveillance Frequency Control n i

+--

Program. The primary-to-secondary leakage is determined using continuous process raciation monitors or radiochemical grab sampling (in accordance with EPRI PWR Primary-to-Secondary Leak Guidelines).

3/4.4.7 THIS SECTION DELETED SALEM - UNIT 1 B 3/4 4-4b Amendment No. 299 (PSEG Issued)

REAC'I' O R COOLANT S YS T EM BA S E S 3/ 4 . 4 . 7 REACTOR COOLANT S YSTEM LEAKAGE 3/4 . 4 . 7 . 1 LEAKAGE DETECTION S Y S T EMS The RCS l e a k a ge de t e ction s y s t ems required by thi s speci f i c a t i on a re provided t o mon i t o r and de t e ct l e a kage f rom t h e Re a c t o r C o o l a n t P r e s s u re B o u nd a r y . Th e s e d e t e c t i on s ys t ems a r e cons i s t e n t wi t h the r e commenda t i o n s o f Re gulat o r y Guide 1 . 45,

" Re a cto r C o o l ant P r e s s u r e Bo u n d a r y Lea kage D e t e ct i o n S y s t ems , " Ma y 1973 .

3/4 . 4 . 7 . 2 O PERAT I ONAL LEAKAGE Indu s t ry exp e r i e n ce h a s s h own t hat whi l e a l imi t e d amount of l e a k age is e xp e c t ed from the R C S , t h e unidenti f i ed p o r t i o n o f t h i s l eakage can b e redu ced t o a thre s h o l d v a l u e of less than 1 GPM . T hi s t h r e s ho l d va l u e is s u f fi c i e n t l y l ow to ensure e a r l y de t e c t i on o f a ddi t i on a l leakage .

The 10 GPM I DENT I F I ED L EAKAGE l imi t a t i o n p ro vi d e s a l l owance fo r a limi t ed amo unt o f l e a kag e from known s ources wh o s e p r e s e n c e will n o t int e r f e r e w i t h t h e d e t e ct i on of UNI D ENT I FI ED L EAKAGE by the l e a k a g e d e t e c t i o n s ys t ems .

--=----*----*------...----**....----*-... ... * -------*---................... ....

rsora t i o n

,H-l-01-):J..C;!. i remen t s f o r R C S P r e u r; Va l ve s p r ovide added as s urance of valve inte gri fYt;h-e- ¥i n g t h e probab i l i ty o f g r o s s valve f a i l u r e a n d co ns e qu en t i n t e r s y s t em LOCA. Le'a-hrI(l the RCS Pres s ure I s olation Va l v e s is I D ENTI FI ED LKAGE a n d wi l l b e cons idered as llowed limi t .

P RE S SURE BOUNDARY L EAKAGE o f any magnitude is unaccep table s in c e it may b e indi cative o f an impending gro s s fail u r e o f the p r e s s u r e bounda ry .

T h e r e fore , the p r e s en c e o f a n y PRES SURE BOUNDARY LEAKAGE requ i r e s t h e un i t to be p r omp t l y p l a ced in COLD S HUTDOWN .

Prima r y to S e cond a ry Lea kage Throu9h Any One S G T h e p r ima r y - t o - s e c ondary l e a ka g e ra t e limit a pp li e s t o l e a kage t h ro u gh any one S t e am Genera t o r . The l imit of 1 5 0 g a l l ons per day p e r s t eam gen e ra t o r is b a s ed on the op e r a t i onal l e a k a g e p e r f o rmance c ri t e ri o n i n NEI 97 0 6 , S t e am Gene r a t o r P r og r am Gui d e l i n e s . The S t eam Gene r a t o r P rog ram o p e r a t i o n a l l e akage p e r f o rma nce cri t e r ion in N E I 9 7 - 0 6 s t a t e s , "The RCS op e r a t i onal p rima ry-t o - s e conda ry l e a k a g e through a ny one SG s h a l l be limi t e d to 1 5 0 ga l l on s p e r d ay . n The l imit i s b a s ed on ope r a t ing expe ri e n c e wi th s te am g ene r a t o r t ube degradat i on me ch ani s ms that re s ul t in tube lea kage . The op e r a t i o n a l l e a kage rate c r i t e r i on i n conj u n c t i o n with the imp l emen t at i o n of the S t e am Gen e r a t o r P r o g r am is an e f f e ctive meas u r e for mi n i mi z i ng the frequency o f s te am genera t o r tube rup t u r e s . The d o s a g e con t ribut i on from t he tube l e a k a g e will be wi thin 10 CFR 5 0 . 6 7 l i mi t s in t h e event of e i th e r a s t eam g ene rator tub e rupture o r s t e am l i n e bre a k . The anal y s e s a r e b a s e d on the

• t o t a l p r ima ry to s e condary l ea ka g e f rom all SGs o f 1 g a l l o n per mi n u t e a s a r e s u l t of a c c i d e n t i nduced condi t i o n s .

[ sc SALEM - UN I T 2 B 3/4 4-4 Arnendmen t No . 2 62

( P S E G I s s ue d )