Text

License Amendment Request to Correct Technical Specification and Facility Operating License Editorial Items
	ML092680244
Person / Time
Site:	Salem
Issue date:	09/14/2009
From:	Braun R; Public Service Enterprise Group
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	LAR S09-03, LR-N09-0164
	Download: ML092680244 (66)
	v • d • e

PSEG Nuclear LLC P.O. Box 236,, Hancocks Bridge, NJ 08038-0236 0 PSEG NuclearL.L. C.

SEP 1 4 2009 10 CFR 50.90 LR-N09-0164 LAR S09-03 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Salem Generating Station, Unit 1 and 2 Facility Operating License Nos. DPR-70 and DPR-75 NRC Docket Nos. 50-272 and 50-311

Subject:

License Amendment Request to Correct Technical Specification and Facility Operating License Editorial Items In accordance with the provisions of 10CFR50.90, PSEG Nuclear LLC (PSEG) requests an amendment to the facility operating licenses listed above. In accordance with 10CFR50.91(b)(1), a copy of this submittal has been sent to the State of New Jersey.

The proposed changes correct editorial items in the Technical Specifications (TS) and Facility Operating License (FOL) for Salem Units 1 and 2. These items are either historical in nature, or are errors inadvertently created during the submittal of other license amendment requests and subsequent license amendments. The proposed changes are administrative in nature and fall into one of four categories: (1) correct typographical errors, (2) correct format errors, (3) correct administrative differences between Units, or (4) delete historical requirements that have expired.

The TS affected are 3.1.3.3 ACTION b (Unit 1 and 2), 3.2.1 (Unit 2), Surveillance Requirement (SR) 4.2.2.2 (Unit 1 and 2), Table 3.2-1 (Unit 1), Table 3.3-6 (Unit 1 and 2), Table 4.3-3 (Unit 1),

3.3.3.14 (Unit 1 and 2), SR 4.4.11.2 (Unit 2), 6.9.1.5.c (Unit 2), 6.9.1.9 (Unit 2), Amendment 222 (Unit 1) one time changes (various SR), and Amendment 230 (Unit 1) one time changes (3.1.3.1.2, SR 4.1.3.1.1). The FOL sections affected are Unit 1 FOL Attachment 1 and Unit 2 FOL Condition 2.C.3 through 2.C.9 and 2.C.11 through 2.C.25.

Attachment I of this submittal provides an evaluation supporting the proposed changes. provides the marked-up TS and FOL pages, with the proposed changes indicated.

No regulatory commitments are contained in this submittal.

The changes in this LAR are not required to address an immediate safety concern; PSEG requests approval of this 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.

Document Control Desk Page 2 LR-N09-0164 If you have any questions or require additional information, please do not hesitate to contact Mr.

Jeff Keenan at (856) 339-5429.

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

Executed on 91,/A /o9 (Date)

Sincerely, Robert C. Braun Site Vice President Salem Generating Station Attachments (2)

S. Collins, Regional Administrator - NRC Region I R. Ennis, Project Manager - USNRC NRC Senior Resident Inspector - Salem P. Mulligan, Manager IV, NJBNE Commitment Coordinator - Salem PSEG Corporate Commitment Manager

Attachment I LAR S09-03 LR-N09-0164 License Amendment Request to Correct Technical Specification and Facility Operating License Editorial Items

'Table of Contents

1. DE S C R IPT IO N................................................................................................................ 2

2. PRO PO SED C HA NG ES ............................................................................................ 2

3. BACKGROUND .................................................................................... 7

4. TEC HNICA L A NA LYS IS ........................................................................................... 7

5. REG ULATO RY ANALYSIS ........................................................................................ 9

6. ENVIRONMENTAL CONSIDERATION .................................................................... 11

7. R E F E R E NC E S .............................................................................................................. 11 1 of 11

Attachment 1 LAR S09-03 LR-N09-0164

1.0 DESCRIPTION

In accordance with the provisions of 10CFR50.90, PSEG Nuclear LLC (PSEG) requests an amendment to the facility operating licenses DPR-70 and DPR-75.

The proposed changes correct editorial items in the Technical Specifications (TS) and Facility Operating License (FOL) for Salem Units 1 and 2. These items are either historical in nature, or are errors inadvertently created during the submittal of other license amendment requests and subsequent license amendments. The proposed changes are administrative in nature and fall into one of three categories: (1) correct typographical errors, (2) correct format errors, (3) correct administrative differences between Units, or (4) delete historical requirements that have expired.

The TS/FOL affected sections are 3.1.3.3 ACTION b (Unit 1 and 2), 3.2.1 (Unit 2), Surveillance Requirement (SR) 4.2.2.2 (Unit 1 and 2), Table 3.2-1 (Unit 1), Table 3.3-6 (Unit 1 and 2), Table 4.3-3 (Unit 1), 3.3.3.14 (Unit 1 and 2), SR 4.4.11.2 (Unit 2), 6.9.1.5.c (Unit 2), 6.9.1.9 (Unit 2),

Amendment 222 (Unit 1) one time changes (various SR), and Amendment 230 (Unit 1) one time changes (3.1.3.1.2, SR 4.1.3.1.1). The FOL sections affected are Unit 1 FOL Attachment 1 and Unit 2 FOL Condition 2.C.3 through 2.C.9 and 2.C.1 1 through 2.C.25.

2.0 PROPOSED CHANGE

S Item Description TS I FOL Unit Action 1 Amendments 201/197 (LCR 94-41)- 3.1.3.3 1 and 2 Delete TS 3.1.3.3 ACTION b The Margin Recovery Program removed ACTION b 3-loop operation references in TS; 3- (Note: Another revision to TS Loop operation has not been analyzed 3.1.3.3 is pending via PSEG for Salem. TS 3.1.3.3, ACTION b, was LAR S09-01, submitted March not changed (deleted) due to an 22, 2009) oversight.

TS 3.1.3.3 ACTION b provides requirements if rod drop time is determined with 3 reactor coolant pumps running in Modes 1 and 2. TS 3.4.1.1 requires that all reactor coolant pumps be in operation in Modes 1 and 2.

(Reference 1 and 2)

2. LCO 3.2.1 contains a typographical error 3.2.1 2 Correct typographical error in the acronym for the term "Core Operating Limits Report". The acronym should be "COLR"; instead it is represented as "CORL" 2 of 11

Attachment 1 LAR S09-03 LR-N09-0164 Item Description TS / FOL Unit Action

3. The peaking factor term in SR 4.2.2.2 is SR 4.2.2.2, 1 and 2 Correct subscript and incorrectly represented in some cases. page 3/4 2- superscript errors in SR 4.2.2.2, Standard convention has the subscript 7 page 3/4 2-7.

followed by the superscript. The subscript and superscript are reversed in some cases on page 3/4 2-7 of SR 4.2.2.2.

4. SR 4.2.2.2 contains the following note SR 4.2.2.2 2 Delete historical Cycle 11 related to Cycle 11: footnote

"*For Cycle 11, when the number of available movable detector thimbles is greaterthan or equal to 50% and less than 75% of the total, the 5%

measurement uncertaintyshall be increased to [5% + (3-T/14.5)(1%)]

where T is the number of available thimbles."

This note is no longer applicable to subsequent operating cycles; current Salem Unit 2 cycle is Cycle 17.

5. A footnote in Unit 1 TS Table 3.2-1 Table 3.2-1 1 Delete the word "increase" (2 states that the pressurizer pressure limit instances) from the footnote in is not applicable "during either Unit 1 TS Table 3.2-1 THERMAL POWER ramp increase in excess of 5% RATED THERMAL POWER per minute or a THERMAL POWER step increase in excess of 10%

RATED THERMAL POWER."

The corresponding footnote in Unit 2 TS Table 3.2-1 states that the limit is not applicable "during either a THERMAL POWER ramp in excess of 5% RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10% RATED THERMAL POWER."

The word 'increase' in the Unit 1 TS is superfluous and inconsistent with Unit 2 TS, and inconsistent with NUREG 1431 (page 3.4.1-1).

3 of 11

Attachment 1 LAR S09-03 LR-N09-0164 Item Description TS / FOL Unit Action

6. Amendment 280 & 263 (LCR S06-03) - Table 3.3-6 1 and 2 Correct format and The TS camera ready pages prepared typographical errors in Table for LCR S06-03 contained format and 3.3-6.

typographical errors in TS Table 3.3-6; these errors were subsequently included in the TS pages issued with the amendments. Specifically, some symbols were incorrectly reproduced (Unit 1 and 2), and a table heading was misaligned (Unit 2)

(Reference 3)

7. Amendment 225 - a typographical error Table 4.3-3 1 Correct footer alignment was noted in the footer of Table 4.3-3, page 3/4 3-38a; the number "1" is missing following the word "Unit" (the "1" has been tabbed-over next to the page number in the center of the footer).

S(Reference 4)

8. Amendment 282/265 (LCR S05-12) - 3.3.3.14 1 and 2 Revise TS 3.3.3.14 to remove These amendments approved the the reference to the deleted TS relocation of Incore Detection Monitoring 3.3.3.2. The TS will be revised TS 3/4.3.3.2 to the UFSAR. During the as follows: '...and use the amendment implementation it was incore movable detector identified that the relocated TS 3.3.3.2 is system, satisfying-the also referenced in TS 3.3.3.14, Power OPERABILIT'Y .e.uir.mnt.

Distribution Monitoring System (PDMS). listed in Specification 3.3.3.2, to The action in TS 3.3.3.14 is to refer to obtain any required core power TS 3.3.3.2 if the PDMS was declared distribution measurements.'

inoperable and the incore moveable detector system was to be used.

LCR S05-12 was submitted on June 30, 2006 (PSEG letter LR-N06-0079).

(Reference 5) 4 of 11

Attachment 1 LAR S09-03 LR-N09-0164 Item Description TS I FOL Unit Action

9. SR 4.4.11.2 - the SR is historical and SR 2 Delete SR 4.4.11.2 applicable only to the first 3 cycles of 4.4.11.2 operation:

"Augmented Inservice Inspection Programfor Steam GeneratorChannel Heads - The No. 21 Steam Generator channel head shall be ultrasonically inspected in a selected area during each of the first three refueling outages using the same ultrasonicinspection procedures and equipment used to generate the baseline data. These inservice ultrasonicinspections shall verify that the cracks observed in the stainless steel claddingprior to operation have not propagatedinto the base material."

The SR is no longer applicable to cycles subsequent to Cycle 3 (current Salem Unit 2 cycle is Cycle 17.)

10. Incorrect TS

Reference:

TS 6.9.1.5.c 6.9.1.5.c 2 Correct reference to TS 3.4.9 references TS 3.4.8, however it should reference TS 3.4.9 (Reactor Coolant System Specific Activity) for Unit 2.

Section 6.9.1.5 is an administrative section of TS that addresses periodic reports.

11. Incorrect TS

Reference:

TS 6.9.1.9.a.1 6.9.1.9 2 Correct reference to TS references TS 3/4.1.1.4; however it 3/4.1.1.3 should reference 3/4.1.1.3 (Moderator Temperature Coefficient) for Unit 2.

Section 6.9.1.9 is an administrative section of TS that addresses periodic reports.

12. Amendment 222 allowed a one-time See 1 Delete historical Cycle 13 notes extension of the TS surveillance interval description to the end of fuel Cycle 13 for certain TS surveillance requirements (SRs).

Specifically, the amendment extended the surveillance interval in (a) SR 4.3.2.1.3; (b) SRs 4.8.2.3.2.f and 4.8.2.5.2.d; (c) SR 4.8.2.5.2.c.2; (d) SR 4.8.3.1.a.1.; (e) SR 4.1.2.2.c; (f) SRs 4.3.1.1.1, Table 4.3-1, 4.3.2.1.1, Table 4.3-2, and 4.3.3.7, Table 4.3-11; (g) SR 4.5.1.d; (h) SR 4.5.2.e.1; (i)SR 4.7.6.1.d.2; (j) SR 4.7.10.b; and (k) SR 5 of 11

Attachment 1 LAR S09-03 LR-N09-0164 Item Description TS I FOL Unit Action 4.8.1.1.2.d.7. Because of the length of outage 1 R1 2 and delays in restart, the SRs would have been overdue prior to reaching the next refueling outage (1R13).

These Cycle 13 notes are no longer applicable to subsequent operating cycles; current Salem Unit 1 cycle is Cycle 20. Note that the Cycle 13 notes for item (a), SR 4.3.2.1.3, (b), SR 4.8.2.5.2.d (only), (c), SR 4.8.2.5.2.c.2, and (k), SR 4.8.1.1.2.d.7, have already been removed from TS.

(Reference 6)

13. Amendment 230 modified TS 3.1.3.2.1, 3.1.3.2.1 1 Delete historical Cycle 14 notes Action a. 1, to determine the position of 4.1.3.1.1 Rod 1 SB2 from once every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to 4.1.3.4 within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following any movement of the rod until repair of the rod indication system was completed. This change was applicable for the remainder of Unit 1 Cycle 14, or until an outage of sufficient duration occurred to repair the system.

Surveillance Requirements (SR) of TSs 4.1.3.1.1 and 4.1.3.4 were also modified to require that the position of Rod 1 SB2 be determined (by the incore system) only following movement of the rod until repair of the indication system was completed.

These requirements are historical, they were only applicable for Cycle 14 and are no longer applicable to subsequent operating cycles; current Salem Unit 1 cycle is Cycle 20.

(Reference 7)

14. Unit 1 FOL Attachment 1, "Incomplete FOL 1 Delete FOL Attachment 1 and preoperational Tests, Startup Tests, and Attachment the reference to Attachment 1 Other Items Which Must be Completed" 1 on FOL page 5c.

is a listing of required actions related to Unit 1 initial plant start-up. These items are all historical and can be removed from the FOL.

6 of 11

Attachment 1 LAR S09-03 LR-N09-0164 Item Description TS / FOL Unit Action

15. Unit 2 FOL Condition 2.C.3 through FOL 2.C 2 Delete FOL items 2.C.3 through 2.C.9 and 2.0.11 through 2.C.25 (Unit 2.0.9 and 2.C.11 through 2). These conditions comprise actions 2.0.25.

related to (a) Unit 2 initial plant start-up and operation through the first cycle and refueling outage, (b) the second refueling outage, (c) issuance of the license, or (d) actions required to be completed prior to June 1, 1983. These items are all historical and can be removed from the FOL.

The marked up TS and FOL pages are provided in Attachment 2.

3.0 BACKGROUND

Background information is provided in the Table in Section 2.0

4.0 TECHNICAL ANALYSIS

The proposed changes are administrative in nature that correct (1) typographical errors, (2) correct format errors, (3) correct administrative differences between Units, or (4) delete historical requirements that have expired, as discussed below.

The proposed deletion of Salem Unit 1 and 2 TS 3.1.3.3, ACTION b is editorial in nature, correcting an oversight in LCR 94-21. LCR 94-41, The Margin Recovery Program, and the subsequent Amendments 201 and 197 removed 3-loop operation references in TS; 3-Loop operation has not been analyzed for Salem. Since 3-loop operation is not approved for Salem, there should be no references for its use in TS. TS 3.1.3.3 ACTION b provides requirements if rod drop time is determined with 3 reactor coolant pumps running in Modes 1 and 2. TS 3.4.1.1 requires that all reactor coolant pumps be in operation in Modes 1 and 2. TS 3.1.3.3, ACTION b, was not changed (deleted) due to an oversight in the preparation of the TS markups in LCR 94-21.

The peaking factor term in SR 4.2.2.2, is incorrectly represented in some cases. Standard convention has the subscript followed by the superscript. The subscript and superscript are reversed in some cases on page 3/4 2-7 of SR 4.2.2.2; this submittal will correct those errors.

The proposed change to LCO 3.2.1 is editorial in nature, correcting a typographical error. The acronym for the term "Core Operating Limits Report" is represented as "CORL"; instead it should be "COLR".

The proposed change to Salem Unit 2 TS 4.2.2.2 is editorial in nature, eliminating a condition approved only for Cycle 11, which has expired.

The proposed change to Salem Unit 1 TS Table 3.2-1 is editorial in nature, correcting a typographical error; the word 'increase' is inserted superfluously in two places. A footnote in Unit 1 TS Table 3.2-1 states that the pressurizer pressure limit is not applicable "during either THERMAL POWER ramp increase in excess of 5% RATED THERMAL POWER per minute or a 7 of 11

Attachment I LAR S09-03 LR-N09-0164 THERMAL POWER step increase in excess of 10% RATED THERMAL POWER." The corresponding footnote in Unit 2 TS Table 3.2-1 states that the limit is not applicable "during either a THERMAL POWER ramp in excess of 5% RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10% RATED THERMAL POWER." The word 'increase' in the Unit 1 TS is superfluous and inconsistent with Unit 2 TS, and inconsistent with NUREG 1431 (page 3.4.1-1).

The proposed change to Salem Unit 1 and 2 TS Table 3.3-6 is editorial in nature, correcting typographical and formatting errors. The TS camera ready pages prepared for LCR S06-03 (Amendments 280 and 263) contained format and typographical errors in TS Table 3.3-6; these errors were subsequently included in the TS pages issued with the amendments. Specifically, two symbols (< and [t) were incorrectly reproduced (Unit 1 and 2), and a table heading was misaligned (Unit 2).

The proposed change to Salem Unit 1 Table 4.3-3 is editorial in nature, correcting a typographical error. Following the implementation of Amendment 225 a typographical error was noted in the footer of Table 4.3-3, page 3/4 3-38a; specifically the number "1" is missing following the word "Unit" (the "1" has been tabbed-over next to the page number in the center of the footer). This submittal will correct the error.

The proposed change to Salem Unit 1 and 2 TS 3.3.3.14, is editorial in nature, correcting an oversight in LCR S05-12. LCR S05-12 and subsequent Amendments 282 and 265 relocated the Incore Detection Monitoring TS 3/4.3.3.2 to the UFSAR (PSEG intends to ultimately place the requirements in the Technical Requirements Manual (TRM)). During the amendment implementation it was identified that the relocated TS 3.3.3.2 is also referenced in TS 3.3.3.14, Power Distribution Monitoring System (PDMS). The action in TS 3.3.3.14 is to refer to TS 3.3.3.2 if the PDMS was declared inoperable and the incore moveable detector system was to be used. This submittal will revise TS 3.3.3.14 to delete the reference to TS 3.3.3.2 and state that if the PDMS is declared inoperable then the incore movable detector system should be used to obtain any required core power distribution measurements. The requirements for the incore movable detector system have been relocated to a document controlled by 10CFR 50.59 (e.g., the UFSAR or TRM).

The proposed change to Salem Unit 2 SR 4.4.11.2 is editorial in nature, eliminating a condition, approved only for the first 3 refueling outages, which has expired. The SR is historical and no longer applicable to cycles subsequent to Cycle 3 (current Salem Unit 2 cycle is Cycle 17).

The proposed change to Salem Unit 2 TS 6.9.1.5.c is editorial in nature, correcting a typographical error. TS 6.9.1.5.c incorrectly references TS 3.4.8; it should reference TS 3.4.9 (Reactor Coolant System Specific Activity) for Unit 2. Section 6.9.1.5 is an administrative section of TS that addresses periodic reports.

The proposed change to Salem Unit 2 TS 6.9.1.9.a.1 is editorial in nature, correcting a typographical error. TS 6.9.1.5.c incorrectly references TS 3 TS 3/4.1.1.4; it should reference TS 3/4.1.1.3 (Moderator Temperature Coefficient) for Unit 2. Section 6.9.1.9 is an administrative section of TS that addresses periodic reports.

The proposed change to the SRs changed by Amendment 222 is editorial in nature, eliminating conditions approved only for Cycle 13, which has expired.

8 of 11 LAR S09-03 LR-N09-0164 The proposed change to the TS and SR changed by Amendment 230 is editorial in nature, eliminating conditions approved only for Cycle 14, which has expired.

Unit 1 FOL Attachment 1, "Incomplete preoperational Tests, Startup Tests, and Other Items Which Must be Completed" is a listing of required actions solely related to Unit 1 initial plant start-up. These items are all historical and can be removed from the FOL.

Unit 2 FOL Condition 2.C.3 through 2.C.9 and 2.C.1 1 through 2.C.25 comprise actions solely related to (a) Unit 2 initial plant start-up and operation through the first cycle and refueling outage, (b) the second refueling outage, (c) issuance of the license, or (d) actions required to be completed prior to June 1, 1983. These items are all historical and can be removed from the FOL.

5.0 REGULATORY ANALYSIS

10 CFR 50.36 (a)(1) requires that each applicant for a license authorizing operation of a production or utilization facility shall include in his application proposed TS in accordance with the requirements of section 50.36. The TS are part of the FOL and any changes to the FOL and TS must be in accordance with 10 CFR 50.90. The corrections proposed by this license amendment request conform to these regulations.

5.1 No Significant Hazards Consideration PSEG requests an amendment to the Salem Unit 1 and 2 Operating Licenses. The proposed changes correct editorial items in the Technical Specifications (TS) and Facility Operating License (FOL) for Salem Units 1 and 2. These items are either historical in nature, or are errors inadvertently created during the submittal of other license amendment requests and subsequent license amendments. The proposed changes are administrative in nature and fall into one of four categories: (1) correct typographical errors, (2) correct format errors, (3) correct administrative differences between Units, or (4) delete historical requirements that have expired.

The TS involved are 3.1.3.3 ACTION b (Unit 1 and 2), 3.2.1 (Unit 2), Surveillance Requirement (SR) 4.2.2.2 (Unit 1 and 2), Table 3.2-1 (Unit 1), Table 3.3-6 (Unit 1 and 2), Table 4.3-3 (Unit 1),

3.3.3.14 (Unit 1 and 2), SR 4.4.11.2 (Unit 2), 6.9.1.5.c (Unit 2), 6.9,1.9 (Unit 2), Amendment 222 (Unit 1) one time changes (various SR), and Amendment 230 (Unit 1) one time changes (3.1.3.1.2, SR 4.1.3.1.1). The FOL sections affected are Unit 1 FOL Attachment 1 and Unit 2 FOL Condition 2.C.3 through 2.C.9 and 2.C.11 through 2.C.25.

PSEG has evaluated the proposed changes to the TS and FOL for the stations listed above, using the criteria in 10CFR50.92, 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 consideration.

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

Response: No The proposed changes to TS and the FOL are administrative in nature that correct typographical errors, correct format errors, correct inconsistencies between Units, or delete 9 of 11 LAR S09-03 LR-N09-0164 historical requirements that have expired. These changes do not affect the intent of any TS requirements.

The proposed change does not have any impact on structures, systems and components (SSCs) of the plant, and no affect on plant operations. The proposed change does not impact any accident initiators or analyzed events or assumed mitigation of accident or transient events. They do not involve the addition or removal of any equipment, or any design changes to the facility. Therefore, this proposed change does not represent a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No The proposed changes to TS and the FOL are administrative in nature that correct typographical errors, correct format errors, correct inconsistencies between Units, or delete historical requirements that have expired. These changes do not affect the intent of any TS requirements.

The proposed change does not involve a modification to the physical configuration of the plant (i.e., no new equipment will be installed) or change in the methods governing normal plant operation. The proposed change will not impose any new or different requirements or introduce a new accident initiator, accident precursor, or malfunction mechanism.

Additionally, there is no change in the types or increases in the amounts of any effluent that may be released off-site and there is no increase in individual or cumulative occupational exposure. 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 TS and the FOL are administrative in nature that correct typographical errors, correct format errors, correct inconsistencies between Units, or delete historical requirements that have expired. These changes do not affect the intent of any TS requirements.

The proposed change incorporates corrections to the TS and FOL and result in improved accuracy of these licensing documents. There is no change to any design basis, licensing basis or safety limit, no change to any parameters; consequently no safety margins are affected. Therefore, 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 10CFR50.92 (c), and, accordingly, a finding of no significant hazards consideration is justified.

In conclusion, based on the considerations discussed above, (1) there is a reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the NRC's 10 of 11

Attachment I LAR S09-03 LR-N09-0164 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 10 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 a 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 need be prepared in connection with the proposed amendment.

7.0 REFERENCES

(1) Letter from NRC to PSEG: "Salem Nuclear Generating Station, Unit 1 (TAC NO M95383)", Amendment 201, dated November 26, 1997. (ML011720441)

(2) Letter from NRC to PSEG: "Salem Nuclear Generating Station, Unit 2 (TAC NO M95384)", Amendment 197, dated January 8, 1999. (ML0117230279)

(3) Letter from NRC to PSEG: "SALEM NUCLEAR GENERATING STATION, UNIT NOS. 1 AND 2, ISSUANCE OF AMENDMENTS RE: ACCIDENT MONITORING INSTRUMENTATION AND SOURCE CHECK DEFINITION (TAC NOS. MD1654, MD1655, MD1656 AND MD1657)", Amendment 280/263, dated April 19, 2007.

(ML070920317)

(4) Letter from NRC to PSEG: "SALEM NUCLEAR GENERATING STATION, UNIT NO. 1 AND 2, ISSUANCE OF AMENDMENT RE: ADMINISTRATIVE AND EDITORIAL CHANGES (TAC NOS. MA0180 AND MA0181)", Amendment 225/206, dated November 2, 1999. (ML993240246)

(5) Letter from NRC to PSEG: "SALEM NUCLEAR GENERATING STATION, UNIT NOS. 1 AND 2, ISSUANCE OF AMENDMENTS RE: RELOCATION OF TECHNICAL SPECIFICATION REQUIREMENTS FOR THE MOVABLE INCORE DETECTORS AND RADIOACTIVE GASEOUS EFFLUENT OXYGEN MONITORING INSTRUMENTATION (TAC NOS. MD2505 AND MD2506)", Amendment 282/265, dated June 6, 2007.

(ML071200393)

(6) Letter from NRC to PSEG: "SALEM NUCLEAR GENERATING STATION, UNIT NO. 1 -

ISSUANCE OF AMENDMENT RE: ONE-TIME EXTENSION OF SURVEILLANCE INTERVAL (TAC NO. MA4554)", dated May 4, 1999 (ML011730078)

(7) Letter from NRC to PSEG: "SALEM NUCLEAR GENERATING STATION, UNIT NO. 1, ISSUANCE OF AMENDMENT RE: EXIGENT REQUEST FOR CHANGES TO TECHNICAL SPECIFICATIONS -POSITION INDICATION SYSTEM (TAC NO.

MA8840)", dated May 26, 2000 (ML003719424) 11 of 11 LAR S09-03 LR-N09-0164 TECHNICAL SPECIFICATION PAGES WITH PROPOSED CHANGES The following Technical Specifications and Facility Operating License pages for Facility Operating License DPR-70 are affected by this change request:

Technical Specification Page 4.1.2.2.c 3/4 1-9 4.1.3.1.1 3/4 1-18a 3.1.3.2.1 3/4 1-19 3.1.3.3 3/4 1-21 4.1.3.4 3/4 1-22 4.2.2.2 3/4 2-7 Table 3.2-1 3/4 2-14 Table 4.3-1 3/4 3-13 Table 4.3-2 3/4 3-31a Table 3.3-6 3/4 3-36a Table 4.3-3 3/4 3-38a Table 4.3-11 3/4 3-57a 3.3.3.14 3/4 3-71 4.5.1 .d 3/4 5-2 4.5.2.e.1 3/4 5-5 4.7.6.1 3/4 7-21 4.7.10. b 3/4 7-34 4.8.2.3.2.f 3/4 8-9a 4.8.3.1 .a.1 3/4 8-14 Facility Operating License Page Attachments 5c 1 through 4 The following Technical Specifications and Facility Operating License pages for Facility Operating License DPR-75 are affected by this change request:

Technical Specification Page 3.1.3.3 3/4 1-18 3.2.1 3/4 2-1 4.2.2.2 3/4 2-7 Table 3.3-6 3/4 3-39a 3.3.3.14 3/4 3-66 4.4.11.2 3/4 4-33 6.9.1.5.c 3/4 6-21 6.9.1.9 3/4 6-24 Facility Operating License Page Section 2.C 4 through 21 LAR S09-03 LR-N09-0164 UNIT I TECHNICAL SPECIFICATION PAGES WITH PROPOSED CHANGES Facility Operating License DPR-70

REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

c. " At least once per 18 months during shutdown by verifying that each automatic valve in the flow path actuates to its correct position on a safety injection test signal.

d. At least once per 18 months by verifying that the flow path required by specification 3.1.2.2.a delivers at least 33 gpm to the Reactor Coolant System.

A A ene time e~tensien to this surveillanee requirement which in-satisfied by performanee of the Manual S! test is granted during fn eyele thirteen allowing Unit 1 eporotions to eentinue to the thrtnnt refueling eutage (IR13). The surveillanee testing is to be eompletod ý&ýat the apprepriato time during the !R13 eutage, prior to the unit retuniýPng to Mede 41upon eunage eemplktion.

SALEM - UNIT 1 3/4 1-9 Amendment No. 22-

a) A reevaluation of each accident analysis of table 3.1-1 is performed within 5 days; this reevaluation shall confirm that the previously analyzed results of these accidents remain valid for the duration of operation under these conditions.

b) The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c) A core power distribution measurement is obtained and FQ (Z) FNAH are verified to be within their limits within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

d) The THERMAL POWER level is reduced to less than or equal to 75% of RATED THERMAL POWER within one hour and within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the high neutron flux trip setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER. THERMAL POWER shall be maintained less than or equal to 75% of RATED THERMAL POWER until compliance with ACTIONS 3.1.3.1.c.3.a and 3.1.3.1.c.3.c above are demonstrated.

SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The position of each full length rod shall be determined to be within the limits established in the limiting condition for operation at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (allowing for one hour thermal soak after rod motion) except during time intervals when the Rod Position Deviation Monitor is inoperable, then verify the group positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .+/--

4.1.3.1.2 Each full length rod not fully inserted in the core shall be determined to be OPERABLE by movement of at least 10 steps in any one direction at least once per 31 days.

During Cyele 14, the pesitien of Red 1.9B2 will: be determined indlirootly by the moevoblo ineeiro doeetootor within 8 houro~ following its motvomotnt1 udntil:

the ropair of tho, i:ndication system for thios red. During roocter stortup, the fully withdrcawn pooition of Red 1SB2 wimll be doet-ermfinod by lur-Erant trococs and subsequently verified by the movablo incoro dotootoros prior to entry into Modo I.

SALEM - UNIT 1 3/4 1-18a Amendment No. 2-3q

REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEMS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.3.2.1 The shutdown and control rod position indication systems shall be OPERABLE and capable of determining the actual and demanded rod positions ,as follows:

a. Analog rod position indicators, within one hour after rod motion (allowance for thermal soak);

All Shutdnwn Banks: +/-18 steps at

85% reactor power or if reactor power is > 85% RATED THERMAL POWER +/- 12 steps of the group demand counters for withdrawal ranges of 0-30 steps and 200-228 steps.

Control Rank A: +/-18 steps at

85% reactor power or if reactor power is

> 85% RATED THERMAL POWER +/- 12 steps of the group demand counters for withdrawal ranges of 0-30 steps and 200-228 steps.

Control Bank B: +/-18 steps at

85% reactor power or if reactor power is

> 85% RATED THERMAL POWER +/- 12 steps of the group demand counters for withdrawal ranges of 0-30 steps and 160-228 steps.

Control Bank C and D: +/-18 steps at

85% reactor power or if reactor power is > 85% RATED THERMAL POWER +/- 12 steps of the group demand counters for withdrawal ranges of 0-228 steps.

b. Group demand counters; +/- 2 steps of the pulsed output of the Slave Cycler Circuit over the withdrawal range of 0-228 steps.

APPLICABILITY: MODES 1 and 2.

ACTION:

a. With a maximum of one analog rod position indicator per bank inoperable either:

1. Determine the position ofthe non-indicating rod(s) indirectly using the power distribution monitoring sy em (if power is above 25% RTP) or using the movable incore vectors (if power is less than 25% RTP or the power distributio monitoring system is inoperable) at least once per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s- and within one hour after any motion of the non-indicating rod which exceeds 24 steps in one direction since the last determination of the rod's position, or

2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

Durcing Cyole 14, the position of Red 1SB2 will4 bo dotormffinod indireotly by the moivable ineoro dotetoictr wit~hin 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months follewing its movomont until the repair of the indication system for this rod.

During r.a.t.r startup, the fully wit.hdr.awn position .B2 f Red will he doetormined by eurront troopes ondl subsequently vorifiodee by the

.movablo incors dototrs prior to entry into Mode I.

b. With two or more analog rod position indicators per bank inoperable, within one hour restore the inoperable rod position indicator(s) to OPERABLE status or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . A maximum of one rod position indicator per bank may remain inoperable following the hour, with Action (a) above being applicable from the original entry time into the LCO.

SALEM - UNIT 1 3/4 1-19 Amendment No. 2-34

REACTIVITY CONTROL SYSTEMS ROD DROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.3 The individual full length (shutdown and control) rod drop time from 228 steps withdrawn shall be

2.7 seconds from beginning of decay of stationary gripper coil voltage to dashpot entry with:

Note: A proposed change to 3.1.3.3 is

a. Tavg Ž 541'F, and pending via PSEG LAR S09-01, submitted

b. All reactor coolant pumps operating. March 22, 2009.

APPLICABILITY: MODE 1 & 2.

ACTION:

a. With the drop time of any full length rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.

b. With the red drop times wit-hin limits but dotormffinod with 3 r...t.r eoolant pum..ps p*.rating, ep.r.ation may preeeod provided THERDIý POWER is rostrit td to 7 of DPTED THERmDIA POWER.

SURVEILLANCE REQUIREMENTS 4.1.3.3 The rod drop time of full length rods shall be demonstrated through measurement prior to reactor criticality:

a. For all rods following each removal of the reactor vessel head,

b. For specifically affected individual rods following any maintenance on or modification to the control rod drive system which could affect the drop time of those specific rods, and

c. At least once per 18 months.

SALEM - UNIT 1 3/4 1-21 Amendment No. 2-1-k

REACTIVITY CONTROL SYSTEMS SHUTDOWN ROD INSERTION LIMIT LIMITING CONDITION FOR OPERATION 3.1.3.4 All shutdown rods shall be FULLY WITHDRAWN.

APPLICABILITY: MODES 1*, and 2*#@

ACTION:

With a maximum of one shutdown rod not FULLY WITHDRAWN, except for surveillance testing pursuant to Specification 4.1.3.1.2, within one hour either:

a. FULLY WITHDRAW the rod, or,

b. Declare the rod to be inoperable and apply Specification 3.1.3.1.

SURVEILLANCE REQUIREMENTS 4.1.3.4 Each shutdown rod shall be determined to be FULLY WITHDRAWN by use of the group demand counters, and verified by the analog rod position indicators**,*-**:

a. Within 15 minutes prior to withdrawal of any rods in control banks A, B, C, or D during an approach to reactor criticality, and

b. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

See Special Test Exceptions 3.10.2 and 3.10.3

- For power levels below 50% one hour thermal "soak time" is permitted.

During this soak time, the absolute value of rod motion is limited to six steps.

During Cyele 14, the position of Red !SB2 will be determined indireetly by the movablo macero deteetor~s within 8 houris following its movementz until the repair of the indieation system for this roed. DPuwring reactor startup, the fully withdrawn position of Red !GB2 will be determined by udr-rent traces and subsequently verified by the moevable ineer-e deteetors p-rior to entry into Mode I.

With Keff greater than or equal to 1.0 Surveillance 4.1.3.4.a is applicable prior to withdrawing control banks in preparation for startup (Mode 2).

SALEM - UNIT 1 3/4 1-22 Amendment No. 2-30

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) b) At least once per 31 EFPD, whichever occurs first.

2. When the Fxy C is less than or equal to the FxyRTP y limit for the appropriate measured core plane, additional core power distribution measurements shall be taken and FxyC compared to FxyRTP ,y and FxyL at least once per 31 EFPD.

e. The Fxy limit for Rated Thermal Power (FyRTP ,y) shall be provided for all core planes containing bank "D" control rods and all unrodded core planes in the COLR per specification 6.9.1.9.

f. The Fxy limits of e, above, are not applicable in the following core plane regions as measured in percent of core height from the bottom of the fuel:

1. Lower core region from 0 to 15% inclusive.

2. Upper core region from 85 to 100% inclusive.

3. Grid plane regions at 17.8 + 2%, 32.1 +/-2%, 46.4 +/-2%, 60.6 +/-2%,

and 74.9 +2% inclusive.

4. Core plane regions within +2% of core height (+/-2.88 inches) about the bank demand position of the bank "D" control rods.

g. Evaluating the effects of Fxy on FQ(Z) to determine if FQ(Z) is within its limit whenever Fxyc y exceeds FxyL I SALEM - UNIT 1 3/4 2-7 Amendment No. 2

TABLE 3.2-1 DNB PARAMETERS PARAMETER LIMITS 4 Loops In Operation 0

Reactor Coolant System Tavg 582.9 F Pressurizer Pressure 2 2200 psia*

Reactor Coolant System Flow Ž 341,000 gpm#

Limit not applicable during either THERMAL POWER ramp inerease in excess of 5% RATED THERMAL POWER per minute or a THERMAL POWER step inerease in excess of 10% RATED THERMAL POWER. I Includes a 2.4% flow measurement uncertainty plus a 0.1% measurement uncertainty due to feedwater venturi fouling.

SALEM - UNIT 1 3/4 2-14 Amendment No. 201

TABLE 4.3-1 (Continued)

NOTATION With the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal.

(1) - If not performed in previous 31 days.

(2) - Heat balance only, above 15% of RATED THERMAL POWER.

(3) - Compare incore to excore axial offset above 15% of RATED THERMAL POWER. Recalibrate if absolute difference Ž 3 percent.

(4) - Manual SSPS functional input check every 18 months. -*-

(5) - Each train or logic channel shall be tested at least every 62 days on a STAGGERED TEST BASIS.

(6) - Neutron detectors may be excluded from CHANNEL CALIBRATION.

(7) - Below P-6 (Block of Source Range Reactor Trip) setpoint.

(8) - Deleted (9) - The CHANNEL FUNCTIONAL TEST shall independently verify the OPERABILITY of the Undervoltage and Shunt Trip mechanism for the Manual Reactor Trip Function.

The Test shall also verify OPERABILITY of the Bypass Breaker Trip circuits.

(10) - DELETED (11) - The CHANNEL FUNCTIONAL TEST shall independently verify the OPERABILITY of the Reactor Trip Breaker Undervoltage and Shunt Trip mechanisms.

(12) - DELETED

- A ene tioer xtons.in t . this urv.illanh e requirefment whi.h is satisfied by perfermoanee ef the Manual S1 test is granted during fuel eyele thirteen allowing,-

Unit 1 eperatiens to continue to the thirteenth refueling outage (1R13) . The-surveillance testing is te be completed at the appropriate time duaring the !R!3 eutage, prier to the unit returning to Mode 41 upon outage completion.

SALEM - UNIT I 3/4 31 Amendment No. 222

TABLE 4.3-2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCT I ONAL SURVEILLANCE FUNCTIONAL UNIT

1. SAFETY INJECTION, CHECK TURBINE TRIP AND FEEDWATER CALIBRATION ISOLATION 0/

TEST it-k REQUIRED

a. Manual Initiation N.A. N.A. 1,2,3,4

b. Automatic Actuation Logic N.A. N.A. M(2) 1,2,3,4

c. Containment Pressure-High S R Q(3) 1,2,3

d. Pressurizer Pressure--Low S R Q 1,2,3

e. Differential Pressure Between S R Q 1,2,3 Steam Lines--High

f. Steam Flow in Two Steam S R Q 1,2,3 Lines--High coincident with Tavg--Low-Low or Steam Line Pressure-Low

2. CONTAINMENT SPRAY

a. Manual Initiation N.A. N.A. R 1,2,3,4

b. Automatic Actuation Logic N.A. N.A. M(2) 1,2,3,4

c. Containment Pressure--High-High S R Q(3) 1,2,3

A ono timot e2Etenslen to this sur-voillanoo rogeuirefmont whioh is satisfiodl by peifr-foranoo of the Manudal:

6I test is granted duaring fuel eyolo thirteen allowing Unit 1 operatiens to continue to the thirteenth roefuoling outago (IR13) . The survoillanoo testing is to be eomplotod at thie appropriato timoi duringf the !R13 outago, prior to the unit returning to Hodo 4 upon outage eompletion.

SALEM - UNIT 1 3/4 3-31a Amendment No. 2-2 TABLE 3.3-6 (Continued)

RADIATION MONITORING INSTRUMENTATION MINIMUM CHANNELS APPLICABLE ALARM/TRIP MEASUREMENT INSTRUMENT OPERABLE MODES SETPOINT RANGE ACTION

2. PROCESS MONITORS

b. Noble Gas Effluent Monitors 0/

3. 0xl0-2pCi/cm3 1

-1 0-IOCi /cm 3

0/

1) Medium Range Auxiliary 1 1,2, 3&4 23 Building Exhaust System (Alarm only)

(Plant Vent) 3 5

2) High Range Auxiliary 1 1,2, 3&4 *I. OxlO02Ci/cm 10l0 pCi*/cm3 23 Building Exhaust System (Alarm only)

(Plant Vent)

3) Condenser Exhaust 1 1,2, 3&4 l. 27x10 4 cpm 1-106 cpm 23 System (Alarm only)

3. CONTROL ROOM

a. Air Intake - 2/Intake##

/0

2.48x103 cpm 101_107 cpm 24, 25 Radiation Level

1. Control Room air intakes shared between Unit 1 and 2.

- ALL MODES and during movement of irradiated fuel assemblies and during CORE ALTERATIONS.

SALEM - UNIT 1 3/4 3-36a Amendment No. 249

TABLE 4.3-3 (Continued)

RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES IN WHICH CHANNELS SOURCE CHANNEL FUNCTIONAL SURVEILLANCE INSTRUMENT CHECKS CHECKS, CALIBRATION TEST REQUIRED

2. PROCESS MONITORS

b. Noble Gas Effluent Monitors

1) Medium Range Auxiliary S M R Q 1, 2, 3 & 4 Building Exhaust System (Plant Vent)

2) High Range Auxiliary M R Q 1, 2, 3 & 4 Building Exhaust System (Plant Vent)

3) Condenser Exh. Sys. S M R Q 1, 2, 3 & 4

3. CONTROL ROOM

a. Air Intake - Radiation Level S M R Q

- ALL MODES and during movement of irradiated fuel assemblies and during CORE ALTERATIONS.

SALEM - UNIT 1 1 4 3-38a Amendment No. 2-&G I

TABLE 4.3-11 (Continued)

SURVEIL]LANCE REQUIREMENTS FOR ACCIDENT M(ONITORING INSTRUMENTATION CHANNEL CHANNEL CHANNEL FUNCTIONAL INSTRUMENT CHECK C ALIBRATION TEST

12. PORV Position Indicator M N.A. R

13. PORV Block Valve Position Indicator M N.A. Q*

14. Pressurizer Safety Valve Position M N.A. R Indicator

15. Containment Pressure - Narrow Range M N.A. N.A.

16. Containment Pressure - Wide Range M R N.A.

17. Containment Water Level - Wide Range M R** N.A.

18. Core Exit Thermocouples M R N.A.

19. Reactor Vessel Level Instrumentation M R N.A.

System (RVLIS)

20. Containment High Range Accident Radiation S R Q Monitor

Unless the block valve is closed in order to meet the requirements of Action b, or c in specification 3.4.3.

ei ccci cimsnf~ cr;oonfsioni cc- this surveillance requiremaent is grantedl dluring fuel eyole thirteen allowing Unit 1 e.p.rti.ns to duntinuo to the thirteenth rcefudeling outage (1R13) . The sur~veill anee is to be eompleted at the appreopriato tiffe during the 1R13 outage, prier to the unit r-eturiaifng to Modle 4 upon oudt-age eompletion.

SALEM - UNIT 1 3/4 3-57a Amendment No. 2-7 INSTRUMENTATION POWER DISTRIBUTION MONITORING SYSTEM LIMITING CONDITION FOR OPERATION (Continued)

APPLICABILTY: MODE 1, above 25% RATED THERMAL POWER (RTP)

ACTION:

With any of the operability criteria listed in 3.3.3.14.a, 3.3.3.14.b, or 3.3.3.14.c not met, either correct the deficient operability condition, or declare the PDMS inoperable and use the incore movable detector systemT satisfying the OPERABIL.ITY requirements listed in Gpeeifieatien 3.3.3.2, to obtain any required core power distribution measurements. Increase the measured core peaking factors using the values listed in the COLR for the PDMS inoperable condition.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.14.1 The operability criteria listed in 3.3.3'14.a, 3.3.3.14.b, and 3.3. 3. . 14. c shall be verified to be satisfied prior to acceptance of the PDMS core power distribution measurement results.

4.3.3.14.2 Calibration of the PDMS is required:

a. At least once every 180 Effective Full Power Days when the minimum number and core coverage criteria as defined in 3.3.3.14.b.1 and 3.3.3.14.b.2 are satisfied, or

b. At least once every 31 Effective Full Power Days when only the minimum number criterion as defined in 3.3.3.14.b.3 is satisfied.

SALEM - UNIT 1 3/4 3-71 Amendment No. 2-74

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

b. At least once per 31 days and within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after each solution volume increase of Ž 1% of tank volume by verifying the boron concentration of the accumulator solution.

c At least once per 31 days when the RCS pressure is greater than 1000 psig by verifying that the power lockout switch is in lockout.

d.-* At least once per 18 months by verifying that each accumulator isolation valve opens automatically upon receipt of a safety injection test signal.

A no time eixtension to tnis survell~anee roguirement.whieh is satisfied by performanee of the Manual S! test is gr~nted durn fuoel eyelo thirteen allowing Unit 1 oporations to eontinue to the thirteenth refueling eutago (IR13) . The surveillanee testing is to bo eemplotod at the apprepriate timoe during the !R13 eutago, prior to the unit returning to Modo 4 upon outago eomplotion.

SALEM - UNIT I 3/4 5-2 Amendment No. 222

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

c. By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the containment sump and cause restriction of the pump suctions during LOCA conditions. This visual inspection shall be performed:

1. For all accessible areas of the containment prior to establishing CONTAINMENT INTEGRITY, and

2. At least once daily (24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months consecutive period) the areas affected within containment by containment entry and during the final entry when CONTAINMENT INTEGRITY is established.

d. At least once per 18 months by:

ý1. A visual inspection of the containment sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or corrosion.

e. \ least once per 18 months, during shutdown, by:

l.+/- Verifying that each automatic valve in the flow path actuates to its correct position on a safety injection test signal.

2. Verifying that each of the following pumps start automatically upon receipt of a safety injection test signal:

a) Centrifugal charging pump b) Safety injection pump c) Residual heat removal pump N

satisfied by perferm~anec ef the Manual SI test is griant~ed duraing fudel cycle thirteen allewing Unimt I operations to continuie to the eompete attheapproprit tiedrn he 1R13 outage, parioer te th~e uni-t returning to Mode 4 upon outagc comfpletioen.

SALEM - UNIT 1 3/4 5-5" Amendment No. 2-2-2

PLANT SYSTEMS SURVEILLAN REQUIREMENTS (Continued) 2.- Verifying that on a safety injection test intake high radiation test signal or control room signal, the system automatically I actuates in the pressurization mode by opening the outside air supply and diverting air flow through the HEPA filter and charcoal adsorber bank.

3. Deleted

4. Verifying that on a manual actuation signal, the system will actuate to the required pressurization or recirculation operating mode.

5. Verify each CREACS train has the capability to remove the assumed heat load.

e. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove Ž 99% of the DOP when they are tested in-place while operating the filter system at a flow rate of 8000 cfm +/- 10%.

f. After each complete or partial replacement of a charcoal absorber bank by verifying that the charcoal absorbers remove Ž 99% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place while operating the filter system at a flow rate of 8000 cfm +/- 10%.

4.7.6.2 Perform required CRE unfiltered air inleakage testing in accordance with the Control Room Envelope Habitability Program (Refer to T.S.

6.18).

A one time oectensien te this surveillanee ir i E'E IirmnfH - ;i nlF 1 F WRi 5 atisfied by perfor-manco of the Manual S! tesot is t:1= Ct .nted during fuoel eyelo LEirecon awoewing unwo + operatwons eo continuo te nc oo irooentn roruoling

.utoag,1R13) . The .ur.villln.o testing is to beo rmplo*td at the appreopu-arit timot duiringj the 1R13 outago, prior to the unit returning to Modo 4 upon eutoge.

ompilolt ien.

K SALEM - UNIT 1 3/4 7-21 Amendment No. 286

LIMITING CONDITION FOR OPERATION ACTION: MODES 5 and 6 or during movement of irradiated fuel assemblies. *

a. With one chiller inoperable:

1. Remove the appropriate non-essential heat loads from the chilled water system within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and;

2. Restore the chiller to OPERABLE status within 14 days or;

3. Suspend CORE ALTERATIONS and movement of irradiated fuel assemblies.

b. With two chillers inoperable:

1. Remove the appropriate non-essential heat loads from the chilled water system within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and;

2. Align the control room emergency air conditioning system (CREACs) for single filtration operation using the Salem Unit 2 train within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and;

3. Restore at least one chiller to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or;

4. Suspend CORE ALTERATIONS and movement of irradiated fuel assemblies.

c. With one chilled water pump inoperable, restore the chilled water pump to OPERABLE status within 7 days or suspend CORE ALTERATIONS and movement of irradiated fuel assemblies.

SURVEILLANCE REQUIREMENTS 4.7.10 The chilled water loop which services the safety-related loads in the Auxiliary Building shall be demonstrated OPERABLE:

a. At least once per 31 days by verifying that each manual valve in the chilled water system flow path servicing safety related components that is not locked, sealed, or otherwise secured in position, is in its correct position.

b. X-*At least once per 18 months, by verifying that each automatic valve actuates to its correct position on a Safeguards Initiation signal.

c. At least once per 92 days by verifying that each chiller starts and runs.

During Modes 5 and 6 and during movement of irradiated fuel assemblies, chilled water components are not considered to be inoperable solely on the basis that the backup emergency power source, diesel generator, is inoperable.

'*÷A one time eixtension to this survei!!ancc requirement for performancc of relay time responsc and se.u.ne. testing f the safe.guard equipment e.ntr.l (t.C..

systom, whi*h partially satio*fies the.uc...illane. requirement, is granted during fu, ,

cyele thirteen allewing Unit 1 operaltons t0 C©*onnue te 0Er LrllreCenCli rcrul+/-In9 eutage (IR13) . The survoFillaneo testing il; tR be completed at the apprepriate time during the . 1R1. utage, prier to the unit returning t. Med. 4 upon outage complct ion.

SALEM - UNIT 1 3/4 7-34 Amendment No. 22

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

3. The connection resistance is:

< 150 micro ohms for inter-cell connections,

< 350 micro ohms for inter-rack connections,

< 350 micro ohms for inter-tier connections,

70 micro ohms for field cable terminal connections, and

<2500 micro ohms for the total battery connection resistance which includes all inter-cell connections (including bus bars), all inter-rack connections (including cable resistance), all inter-tier connections (including cable resistance), and all field terminal connections at the battery.

\ e. At least once per 18 months by verifying that the battery charger will supply at least 170 amperes at 125 volts for at least 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

f.- At least once per 18 months, during shutdown, by verifying that the battery capacity is adequate to supply and maintain in OPERABLE status all of the actual or simulated emergency loads for the design duty cycle when the battery is subjected to a battery service test.

g. At least once per 60 months, during shutdown, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test.

Satisfactory completion of this performance discharge test shall also satisfy the requirements of Specification 4.8.2.3.2.f if the performance discharge test is conducted during a shutdown where that test and the battery service test would both be required.

h. At least once per 12 months, during shutdown, if the battery shows signs of degradation OR has reached 85% of the service life with a capacity less than 100% of manufacturers rating, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test. Degradation is indicated when the battery capacity drops more than 10% of rated capacity from its capacity on the previous performance test, or is below 90% of the manufacturer's rating.

i. At least once per 24 months, during shutdown, if the battery has reached 85% of the service life with capacity greater than or equal to 100% of manufacturers rating, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test.

+A one time extension to this surveillance requirement is granted-during fuel eycle thirteen allowing Unit I operations to continude t the thirteenth refueling outage (!1113). The surveillance is to be e eempleed at the apprepriate cine Ruring tfne 1R1i oucage, prior co tno unit returning to Mode 4 upon outage completion.

SALEM - UNIT 1 3/4 8-9a Amendment No. 222

ELECTRICAL POWER SYSTEMS 3/4 8.3 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES LIMITING CONDITION FOR OPERATION 3.8.3.1 All containment penetration conductor overcurrent protective devices required to provide thermal protection of penetrations shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With one or more of the required containment penetration conductor overcurrent protective device(s) inoperable:

a. Restore the protective device(s) to OPERABLE status or de-energize the circuit(s) by tripping either the primary or backup protective device, or racking out or removing the primary or backup device within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , declare the affected system or component inoperable, and verify the primary or backup protective device to be tripped, or the primary or backup device racked out or removed at least once per 7 days thereafter; or

b. Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.8.3.1 All required containment penetration conductor overcurrent protective devices shall be demonstrated OPERABLE:

a. At least once per 18 months:

I.*,-*For at least one 4.16 KV reactor coolant pump circuit, such that all reactor coolant pump circuits are demonstrated OPERABLE at least once per 72 months, by performance of:

(a) A CHANNEL CALIBRATION of the associated protective relays, and (b) An integrated system functional test which includes simulated automatic actuation of the system and verifying that each relay and associated circuit breakers and control circuits function as designed.

A one time .. t.n.i.

n te thi* urv7illaneo rguirom..nt for inspection calibration and mtggaring .f the 1F 4KV Euc eve3ad relays, which partially satisfies this siurveillanee requirement, is granted duiring fuel eyele thirteen allowing unit1 eporatiens to continua to the thirteenth refueling eutage (IR13) . The saurvoýillanco testing is to be eomplotod at the apprepriate time during the !R1-3 eutage, prier to the unit returning to Modo 4 u:pon eut-age eompletion.

- A ono time oxEtension to this surveillanco reguirement for inspoction calibration and moeggaring of the !A, 1B, and iC 460 tranfrmor relays and C.T.'., whi. h partially, satisfy this sourveillanee roguairemont, is grantod during2 fuel eyolo thirt..n allowing .p.ration.

Unit 1 t teo ntinus to tho thirteenth refuoling eutage (1R13) . The surveillanco tosting is to be complotod at the appropriate time during the !R13 outage, prior to the unit returning to Mede 4 udpon eutage eomplot ionB.

SALEM - UNIT I 3/4 814 Amendment No. 2-22 LAR S09-03 LR-N09-0164 UNIT I FACILITY OPERATING LICENSE PAGES WITH PROPOSED CHANGES Facility Operating License DPR-70

- 5c-

10. TERMINATION Pursuant to the provisions of 10 CFR 75.41, the Commission will inform the licensee, in writing, when its installation is no longer subject to Article 39(b) of the principal text of the US/IAEA Safeguards Agreement. The IAEA Safeguards License Conditions incorporating Code 7. of the Facility Attachment as part of NRC License DPR-70 will be terminated as of the date of such notice from the Commission. However, since the IAEA may elect to maintain the licensee's installation under Article 2(a) of the Protocol, provisions equivalent to Codes 1. through 6. of the Facility Attachment (with possible appropriate modifications) may still apply, and accordingly all other IAEA Safeguards License Conditions to NRC License No. DPR-70 will remain in effect until the Commission notifies the licensee otherwise. If this option is not selected by the IAEA, the Commission will then notify the licensee that all License Conditions pertaining to the US/IAEA Safeguards Agreement are terminated.

J. RELOCATED TECHNICAL SPECIFICATIONS PSEG Nuclear LLC shall relocate certain technical specification requirements to licensee-controlled documents as described below. The location of these requirements shall be retained by the licensee.

a. This license condition approves the relocation of certain technical specification requirements to licensee-controlled documents (UFSAR), as described in the licensee's applications with the staff's safety evaluation approval and Amendment No. as noted below:

Licensee's Applications Safety Evaluations Amendment Nos.

September 25, 1996 January 30, 1997 189 Implementation shall include the relocation of technical specifications requirements to the appropriate licensee-controlled document as identified in the licensee's application.

FOR THE NUCLEAR REGULATORY COMMISSION Original Signed by Roger S. Boyd Roger S. Boyd, Director Division of Project Management Office of Nuclear Reactor Regulation Attachments:

1. lneecrnplete Preeperatienal Tests, Startup Tests,. and th.. R. DELETED r t..

Which Must Be Ccrnipleted

2. Page Changes to Technical Specifications, Appendix A Date of Issuance: December 1, 1976 Amendment No. 24

ATTACHMENT 1 LICENSE DPRT70 This attachmnent identifies certain preoperational tests, startup tests, and Other- item which must be completed to the Commission's satisfaction prior-to proceeding to certain specified Oper-ational Modes. Public Ser.'ice Electr-ic and Gas Company shall not proceeed beyond the author-ized Operational Modes without pr-ior written authorizato from the Commission-.

A. Public Sen'ice Electr-ic and Gas Company may at the license issue date proee directly to Operational Mode 6 (initial fuel loading), and may subsequently proeee-to Oper-ational Mode 5 (cold shuttdown).

B. Prior-to proeeeding to Operational Mode 4 (hot shutdown); Public sen'vice Electric coola-nt system per SUP 20. 1. Subsequent to the ver-ification by the Of-ficeo Inspection and Enforceement of the acceptable completion of this item, and u~pon wr-itten authorization by the Commission, Publie sen'Lie Electr-ic and Gas company may proceed to Operational Mode 4 (hot shutdown),

C. Priorf to proceeding to Operational Mode 3 (hot standby), Puiblic Sen'ice Electfic and Gas Company shall complete the followxing itemns:

I ~ r-flTT-rb . . . . 11--T' -i -1 1 ,1T'%TT,%f, I. Testing epefmiiun ufl~is i-1Ipufiw fcifouuiatii valves11x, ii-r~.y - n z-uz ver 1

SUP. 50.0.

2. Testing motor-winding tempi er-atur-es ofR1R inuma motor-s Nos. 11 and 12 per-SUP 12.

3. Testing the following snubbers per-SUP 50.41!

MMT- 11 29A RHRH-l1-29B RI-RH 12 3A2 R1R: 12 34-C

4. Testing the boron reycle system per-SUP 10.5,

5. Demonstrate beta dosimet, y capability.

6. Testing process radiation monitors, excluding those r-equtired for-fuel loading, per-SUP 214.

7. Testing service water system per SUP 28.

(Revised September 10, 197-6)

8.-Testing E7hilled water porion of the control F8 ,.it air-conditioning system pe i

!ýl p IQ 0

t)

J. Prepare the following radioehemistry pr-ocedur-es:

(a) PD3.3.010 pro.edur.e to determin.e the average ener.gy f gamm.a emitting is*t*pe;,,

(b) PD 3.3.011 procedurye for-detecting fission gases by gamma spectroscepy in the pr~esence of other gases; (e) PD 3.3.003 pr-oceduire to determine the dose equivalent iodine 131 in the Fyeee1ant.

10. Replace the existing standby charcoal filters in the aux~iliar-y building ventilation system with charcoeal filters capable of r-enoving 90 percent of the erganic iodiiies.

Subsequent to ver-ificoation by the Office of Inspection and Enforceement of th acceptable eempletion of the above listed items, and upon written authorization trom the commission, me Public benwie Eleetne and U~as Company may proeed t Operational Mode 3 (hot standby).

D. Pro opoeding to Operational Mode 2 (initial cditiealityý, Public Sen'ic

1. Testing high temper-atuire alarm TE463A on.prsuie elief line per- SUP 50.6.

2. Testing control of steamn gener~ator- blowdo~m flow by valves GB8 and GBIO pel SUP 5043.

3. Testing up*per mtor bearing of feaeto+-coolant pump No. 14 per SUP 50.0.

A ~i ~ CiT Wfl Cfl a T'

IQQ4-'IT~

= ~ Lrrr :g-* I '4re-jr;Ior

ýHA PAC44AA4I1 niR-4Ri 1" .. I XT 4ýcA I I --

4 4 n:-Ar AI'-- mW W

5. Testing RID's Nos. 423B, 431A, 433B and 44.B in the reactor coo.lant system pef SUP 50,7,

6. Testing the following snubbers per-SUP 50.4:

(Revised September 10, 1976)

- 1 DPRA 1.46 I -P-RSN 7 1 PRSN 28 -l-PRSN 400 1 PRA 150 1 DPRSN 9 1 PRSN 29 1 PRSN 40t-1 PPRA 54 i-RRSN io 1 DPRN 30 I PPRSN 402 1 PR A 159Q i-PRSN-1P ! PRSN 32A 1 PRSN 405 1 PRA 162 1 DPDR N2 I PRSN 32B 1-PRSN 405A 1 DPRN 13 1 DPRSNT 3 1 PRSN 406 1 PRSN41 1 PRSN 16 1 PRSN 34 1 PRSN 406A SPRSNT 2 1 PRSN-7 1 PRS'N 3 SDDRRSNT 119 1 DDRSNI 37" i 4 .PRSN-3A .4 1 ,PRS.2 1-PRSN 39 4 PRSN 4 1 PRSN 23 1-PRSN 5 I PRSN 25 1 PRSN 5A 1 PRSN 27 1 PRSN 42 Subsequent to verificationf by the Office of Inspeetion and Enforcement of the acceptable completion of the above items, and upo rItte authorization fro~m the Commission, public Se..wi-e Electric -AndG-as- Company may proceed to Oper-ational Mode 2 (initial cri4ticality).

. P*o t, o Aing t. Operation Mode 1 power-operation), the following item shall. be. fompleted:

1. Reactor-Vessel Ove~r--e nur-e Alafm A reacter- vessel ox emr-p~,r-e e alafm shall be installed in the control room. This alaf shall be operable whenever- the system is in cold shutdown or-hot shutdown, shall be actuiated whenever the system pr-essure exceeds the techaical specification limits, and shall not compomia safety related equipmnent.

2. Mainteniance proceedures The maintenance procedur-es delineated in inspection and Enforceement Repei4 50 272,/76 38 shall be completed.

Subsequent to ver-ification by the Office of inspection and Enforcement of the acceptable completion of the above items, and utpon written authorization by the Commffission,.Public ser-vice Electric and Gas Company may proceed in its poe ascnson program to Oper-ational Mode 1, with the power-level limited to twenty per-cent of r-atedl co~re power.

(Revised .December.1, 1976)

F.Prior-to exceeding the forty per-eent powAer- limit, the snubber tests delineated in Rtem F above shall be repeated at a poerve level between thiity and forty, perceent of rated-core poer~e. Upon written acceptance by the Commission of the above. items, Pubhie Ser.'ioe Electric and Gas Company may proeeed in its power-asenio rcam to a power level not exceeding ninety perceent of rated cor-e power.

G. Prior-to exceeding the ninety perceent power-limit, the snubber-tests delineatedi Rtem F above shall be r-epeated at a power-level betwee~n eighty and ninety perceent of rated. Upon wr-itten acceptance by the Commission of these tests, Public S e'icee Electr-ic anad Gas Company may proceeed in its poe aseson pioroamn to fuill Power, Uponef attaining full power-, or-as soon as possible thereafter-, Public Sen'viee Eleti and Gas Company shall per-fonm a final ver-ification test of these snubbers. The Office of Inspeetion and Enforceement will review the results of these vefification tests, and absent any niotification to the contrar-y, Public Ser-vice Electric and Gias Companay may sustainl full power oper-ation.

(Revised December 1, 197-6)

LAR S09-03 LR-N09-0164 UNIT 2 TECHNICAL SPECIFICATION PAGES WITH PROPOSED CHANGES Facility Operating License DPR-75

REACTIVITY CONTROL SYSTEMS ROD DROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.3 The individual full length (shutdown and control) rod drop time from 228 steps withdrawn shall be less than or equal to 2.7 seconds from beginning of decay of stationary gripper coil voltage to dashpot entry-with:

a. Tavg greater than or equal to 541'F, and Note: A proposed change to 3.1.3.3 is

b. All reactor coolant pumps operating. pending via PSEG LAR S09-01, submitted APPLICABILITY: MODES 1 & 2.

March 22, 2009.

ACTION:

a. With the drop time of any full length rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.

I

-1 -1 ryý Ký 3 1 Tylr ý i ý

ýj ý ziýzi

'. xxiý

- I r...t.. el.. .p.r.tin.g, t piumps eperation may procood provided T4ERDAL POWER is rotr-iotd to less than or equal to 76% of RATED THERMAL POWER.

SURVEILLANCE REQUIREMENTS 4.1.3.3 The rod drop time of full length rods shall be demonstrated through measurement prior to reactor criticality:

a. For all rods following each removal of the reactor vessel head,

b. For specifically affected individual rods following any maintenance on or modification to the control rod drive system which could affect the drop time of those specific rods, and

c. At least once per 18 months.

SALEM - UNIT 2 3/4 1-18 Amendment No. q-4

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE (AFD)

LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFE CE shall be maintained within the target band about the t flux di rence as specified in the CORE OPERATING LIMITS REPO P-G) (COLR I APPLICABILITY: MODE 1 ABOVE 50% RATED THERMAL POWER*

ACTION:

a. With the indicated AXIAL FLUX DIFFERENCE outside of the target band about the target flux difference as specified in the COLR and with THERMAL POWER:

1. Above 90% of RATED THERMAL POWER, within 15 minutes:

a) Either restore the indicated AFD to within the target band limits, or b) Reduce THERMAL POWER to less than 90% of RATED THERMAL POWER.

2. Between 50% and 90% of RATED THERMAL POWER:

a) POWER OPERATION may continue provided:

1) The indicated AFD has not been outside of the target band as specified in the COLR for more than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months penalty deviation cumulative during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and

2) The indicated AFD is within the limits as specified in the COLR. Otherwise, reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b) Surveillance testing of the Power Range Neutron Flux Channels may be performed pursuant to Specification 4.3.1.1.1 provided the indicated AFD is maintained within the limits as specified in the COLR. A total of 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months operation may be accumulated with the AFD outside of the target band during this testing without penalty deviation.

See Special Test Exception 3.10.2 SALEM - UNIT 2 3/4 2-1 Amendment No. 7

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)

2. When the FxyC Y is less than or equal to the Fxy RTP XZ limit for the appropriate measured core plane, additional core power distribution measurements shall be taken and FxyC compared to Fxy RTP y and FxyL at least once per 31 EFPD.

e. The Fxy limit for Rated Thermal Power (Fxy RTP) shall be provided for all core planes containing bank "D" control rods and all unrodded core planes in the COLR per specification 6.9.1.9.

f. The Fxy limits of e., above, are not applicable in the following core plane regions as measured in percent of core height from the bottom of the fuel:

1. Lower core region from 0% to 15%, inclusive.

2. Upper core region from 85% to 100%, inclusive.

3. Grid plane regions at 17.8% +/- 2%, 32.1% +/- 2%, 46.4% +/- 2%,

60.6% +/- 2% and 74.9% +/- 2%, inclusive.

4. Core plane regions within +/- 2% of core height (+/- 2.88 inches) about the bank demand position of the bank "D" control rods.

g. Evaluating the effects of Fxy on FQ(Z) to determine if l

FQ(Z) is within its limit whenever Fxyc-y exceeds FxyL

4.2.2.3 When FQ(Z) is measured pursuant to specification 4.10.2.2, an overall measured FQ(Z) shall be obtained from a core power distribution measurement I

and increased by the applicable manufacturing and measurement uncertainties-as specified in the COLR.

SFr C!yle 11, wh - . , - blo movablo dotootor thimbles is grcarcr wnRan or oqual 00 5U6 an0 1es0s Ehan 75% of the total, the 5%

mooouý-romoent uneer-tainty shall be inerea sod to [5% ! (3 T/14 5) (1%)] who Fc T lo #hp nub R9o vj~intin~

SALEM - UNIT 2 3/4 2-7 Amendment No. 2-1-9

RADIATION MONITORING INSTRUMENTATION MINIMUM CHANNELS APPLICABLE ALARM/TRIP MEASUREMENT INSTRUMENT OPERABLE MODES SETPOINT RANGE ACTION

2. PROCESS MONITORS

b. Noble Gas Effluent Monitors

1) Medium Range Auxiliary 1 1,2, 3&4 3. Oxl0-2pCi/cm 3 /

10-1-101 pCi/cm 3

26 Building Exhaust System (Alarm only)

(Plant Vent)

2) High Range Auxiliary 1. Oxl02PCi/cm 3 10-1-105 ýLCi/CM3 1 1,2, 3&4 26 Building Exhaust System (Alarm only)

(Plant Vent) 0/

3) Condenser Exhaust 1 1,2, 3&4 *7.12x10 4 cpm 1_106 cpm 26 System (Alarm only)

3. CONTROL ROOM

a. Air Intake - 2/Intake## **

0/

2.48x10 3 cpm 101_107 cpm 27,28 Radiation Level

1. Control Room air intakes shared between Unit 1 and 2.

- ALL MODES and during movement of irradiated fuel assemblies and during CORE ALTERATIONS.

SALEM - UNIT 2 3/4 3=39a Amendment No. 243

POWER DISTRIBUTION MONITORING SYSTEM LIMITING CONDITION FOR OPERATION (Continued)

APPLICABILTY.- MODE 1, above 25% RATED THERMAL POWER (RTP)

ACTION:

With any of the operability criteria listed in 3.3.3.14.a, 3.3.3.14.b, or 3.3.3.14.c not met, either correct the deficient operability condition, or declare the PDMS inoperable and use the incore movable detector system-Fi satisfying the OPERABILITY requirements listed in Specification 3.3.3.2, to obtain any required core power distribution measurements. Increase the measured core peaking factors using the values listed in the COLR for the PDMS inoperable condition.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.14.1 The operability criteria listed in 3.3.3.14.a, 3.3.3.14.b, and 3.3.3.14.c shall be verified to be satisfied prior to acceptance of the PDMS core power distribution measurement results.

4.3.3.14.2 Calibration of the PDMS is required:

a. At least once every 180 Effective Full Power Days when the minimum number and core coverage criteria as defined in 3.3.3.14.b.1 and 3.3.3.14.b.2 are satisfied, or

b. At least once every 31 Effective Full Power Days when only the minimum number criterion as defined in 3.3.3.14.b.3 is satisfied.

Salem - Unit 2 3/4 3-66 Amendment No. 2-5-&

REACTOR COOLANT SYSTEM 3.4.11 STRUCTURAL INTEGRITY ASME CODE CLASS 1, 2 and 3 COMPONENTS LIMITING CONDITION FOR OPERATION 3.4.11.1 The structural integrity of ASME Code Class 1, 2 and 3 components shall be maintained in accordance with Specification 4.4.11.1.

APPLICABILITY: ALL MODES.

ACTION:

a. With the structural integrity of any ASME Code Class 1 component(s) not conforming to the above requirements, restore the structural integrity of the affected component(s) to within its limit or isolate the affected component(s) prior to increasing the Reactor Coolant System temperature more than 50'F above the minimum temperature required by NDT considerations.

b. With the structural integrity of any ASME Code Class 2 component(s) not conforming to the above requirements, restore the structural integrity of the affected component(s) to within its limit or isolate the affected component(s) prior to increasing the Reactor Coolant System temperature above 200 0 F.

c. With the structural integrity of any ASME Code Class 3 component(s) not conforming to the above requirements, restore the structural integrity of the affected component(s) to within its limit or isolate the affected component(s) from service.

SURVEILLANCE REQUIREMENTS 4.4.11.1 In addition to the requirements of Specification 4.0.5, each Reactor Coolant Pump flywheel shall be inspected per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1, August 1975. In lieu of Position C.4.b(l) and C.4.b(2), a qualified in-place UT examination over the volume from the inner bore of the flywheel to the circle one-half of the outer radius or a surface examination (MT and/or PT) of exposed surfaces of the removed'flywheels may be conducted at 20 year intervals.

4.*.11.2 Augmented inserviee inspeetien-Program fer Steam C.n.rater Channel Heads The Ne. 21 Steam Conerater ehannel head shall be ultrasonically inspected in a selIeted area during each ef the first three refueling eutages using the same ultrasonic inispoction procoduroc anad equipment used to generate the baseline data.

These inservico ultrasonic inspections shall verify that the cracks obsorved in the stainless steel cladding prier to eperatien have net propagated into the base material.

SALEM - UNIT 2 3/4 4-33 Amendment No. 2-58&

ADMINISTRATIVE CONTROLS 6.9.1.5 Reports required on an annual basis shall include:

a. DELETED

b. DELETED

c. The results of any specific activity analyses in which the primary coolant exceeded the limits of Specification 3.4.49. The following information shall be included: (l) Reactor power history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was'exceeded; (2)

Results of the last isotopic analysis for radioiodine performed prior to exceeding the limit; results of analysis while the limit was exceeded and results of one analysis after the radioiodine activity was reduced to less than the limit. Each result should include date and time of sampling and the radioiodine concentrations; (3) Clean-up system flow history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded; (4) Graph of the 1-131 concentration and one other radioiodine isotope concentration in microcuries per gram as a function of time for the duration of the specific activity above the steady state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radioiodine limit.

MONTHLY OPERATING REPORT 6.9.1.6 DELETED SALEM - UNIT 2 6-21 Amendment No. 262

ADMINISTRATIVE CONTROLS

6. 9.1 .9 CORE OPERATING LIMITS REPORT (COLR)

a. Core operating limits shall be establi ed prior to each reload cycle, or prior to any remaining por on of a reload cycle, and shall be documented in the COLR for the following:

1. Moderator Tempe e Coe ient Beginning of Life (BOL) and End of Life OL) limits and 0 ppm surveillance limit for Specificatioý 3/4.1.1.3 4,

2. Control Bank Insertion Limits for Specification 3/4.1.3.5,

3. Axial Flux Difference Limits and target band for Specification 3/4.2.1,

4. Heat Flux Hot Channel Factor, FQ, its variation with core height, K(z), and Power Factor Multiplier PFxY Specification 3/4.2.2, and

5. Nuclear Enthalpy Hot Channel Factor, and Power Factor Multiplier, PF8H for Specification 3/4.2.3.

6. Refueling boron concentration per Specification 3.9.1

b. The analytical methods used to determine the core operating limits shall. be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1. WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology, (W Proprietary), Methodology for Specifications listed in 6.9.1.9.a.

SALEM - UNIT 2 6-24 Amendment No. 2-67 LAR S09-03 LR-N09-0164 UNIT 2 FACILITY OPERATING LICENSE PAGES WITH PROPOSED CHANGES Facility Operating License DPR-75

(2) Technical Specifications and Environmental Plan The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 275, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

Spocial Lew Pow. r Test Program PSGE4&C al complete the training porti o f the Speeial Lew Power Toot Program in acoordanco ;aith -P S,R :&G's letter dated September 5, 1980 and in accordanoo with the Commission's safoty Evaluation Roport "Spocial Low Powor Test Program", dated August 22, 1:980 (See Amendment No. 2 to IDPR 75 for the Salem Nueloa-r Generating Station, Unit No. 2) prior to operating the facil-it-y at a powor level abovo five pereont.

Within 3! days following complotion of the power asconojon tosting program outlinod in Chapter 13 of the Final Gafoty Analysis Roport, PSE&C shall perform a boron mixing and cooldown tost using decay heat and Natural Circeulation. PSE&CG shall s UB m .... test preaur. to tne for.R.revi.w and approval pri:or to porforman.e of t test. The ro sult of.

-I-q

-i -Aoo shall be submitted to the NRC prior to starting up

.. iLke:ng Uflc ]I2Lrs retuellng eutage.

(4 ) Tpi#iql T-Ppt P:r------

PSE&C shall conduct the post fuel loading initial test program (set forth in Chapter 13 of the Final Safety Analysis Report, as amended) without making any major modifications of this program unless modifications have boon identified and have receivedpro NRC approval. Major modifications arc defined as.

(a) Elimination of any tost identified in Chapter 13 of the Final Safety Analysis Report, as amended, as essential; (b) Modific-ation-of- test objectives, methods or acceptance criteria for any toot identified in Chapter 1:3 of the Final:

Safety Aalysis ReTprt, as amonded, as essential; (e) Perfor-mance of any toot at a power level different by mere than five percont of rated power from there described! and Amendment Ne. 275

- 5-6 -

PAGES 5 AND 6 ARE INTENTIONALLY BLANK ITEMS 3 THROUGH 9 DELETED (d) Failure to complete all tests inc :Lupeiein the described program (planned or scheduled for S - - levels up to the

-authoer-ized power- level) prior to OXOOCpiflu ai coro rIu9=rnun ef 120 effective full power days.

instrument- Trp seepints PSE&C shall submit for NRC review within sim months of the date of issuance of this operating licence the following values forp instrumentation channel:

(a) the Technical Specification allowiable value (the Technical Specification trip setpoint plus the instrumer drift assumed in the accident analysis),

t (b) .

the in strument .. drift assumed to occur during the Inter-.

betweten Technical Specification surveillance tests;

. the .) components ef the cumulative instrument bias; and (d) t~he maximum margin between the Technical Specification value assumed n the aecciden.t analysis.

TL .... T J -- L (f-ri I t, Ufllf irWn TP i-crM~

T1q Prior to exceeding five percent rated thermal power, PSE&C w :ill resolve to the satisfactioen of th-e NRC's Offiee of Inspectie and Enforcement all remaining construction and testi-ng deficiencies on the SP4II 6 Open items List designated for cmletion prior to the commencement ef power range testing.

All -i-steditems deferred beyond the commencement of power range testing will be subject to review by NRC Region 1 (7) it vc " u:t,:t tto:t v -- ----- ------ -- -

By June 1, 1983, PSE&C shall implement to the satisfaction of the ýýC the provisiens of Regwulatory Cuide 1.97, "Instrumentatien for Light Water Cooeledl Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," a modified by PCE&C's commitments to N1JRC 0588 and NUREC 0727.

(a) Within 4 monthc after issuanco of the license, PSE&C sThall provide a Technical Specification listing of mechanical snubbers. in the interim, PSE&C will- conduct a comprehensive mechanical snubber inispection program implemented by plant instructions.

(b) The functional testing of hydraulic and mnechanical snubbers -in accorxdance with Technical Specification 3.7.9 shall copmmence with thc first refueling outage. Thc inilti-al Efunctional testimng shall be comp-leted prior to resuming power operation following the first refueli-ng outage.

PSE&C shall take the following remedial actions, or alternative actions acceptable to the NRC, with regard to the environmental qualification requirements for Class IB equipment:

(a) No later- t~han June 30, 1982, t~he wide ranage resistance

-temperature detectors for the reaer eeelant syste shall be qualified for radiation exposure for the 40 yea-r plant life and appropriate exposure condition due to design basis accidents. Pending completion of suc-h qualificeat-ion and -acceptanceby the NRC, PSE&C shal replace each of these detectors at each refueling outage.

(b) Prior to completion of the first refueling outage or June 30, 1982, whichever is earlies t, PSE&C shall replace the Scotcheast No. 9 resin seals, used at t-he ellectrical ecnnection interface on the NtMCO limit switches, with Conax Electric Conduction Seal Assemblies.

(c) By no later than Juane 20, 1982, all safety related electrical equipment in the facility shall be qualified

-in acco-erd-ance with the provisions of: "Cuidelines for Evaluating Environmental Qualification of Class 1B Electrical Equipment in Operating Reactors" (DOR Cuidel~ines);oer DFJEC 0588, "Iinterim Staff Position i rcnmenea+/-+Ua-lricaeion or Larecy ne-acep ui-cericai Equi pment," December 1979.

_Referencs are to the appropriate sectins of the Safety Evaluation Repeort (MJREC 0517) and its supplements.

7 tl- L t-I and auditable records must beI avalalean maincalnea ac a eenEra- lOCaloen wniCn aescrioe Eao environmontal qualification method used fer all safety related clectrical equipment in cufficient detail to doceument the deg ree of campl1ianee with the DOR Cuidelines o~r 4NU4PEC 0588. Such records should be updated and maintaincd current as cquipmcnt is replaced, further tested, or otherwise further qualified to document

.. mplete compliance by

5U, +/-*U*.

(e) Within 90 days of receipt of the equipment qualification safety evaluation, the licensee shall either (i) provide missing documentation identified in Sections 3 and 4 of the equipment qualification safety evaluation whicha will demonstrate compliance of the applicable equipment with MUREC 0588, or (ii) commit to corrective actions which will result in documentation of compliance of applicable equipment- with NUEG 0588 net- later- than June -30, -1982.

(10) Fire Protection PSEG Nuclear LLC shall implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report, and as approved in the NRC Safety Evaluation Report, dated November 20, 1979, and in its supplements, and in the NRC Safety Evaluation dated January 7, 2004 subject to the following provision:

PSEG Nuclear LLC may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

Amendment No. 242 I

- 8 THROUGH 20 -

PAGES 8 THROUGH 8, 9, ATD 10 20 ARE INTENTIONALLY BLANK ITEMS 11 THROUGH 25 DELETED Amendment No. 1, 2-&, 117

9 i D #. .. . . . . .

(11)

Wihn90- days after issuaneo of the lieense, PSE&G shall demonstratod to the satisfaction of the _NRC_that the present coent-a-inment isolation provisions for the main feodwator lines-eomply with the requirements of General Design Criterion 57 under all postulatcd accidont eondi-tions, or propooc a design chango that will achiovo complianooe. if necoccary, the design change shall be implemented during the first refueling ouatago.

I Prior to excooding 50 percont pe we .r, PSE& shallfcomp 1000e Ofl prooporational testing of the rema ining throoe of six11 circulators to be tested in the main ........ .... .f.

circulating water system.

/ I ') N n 4, m -P 4: 11 11A In ý 4 n n I I linn I pmgýn

_Fp r-: r-.R P pp p:A P pp Alq PSE&G shall also roport for the Salem facility any information roportod for the Hops Greck facility relating to circumstanco which suggest that the risk from -flammablegas clouds (resulting from river traffic accidonts on the Delaware River+

varies significantly from that previously considorod.

Amendment No). 1, 25, -117

(14 oWt ionam Teupplomonti Q5),Spleet n Woolen I Sup p 1 emne-ntý-&

Prior to exceeding 90 percent power, PSE&C shall perform a too program to show that unacceptable watorhammer damage will net result from anticipated feedwater transients to tho ta generator. Prior to performing the tost program, PCE&C shall obtain NRC approval of the test proooduraos.

(15) Prior t: iOn following __ - I -

PSE&G shall submit the details of the inspection program for control rod guide thimble tube wall wear for NRC (e) ......... isolation Vl... S-&ectiOn 5.3.2, Supplement-5)

PSE&C shall install leak test connections on the pressure isolation valves; until installation of the loak ts eenneetiens, PSE&C may substitute multiple valve la tests for Technical Specification 3.4.7.2.f, such ta the cumulativo leakage from twe valves in parallel lines shall not oxceed twe gallons per minute, and the cumulative leakage from three valves in parallel lines shall not exee three gallons per minuto PSE&C shall implement the following design and produl mealtlcateLns wMAL resnect to aicscc oencraeor reliability-(i) Completo a formal training program for all th mechanical and electrical maintenance and-quality control personnel, including supervisers, who -arc responsible for the maintenance and availability of the diesel generators. The depth and quality-of this training program shall be at least equivalent to that of training programs normally conduct~ed-by

-Amendment No. 29

13 (ii) Dovoloc ocora ting procoduros that roquire loading the diosol ongine to a minimum of 25 poroont of full load for ono ho~ur aftor eight hoi irs of continuouc load oporaticJJa or as rocomownenao by tho engino manuif aturor -

(o) Containmont Sump-Mod-elTest .A7'.nr- c A 1>

47 Supplement~--~+ A)'

PSE&C shall submit the confirm~atory results of the sontainmont sump modol test program, along with a dosoription of resulting tests.

PSEiC shall install a socond level1 of undervoltage protectionp for thoe omergency buses.

(g) Roactor Containment Eloctrical Penetrations (Secation 0 A 2) 0- - A\N PSE&C shall add a fuso in seoies with tho primary-dovico of cach one of 12 circuits fed from 230 volt as moetor control contors to provide backup protection for reactor conta-inm~ent electr-ical ponetr-atio-n~s. ERach fuso shall be located in a indepondent compartment in tho co~ntrol center- of theoreen £-

n~rimrv r d; i k3:t) toss or Noen class IE Instrumnenlat~ation and Control Powor-Bus&

- ' -'- ---- - 2 f r. - 5.. 2 PSE&C shall implem ent tho dosign modificeations identified in tho PSE&C lotter datod July 31, 1930 prior to resumingpor operat-ion fol-lowing tho first refueling outag-e-.

/I ','\

k71-

-t i +/- Utjj:ki+/-t:! ýnspeetj-en teeet:ten -4 . S . I , Supplement n )

Prior to resuming power operatien following the socond refueling outage, PSE&G shall subject the law pressure turbines to an insorvice inspeetien. The inspoctien shall consist of

visual and volumotric oxcaminations. The vi suiaI xm-inPation shall be applied to !00 porcont of all the accossible surface of the rotorc, dises Eand baig The vo-lumot-ric oxamination shall use an ultrasonic tochnigue to fully examine the boro and keyway rogion of the di-scs in oach low pressure turbine.

The inspoction roesullts and evaluation of this insorv11ico-inspoction shall be roportod to the NRC and shall be accoptod by the NRC prior- to startup following t-ho socond refueling out ago fI a N 174 1,--ý 4' -, 0 1 C'rnX PCE&shal cnduct .t prooperational vibration dynamic of feet-s oesE program ior ail a ýSMB 1, 2 and 3 piping systems and piping restraints during stai .tup test programs and initial eperatien.,

(19) Differential ,-~!BE:t -O2.Ea1 ~i iJLar Fi= -. n PC3lfIm1RP 1Ail Supplement.,-...7 4)

PSB&G sh~all esorbain Abaseline diata regarding differential pressure acrios the i omlb pressuro taps in tah rtant coolant loop for various pump combinations- dulr-ing startup and initialI operation.

(20) Engineered Safety Fetur Reset Contrel (Sect-ion7.0 Supplement 5) in conformanco with 11- EPullet-in 80 06, PSE&C shall cor-rect the reset actions for the two sets of valvos identified in the PSE&G letter dat~ed Juno 13, 1980, as corrected by the PSE&G letter dated July 18, 1980, prior to operating the facýility at-a power level abov. five percent. PSE&G shall also p.rform the additional testing required by IE Bulletin 80 06 prior to Anporatinn ;;abovo fivo, nplrcon# nnwP4r

15 (21) 1

-,-= 11 I 'r-'m

- -~ - n~ '

-- lp- -

- p - iI

- 9 ý ý I~r (a) Prior to re .um.ng power operation ideoawing the t.rst refuelingoeuatage, PECshall provide ~ a detailed survey 1- -- 1 (b) Prior to operation above five percent power, control room mperators shall be traind -inthe recognition an maitigation of LI.P performance degradation.

iZZ 1 UI lplpf~a iat+B Protootion Organ-isateion (Section PSE&C shall complote the I reorganization actions and programs protection no later than Nevembcr 1, 1981. e ivt- a -a ~

0,3; pateoer:)v T MasoRnrv Walls (Section 3.8.3, Supplement 5)

(a) Prior to operatie ,n anove five percent power, PSE&C shall-I submit the infertnatien requested in the tttt,- letter dated January 8, 1981.

(b) Prior to startup following the first refueling, PSE&G shall resolve the differonce betwee~n theq cta;ff cri tri and- theriteria used by PSE&G to the sati-sfac-tion of the

-NRC -anPd -implementthe required fixes that might result froma such as resolution.

-buUuicmcIlE I

IZ'1I 11"1i 1~.OEi~Dfl Vi3fl uGflplLlofl2 W~ZiE1Ofl ~

Unless otherwise noted, cach of the following con dit ions references the a ppropriato section of Supplement No. 5 to the Safety Evaluatie: Report (IFJRBC 05-17) for the Sal ueneraoicn zeaicin, unRc z, aaesa danuary +/- asire fnuii DU completed to the satisfaction of the NRC by the times indicated.

(b) Short-Tle3 ?,Iysis and Preeedure Re:V4--li (Section 22.2, I.C.!. andl 1.U..;,

The operators shall be br-iefed- R*;n; the roicvisions to the emergency operation instructios ;ii W: L.£Z.L *i .* * ",*A * ",: ",j L..L "." ",* * %Z& :

power days of oper-ation.

Amendment No. -19

(c) Auxiliary Fccdwciter Gy--tem Reliability-Evaluatien (Seeti:

22.2, II.E.l.!)

(4:) PSE&G shall in-stallI auxi~iry rcawaoecr sTor-age tank level -ind-ipatic and alarms in accorFdan-;;

with the S& ot F of: May 5, 198 0 prier to t refueling.

(44) PSE&G shall porform a 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months endurance test an all aux~iliary feedwater system pumnps prior to operatin at 100 porcont pewcr. PSE&C shall previde a reper-t on the results- of ctst NRCwihi thos 60 days of complot-ion of the tests.

(ii44) PSE&G shall resolve to Ng-RC' sRat-isfaet-ion the issue een erning time available for operator action to prevent pump damage prior to operation above five pcrcont power.

/ ..,] \ T"t.... ,,-* r* ......

kt=ty 22grade nmergentzýy rt-týý uttt:ýb - -------- ----- ---- I.

and Section 22.3. IIT.A.2) i) No later than 90 days from the date-o issuance vW status of any items related to emergency preparedness identif ied by FEM-. or the NRC as reauuirinGF furtheracin (i4) PSE&C shall piroevido metcorological and doso assessment rematc interrogation capability to mneet the criteria of Appendix 2, NUREC 0654, Revision 1 as follows: (a) a functional dscripti-n o-f:

upgraded capabilities by January *1, -1982, (b) installation of hardware and software by July -I, 1982 provide-. tat NRC approval -iS r-c....ived by fo.ur months prior to that time and (e) full operation capability by October 1, 1982.

(4:iii) PSE&:G shall provido substantiation that the baek up seuree of meteereligieal inferm-atien -fcromthe NWS Off ico, Greater W-ilmington Airpert adoguatoly charactoricoce the site conditions with respect to wind diroction and wind speed by Jualj, 1, 1981.

(4:v) PSE&G shall providoe substantiation that uncortainties associatod with plume trajectory prediction, associatod with the occurro-nco of sea land broco irclaion wihi the plume oxpesure pathway zono, are compatiblo with the plannod rocommondati-nc fo pretoctivo actions that would be based upon cuch projoctions by July !, 1981 (iw) For these systems in which leakage +'s measurodd during shutdown, PSE&C shall make and roport leak rate measurements prior to initial startup.

(44) For these systems in which leakage is measured during aperationc, PSE&C will make and roport leaks rate measurements within 60 offoctivo full power days of plant operation.

Each of the following conditions roforoncoc the appropriate coction of Supplement No. 5 to the Safety Evaluation Report (NMREG 0517) for the Salem Nucicar Conorati:ng Station, dated January 1981, and shall be completod to the cat-isfac-tion o-f thoe NRC by the times indicatod.

(Soction 22.3, ICI PSE&G shall imploment the requirement of item I.C.1-specified in i~REG 0737, "Clarification of THI Action Plan Requirements,"1 no lator thimplomontation than datec established in kFECB 0737-.

f I-N 1ý1- - -- ý--l - ný - ý -- -- , - / ý--, 4 TT U PSE&G shall submit procedural guidelines fer and a description efthe reactor coolant system vents by July 1, 1981. The reaetor coolant system vents shall b instlledno later than July 1, 1982.

ý .4 PJ:Iq#n Rp p4 1pC4n : pnn nr n Pp-:; -. I. - - ;

PSE&C shall complete modifications to assure adequate access t vital areas and protection ef safety .quipm.nt following an accident resulting in a degraded core not later than January . 1, 1982.

(d) Deleted (e) Relief, Safety and. Blocak vaLvc~ rc~u uirc~mc~n~rzi PSE9C shall qualify the reactor coolant system relief, safety and blocek valves under expected .p.rating conditions for design basis transients and accidents in ac..rdanca w.it.h. the plant specific requirements and schedules established in NTECB 0737, "Clar-ification of THI1 Action Plan Requaireoments."L Fed2: e i--, ta=e n ý PSEIC shall upgraae, as necessary, automatic initiati En 6E the auxilia rv foedwater system and indication of auxiliary f eedwater fl uvW to each steam generator to safety1 grad* eO qualiey no iater tnan July 1, 1981.

Amendment No).23S

(g) ontainment isolati en i)epenaa!ýý kGeet;-en 22.3, 44- EI2 (i) PSE&C chall reduco tho containmont sotpoint prossure that initiates contaminaent isolati on for nonossontial ponetrations to the minimum co inpatible conitUionsI no0 laterC F han Ju 11 1, 1981.

(idi) PSEWs shall -install high radiation isolation signal on tho canto infnefit isuree and vent _selatle (h) Additional Accidont Monitori.

PSE&C shall install and demonstrate the operability of instruments for continuous indication in the control room of the following variables. Each itom shall be completed by the specified date in the condition:

(i) Containment pressuro form minus fivo psig to throe tim.es tho design pressure of tho containment no

- lator than January -1, 1:982;

,) Containment...water loal from (i) the- bottom to the top of the containment sump, and .ii) the bottom of the containment to an elevation equivalent to a

600,000 gallon capacity no later than July !,191 (iii) Containment atmosphere hydrogen concentration from 0 to 10 volume percent no later than July 1, 1982; 2.C-(25) (h) (i~v)

Containment gamma radiation up to !Q' rod/hr. at the first outage of sufficient duration but no later than prior to startup following the first rofueling outage, and (V) Noble gas effluent from sach potential release I

point tram normai concentrations up to 10 -nGltee (Xe 133) no later than prior to startup following the first refueling outage.

PSE&C shall provide the capability to continuously sample gaseous effluents and analyze these samples no later than prior to startup following the first refueling outage.

Until4 the aboeve intlato s completed, PSE&C shall use interim monitoring proceduresan Amendment No. 9

20 p..... a provide the capability t. .ontinuously sample gaseous effluents and analyze these samples no later than January 1, 1982.

Until the above installlatJIon -is completed, PSE&G shall use inter-im monitoring procedures and equipment.

PSE&C shall install and demonstrate the oporabiliqty of additional instruments or controls needed to supplement installed equipment in ordler to provide un-ambiguous, easy to inter-prot indip;ation of inadequate core cooling at the first outage of sufficient duration but no . later than prior t startup following the first rfueli-ng eutage--

I./'%

PSE&G shall submit a detailed analysis of the thermal meehanieal conditions in the reactor vessel during recoyc ry from small breaks with an extended less of all

-r, -- A '..- ý - - - - I - ý - - ý 1, - - 7-, I I o n~

(14E) Anlysis. ef Vei.ing Pet-nt-l (Section 22.3, .7 PSE&C shall analyze the potential for voiding inth rcactor coolant system (RCZ) during anticipatcd transients. PSE&G shall submiýt this analysis no later than January 1, 1982.

22.2, II.K.219 PSE&. shall provide a benchmark analysis of sequential auci liar' foedwater (AFW) flow to the steam generators following -Aloss- of main feedwater no later than January 1, 1982.

(Section 22.3, II.K.3.25),

PSE&C shall deemnby analysis or experiment, the consequences of a loss of cooling water to the reactor coolant pump seals. PSE&G shall submit the results of the evaluation and prepoced modifications no later than January 1, 1992.

Amendment No. 4

- 21 -

(n) Rovisod Small Break Loss of Coolant Accident Methods (Section 22.3, 1I.K.3-30)

PSE&C shall eomply with the requirements of this position as specified in !PJEG 0737, "Clarification of TMI Aetion Plan Requirements."

(e) Compliance Wth,410 QR Par 50.46 (Section 22.3

. K. .31)

PSE&C shall perform plant spfcific caleulations usi NRC approved models fer small break loss of coolant-accidlnts (LuCAs) to show compliance with 10 CFR Part 50.46. PSE&C shall submit these calculations by January 1, 1983,or Cond yar after NRC approvalef LOCA analysis mtodls, whiehnvt r is later, enly if model changes havs been made.

Pie&C shal maintain L -in ffsh t anoe nthm Tacicai Support Conter and an interimn Emergency Operations F-aeili-Ly until such time as the final facilities are complete.

(26) Additional Conditions ii The Additioinal Conditions contained in Appendix C, as revised through Amendment No. 227 are hereby incorporated into this license.. PSEG Nuclear LLC shall operate the facility in accordance with the Additional Conditions.

(27) PSE&G TO PSEG Nuclear LLC License Transfer Condtions

a. PSEG Nuclear LLC shall take all necessary steps to ensure that the decommissioning trust is maintained in accordance with the application, the requirements of the Order Approving Transfer of License and Conforming Amendment, dated February 16, 2000, and the related Safety Evaluation dated February 16, 2000.

b. The decommissioning trust agreement shall provide that:

1) The use of assets in both the qualified and non-qualified funds shall be limited to expenses, related to decommissioning of the unit as defined by the NRC in its regulations and issuances, and as provided in the unit's license and any amendments thereto. However, upon completion of decommissioning, as defined above, the assets may be used for any purpose authorized by law.

Amendment No. 227

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MONTHYEAR2009-09-14 [Table View]

LR-N09-0164, License Amendment Request to Correct Technical Specification and Facility Operating License Editorial Items

Text

Subject:

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

Reference:

Reference:

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

5.0 REGULATORY ANALYSIS

6.0 ENVIRONMENTAL CONSIDERATION

7.0 REFERENCES

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LR-N09-0164, License Amendment Request to Correct Technical Specification and Facility Operating License Editorial Items

Text

Subject:

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

Reference:

Reference:

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

5.0 REGULATORY ANALYSIS

6.0 ENVIRONMENTAL CONSIDERATION

7.0 REFERENCES

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