LR-N10-0032, Request Amendment of Proposed Revision to Technical Specifications - Removal of Reactor Coolant System Structural Integrity Requirements

Request Amendment of Proposed Revision to Technical Specifications - Removal of Reactor Coolant System Structural Integrity Requirements

ML100920052
Person / Time
Site:	Salem, Hope Creek
Issue date:	03/25/2010
From:	Braun R Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LAR H10-02, LAR S10-02, LR-N10-0032
Download: ML100920052 (27)

Text

PSEG P.O. Box 236, Hancocks Bridge, NJ 08038-0236 0 PSEG Nuclear LLC 5 2010 10 CFR 50.90 LAR S10-02 LAR H10-02 LR-N10-0032 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Salem Generating Station - Unit 1 and Unit 2 Facility Operating License Nos. DPR-70 and DPR 75 NRC Docket Nos. 50-272 AND 50-311 Hope Creek Generating Station Facility Operating License No. NPF-57 NRC Docket No. 50-354

Subject:

PROPOSED REVISION TO TECHNICAL SPECIFICATIONS -

REMOVAL OF REACTOR COOLANT SYSTEM STRUCTURAL INTEGRITY REQUIREMENTS In accordance with the provisions of 10 CFR 50.90, PSEG Nuclear, LLC (PSEG) hereby requests an amendment of the Technical Specifications (TS) for the facility operating licenses listed above.

This license amendment request proposes changes for consistency with the requirements of 10 CFR 50.55a. Salem TS 3/4.4.10 (3/4.4.11, Unit 2) and Hope Creek TS 3/4.4.8, "Structural Integrity," would be deleted, removing the reactor coolant system structural integrity requirements. The proposed change is consistent with NUREG 1431, "Standard Technical Specifications, Westinghouse Plants, Revision 3.0," and NUREG-1433, "Standard Technical Specifications General Electric Plants, BWRI4, Revision 3," in that it does not meet the criteria of 10 CFR 50.36 for inclusion in the TS. Concurrently, Salem TS Surveillance Requirement 4.4.10.1.1 (4.4.11.1 for Salem Unit 2), augmented inspection for the Reactor Coolant Pump (RCP) flywheel, will be relocated as TS 6.8.4.k, a new TS Program.

PSEG has determined that this LAR does not involve a significant hazard consideration as determined per 10 CFR 50.92. PSEG's technical and regulatory evaluation of this LAR, the TS changes, and the TS Bases changes (for information only), are provided in Attachments 1, 2 and 3. There are no regulatory commitments in this submittal.

Document Control Desk 2

MAR 26 201 LR-N 10-0032 evaluation of this LAR, the TS changes, and the TS Bases changes (for information only), are provided in Attachments 1, 2 and 3. There are no regulatory commitments in this submittal.

The changes in this LAR are not required to address an immediate safety concern.

PSEG requests approval of this LAR within one year of the submittal date. Once approved, the amendment will be implemented within 60 days from the date of issuance.

These proposed changes have been reviewed by the Plant Operations Review Committee. In accordance with 10 CFR 50.91, "Notice for Public Comment; State Consultation," a copy of this application, with attachments, is being provided to the designated State Official.

Should you have any questions regarding this submittal, please contact Mr. Jeff Keenan at (856) 339-5429.

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

Executed on _,_________

(Date)

Sincerely, Robert C. Braun Senior Vice President Attachments (3)

CC 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 Commitment Coordinator - Hope Creek PSEG Corporate Commitment Manager

LR-N10-0032 Salem Generating Station - Unit 1. and Unit 2 Facility Operating License Nos. DPR-70 and DPR 75 NRC Docket Nos. 50-272 AND 50-311 Hope Creek Generating Station Facility Operating License No. NPF-57 NRC Docket No. 50-354 LICENSE AMENDMENT REQUEST LAR S10-02 and H10-02 PROPOSED REVISION TO TECHNICAL SPECIFICATIONS - REMOVAL OF REACTOR COOLANT SYSTEM STRUCTURAL INTEGRITY REQUIREMENTS Table of Contents

1.

D E S C R IPT IO N................................................................................................

1

2.

PROPOSED CHANGE....................................................................................

1

3.

BA C KG R O U N D...............................................................................................

1

4.

TECHNICAL ANALYSIS................................................................................

3

5.

REGULATORY SAFETY ANALYSIS................................................................

4 5.1 No Significant Hazards Consideration...................................................

5 5.2 Applicable Regulatory Requirements/Criteria........................................

6 5.3 P recedents...........................................................................................

.. 6 5.4 C onclusions...........................................................................................

6

6.

ENVIRONMENTAL CONSIDERATION...........................................................

7

7.

R E FE R E N C ES.................................................................................................

7

LR-N 10-0032

1.0 DESCRIPTION

This submittal requests an amendment to Technical Specification (TS) 3/4.4.10 (11)1 of Facility Operating License Nos. DPR-70 and DPR-75, Salem Units 1 and 2, and TS 3/4.4.8 of Facility Operating License No. NPF-57, Hope Creek Generating Station.

These TS, Reactor Coolant System (RCS) "Structural Integrity", would be deleted, removing the reactor coolant system structural integrity requirements which are redundant to the requirements of 10 CFR 50.55a. The proposed change is consistent with NUREG 1431, "Standard Technical Specifications, Westinghouse Plants, Revision 3.0," and NUREG-1433, "Standard Technical Specifications General Electric Plants, BWR/4, Revision 3," in that it does not meet the criteria of 10 CFR 50.36 for inclusion in the TS. Concurrently, the Salem augmented inspection for the Reactor Coolant Pump (RCP) flywheel of TS Surveillance Requirement (SR) 4.4.10.1.12 will be relocated as TS 6.8.4.k, a new TS Program.

2.0 PROPOSED CHANGE

The RCS Structural Integrity TS and the related Bases are deleted in their entirety.

Concurrently, Salem TS SR 4.4.10.1.1, augmented inspection for the Reactor Coolant Pump (RCP) flywheel, will be relocated as TS 6.8.4.k, a new TS Program.

The requirements of Salem TS SR 4.4.10.1.23, Augmented Inservice Inspection Program for Steam Generator Channel Heads, will be deleted as the surveillance was applicable only to the first three refueling outages and is therefore historical.

The associated TS Index will also be revised to delete reference to the TS and the associated Bases. The changes to the affected TS Bases pages will be incorporated in accordance with the Technical Specifications (TS) Bases Control Program.

Attachments 2 and 3 contain a markup of the affected TS and Bases pages, respectively.

3.0 BACKGROUND

The RCS Structural Integrity TS specifies the requirements for maintaining the structural integrity of the ASME Class 1, 2 and 3 components. This specification was originally intended to support assurance that structural integrity and operational readiness of these components are maintained at an acceptable level throughout the life of the facility. The specification is applicable in all operational modes. However, For Unit 1, the Structural Integrity TS is TS 3/4.4.10, for Unit 2 it is numbered TS 3/4.4.11. For simplicity in the discussion in this submittal, the 3/4.4.10 numbering is used to represent both Units. Attachment 2 and 3 contain the appropriate mark-up of both Unit 1 and Unit 2 TS.

For Unit 1, the RCP flywheel TS SR is 4.4.10.1.1, for Unit 2 the equivalent SR is numbered 4.4.11.1. For simplicity in the discussion in this submittal, the 4.4.10.1.1 numbering is used to represent both Units. Attachment 2 and 3 contain the appropriate mark-up of both Unit 1 and Unit 2 TS.

For Unit 1, the Augmented Inservice Inspection Program for Steam Generator Channel Heads TS SR is 4.4.10.1.2, for Unit 2 the equivalent SR is numbered 4.4.11.2. For simplicity in the discussion in this submittal, the 4.4.10.1.2 numbering is used to represent both Units. Attachment 2 and 3 contain the appropriate mark-up of both Unit 1 and Unit 2 TS.

1

LR-N10-0032 the specification does not provide actions for plant shutdown if its Limiting Condition for Operation (LCO) is not met. In addition, the specification contains no surveillance requirements other than reference to TS 4.0.5, the Inservice Inspection (ISI) Program (and, for Salem only, the augmented RCP flywheel inspection (4.4.10.1.1) and 4.4.10.1.2, Augmented Inservice Inspection Program for Steam Generator Channel Heads (which is historical)). This is because the specification addresses the passive pressure boundary function of ASME Code Class 1, 2 and 3 components as established by compliance with the ISI program. The ISI program is required pursuant to 10 CFR 50.55a, "Codes and Standards." Furthermore, the specification wording could be misconstrued to conflict with normal outage-related activities, including removal of the Reactor Vessel head in preparation for refueling, a time in which the RCS pressure boundary would no longer be intact. This TS does not fulfill any of the criteria of 10 CFR 50.36(d)(2)(ii) for retention in the TS.

Maintaining a program-type requirement within an LCO creates significant interpretation issues for Operations personnel. The RCS structural integrity TS was part of the original TS. It appears to have been included to help ensure that plant heat up and startup would not occur until all required ASME Code Class 1, 2 and 3 components were verified to meet ISI acceptance criteria following inspections performed during a plant outage (normally performed during refueling outages).

Meeting this acceptance criterion helps ensure the integrity of ASME Code Class 1, 2 and 3 components during all modes of operation, including accident events.

However, the RCS pressure boundary and other ASME Code Class 1, 2 and 3 components are purposely breached during Mode 5 and 6 operations to support plant outage activities and such openings are not historically considered a violation of the RCS Structural Integrity TS. Furthermore, the RCS Structural Integrity TS contains no actions suggesting it was designed to accommodate integrity concerns once plant heat up has commenced. ASME Code Class 1, 2 and 3 component structural integrity ISI activities are performed primarily during plant outages when conditions exist that permit access, or are controlled through application of the ISI program during the operational cycle. Other TS are designed to monitor structural integrity during operation and provide actions to shutdown the unit if compliance is not maintained. For example, RCS heat up and cool down rates protect against applying undue stresses as a result of pressure/temperature transients on RCS components and piping. RCS leakage TS provides a means of protecting the RCS integrity by detecting and monitoring leakage. Therefore, it is not necessary to apply the RCS Structural Integrity TS when integrity issues become evident during plant operation above cold shutdown. Because the RCS Structural Integrity TS is redundant to other regulations, it is acceptable to remove the TS requirements from the TSs.

2

LR-N 10-0032

4.0 TECHNICAL ANALYSIS

The purpose of the RCS Structural Integrity TS is to specify the requirements of maintaining the structural integrity of the ASME Class 1, 2 and 3 components.

However, this is redundant to and does not contain the detail of the requirements contained within 10 CFR 50.55a, "Codes and standards." 10 CFR 50.36(c)(2)(ii) states that a TS LCO of a nuclear reactor must be established for each item meeting one or more of the following criteria:

Criterion 1.

Criterion 2.

Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

This specification is not applicable to installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. ASME Code Class 1, 2, and 3 components do not include any instrumentation. This specification does not meet Criterion 1.

A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

The RCS Structural integrity TS requirement is neither a process variable, design feature, or operating restriction that is an initial condition of a design basis accident (DBA) or transient analysis. Structural integrity is not monitored or controlled during plant operation; it is verified during periodic inspections. Compliance is maintained by meeting the requirements of 10 CFR 50.55a through implementation of the Salem Units 1 and 2 and Hope Creek ISI programs. This specification does not meet Criterion 2.

A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

ASME Code Class 1, 2, and 3 components that are part of the primary success path and function to mitigate DBAs or transients that either assume the failure of, or present a challenge to, the integrity/operability of these components are included in the individual specification that covers these components. Each TS SSC must continue to meet the requirements of 10 CFR 50.55a as implemented through the Salem Units 1 and 2 and Hope Creek ISI programs. The RCS Structural Integrity TS Criterion 3.

3

.R-N10-0032 addresses only the passive pressure boundary function of these components. This specification does not meet Criterion 3.

Criterion 4 A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

The requirements covered by this TS have not been shown to be risk significant to public health and safety by either operating experience or probabilistic safety assessment. In addition, failure modes of applicable structures, systems, or components (SSCs) would not be identified from the requirements of this TS. Furthermore, the requirements of this TS do not affect the risk review/unavailability monitoring of applicable SSCs.

Each TS SSC must continue to meet the requirements of 10 CFR 50.55a as implemented through the Salem Units 1 and 2 and Hope Creek IS[

programs. This RCS Structural Integrity specification does not meet Criterion 4.

Therefore, this TS does not fulfill any of the 10 CFR 50.36(c)(2)(ii) criteria for items for which TS must be established. The removal of the RCS Structural Integrity TS and its associated references to structural integrity eliminates from the TS the redundancy of structural integrity requirements that are already covered under 10 CFR 50.55a.

Removal of this specification does not reduce the controls that are necessary to ensure compliance with the ASME Code or the need to maintain the ASME Class 1, 2 and 3 component boundaries. Structural Integrity is maintained by compliance with 10 CFR 50.55a as implemented through the Salem, Units 1 and 2 and Hope Creek ISI Programs. In addition, for Salem only, the augmented inspection for the Reactor Coolant Pump (RCP) flywheel of TS 4.4.10.1.1 will be relocated as TS 6.8.4.k, a new Program. This is consistent with NUREG 1431, "Standard Technical Specifications, Westinghouse Plants, Revision 3.0." The requirements of 4.4.10.1.2, Augmented Inservice Inspection Program for Steam Generator Channel Heads, will be deleted as the surveillance was applicable only to the first three refueling outages and is therefore historical.

Note also that because this TS is redundant to another regulatory requirement, the inspection requirements of 10CFR50.55a, there is no need to relocate the TS to another licensee controlled document.

5.0 REGULATORY SAFETY ANALYSIS The proposed change would amend Technical Specification (TS) 3/4.4.10 of Facility Operating License No. DPR-70, Salem Unit 1, TS 3/4.4.1 of Facility Operating License No DPR-75, Salem Units 2, and TS 3/4.4.8 of Facility Operating License No.

NPF-57, Hope Creek Generating Station. These TS, Reactor Coolant System (RCS) 4

I'R-N10-0032 "Structural Integrity", would be deleted, removing the reactor coolant system structural integrity requirements which are redundant to the requirements of 10 CFR 50.55a. The proposed change is consistent with NUREG 1431, "Standard Technical Specifications, Westinghouse Plants, Revision 3.0," and NUREG-1433, "Standard Technical Specifications General Electric Plants, BWR/4, Revision 3," in that it does not meet the criteria of 10 CFR 50.36 for inclusion in the TS.

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

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

Response: No.

The proposed change to remove the RCS Structural Integrity TS does not impact any mitigation equipment or the ability of the RCS pressure boundary to fulfill any required safety function. Since no accident mitigation or initiators are impacted by this change, no design basis accidents are affected. The removal of the RCS Structural Integrity TS eliminates from the TS the redundancy of requirements that are already covered by the inspections necessary to maintain structural integrity under 10 CFR 50.55a.

Therefore, the proposed changes do not represent a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No The proposed changes do not involve a modification to the physical configuration of the plant (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. Therefore, this proposed change does not create the possibility of an accident of a different kind than previously evaluated.

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

Response: No 5

LR-N 10-0032 Removal of the RCS Structural Integrity TS does not reduce the controls that are required to maintain the RCS pressure boundary for ASME Code Class 1, 2, or 3 components. The removal of the RCS Structural Integrity TS eliminates from the TS the redundancy of requirements that are already covered by the inspections necessary to maintain structural integrity under 10 CFR 50.55a. No equipment or RCS safety margins are impacted due to the proposed change Therefore, this proposed change does not involve a significant reduction in a margin of safety.

Based on the above, PSEG concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

5.2 Applicable Regulatory Requirements/Criteria Although the RCS structural integrity controls are being removed from the TS, PSEG is still required to comply with the ASME Code requirements in accordance with 10 CFR 50.55a.

There are no changes being proposed that would result in non-compliance with any of the regulatory requirements. The evaluations documented above confirm that PSEG will continue to comply with all applicable regulatory requirements.

5.3 Precedent The proposed removal of the RCS Structural Integrity TS is similar to license amendment No. 270 issued to Arkansas Nuclear One, Unit No. 2 on March 1, 2007 (ADAMS ML070570519), and to license amendment No.264 issued to Millstone Unit 2 on February 1,2002 (ADAMS ML020370125).

5.4 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

6

LR-N 10-0032

6.0 ENVIRONMENTAL CONSIDERATION

PSEG has evaluated the proposed amendment for environmental considerations.

The review has determined that the proposed amendment would change requirements with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, and would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b),

no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

7.0 REFERENCES

None 7

LR-N10-0032 TECHNICAL SPECIFICATION PAGES WITH PROPOSED CHANGES Facility Operating License DPR-70 Technical Specification Paqe Index V, XII 3/4.4.10 3/4 4-32, 33 6.8.4.k (new) 6-19e Note: A proposed change to Page 3/4 4-33 and 6-19e is pending via PSEG LAR 509-06, submitted September 23, 2009. The LAR S09-06 change adds new TS 6.8.4.j to new page 6-19e.

Facility Operating License DPR-75 Technical Specification Paqe Index V, XII 3/4.4.11 3/4 4-33 6.8.4.k (new) 6-19g Note: A proposed change to Page 3/4 4-33 and 6-19g is pending via PSEG LAR S09-06, submitted September 23, 2009. The LAR 509-06 change adds new TS 6.8.4.j to new page 6-19g.

Facility Operating License NPF-57 Technical Specification Page E Index I XI, XIX 3/4.4.8 3/4 4-27 Note: A proposed change to Page 3/4 4-27 is pending via PSEG LAR H09-05, submitted November 4, 2009.

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION 3/4.4 3/4.4.1 3/4.4.2.1 PAGE REACTOR COOLANT SYSTEM REACTOR COOLANT LOOPS Normal Operation.........

Hot Standby..............

Hot Shutdown.............

Cold Shutdown............

SAFETY VALVES -

SHUTDOWN.

3/4 3/4 3/4 3/4 3/4 4-1 4-2 4-3 4-3b 4-4 3/4.4.2.2 SAFETY VALVES -

OPERATING.........................

3/4 4-4a 3/4.4.3 RELIEF VALVES.............................

3/4.4.4 PRESSURIZER..............................

3/4.4.5 STEAM GENERATOR (SG)

TUBE INTEGRITY......

3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection System.................

Operational Leakage......................

Primary Coolant System Pressure Isolation 3/4.4.7 DELETED 3/4.4.8 SPECIFIC ACTIVITY.........................

3/4.4.9 PRESSURE/TEMPERATURE LIMITS Reactor Coolant System...................

Pressurizer...............................

Overpressure Protection Systems..........

3/4.4.10 DELETED STRUCTURAL INTEGRITY ASME Cede Glass 1,2, and 3 Comffenents..

3/4 3/4 3/4 4-5 4-6 4-7 Valves 3/4 4-14 3/4 4-15 3/4 4-16a 3/4 4-20 3/4 4-24 3/4 4-29 3/4 4-30 3/4 4-32 3/4.4.11 3/4.4.12 INTENTIONALLY BLANK.......................

HEAD VENTS................................

3/4 4-34 3/4 4-35 SALEM -

UNIT 1 V

Amendment No.4-64

INDEX BASES SECTION PAGE 3/4.3 3/4.3.1 3/4.3.2 3/4.3.3 3/4.3.4 INSTRUMENTATION PROTECTIVE AND ENGINEERED SAFETY FEATURES (ESF)

INSTRUMENTATION..................

MONITORING INSTRUMENTATION.......

TURBINE OVERSPEED PROTECTION.....

B 3/4 3/4 3/4 3-1 3-la 3-4 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION......................

B 3/4 4-1 3/4.4.2 SAFETY VALVES..................................

B 3/4 4-la 3/4.4.3 3/4.4.4 r

3/4.4.5 3/4.4.6 3/4.4.7 3/4.4. 8 3/4.4.9 3/4.4.10 3/4.4.11 3/4.4.12 RELIEF VALVES......................

PRESSURIZER.........................

STEAM GENERATOR (SG)

TUBE INTEGRITY REACTOR COOLANT SYSTEM LEAKAGE.....

DELETED SPECIFIC ACTIVITY..................

PRESSURE/TEMPERATURE LIMITS........

DELETED STRUCTURAL INTECRITY.......

BLANK..............................

REACTOR VESSEL HEAD VENTS..........

............O B 3/4 4-la

.B 3/4 4-2

.B 3/4 4-2 B 3/4 4-4a 3/4 4-5 3/4 4-6 B

B B

3/4 3/4 3/4 4-17 4-17 4-17 SALEM - UNIT 1 XII Amendment No.

278

REACTOR COOLANT SYSTEM 3.4.10 DELETED STRUCTJRAL INTE.CRITY ACME CODEB CLASS 1, 2 and 3 COM4PONEN-TS Li'li'+/-:2if uur32"u:i'iu2 Dtun i

'tjK".llurI' J) 3.4.10.1 The structural integrity of ACME Code Class 1, 2 and 3 components shall bo maintained in accordance with Specificatien 4.4.10.1.1.

APPLICGAILITY.ý AL3 MODES ACTION:

a. With the structural integrity of any ACME Code Class 1 omponent (s) not conforming to tho abovo requirements, restore thc strucetural intogrity of the affected component(s) to within its limiit or isolate the affoctod component(s) prior to increasing the Roactor Coolant system temperature moere than 5Q02 above the minim~um temperature required by lNDT considerations.,

b.

With the structural integrity of any ACME Code Class 2 component(s)

ReE een 9EM Rg te t e a eve reeu rements, restere,Ate struet-ural t nte.ritv ot e.a..ectei come.nent.s) to within its limit or isolate the atte.te. component(s) prior to increasing the Reactor Coolant System temperature above 2000 F.

e. With the structure not conformning to integrity of the a 1 integrity of any ACMEB Cede Class the above reefirements, restore the 3-component (s) structural C C et Ca compfonentws) to wItnin its n+mit or Isolate tee arrececa componentks) tramR service.

SURnVEILLANCE REQUIREMENTS 4.4.10.1-1 The strucetural integrity of ACME Code Class 1, 2 and 3 components nhlql I-I=I-mnn-fr-:

k.

,l'c

+-he reul-remento CCl ilo

'.U.Z, ann

b.

Per the requirements of the augmented inservice inspection program specified in Specification 4.4.10.1.2.

SALEM - UNIT 1 3/4 4-32 Amendment No.

246

REACTOGR GOGLANT SYSTEM Relocated to TS 6.8.4.k

ýTMNT=T T ýXT=

nýýT=nýh=ýTMO

/ý-

4 -,,-A N

In addition to the requirements of Specification 4.0.5, each Reactor Coolant Pump flywheel shall be inspocted per the rocommoendations of Rgulatery Pesition l.4.b oef Rrgulatory Cuidre 114, Revision 1, Anugust

)

1975.

in lieu of Position C.4.b(l) and C.4.b(2), a guaalifiod in placo UT oxominatien ever the volumo from the innor boro of the flywheel to the cirelo ono half of the eutor radius or a suirfaco oxomninatien (MT and/or-PTE) of oxpesod surfacos of the refmovod flywheels mnay be conducted at 20 year intervals.

Channel Heoads The steam goneratosr- -h-annol heads shall be ultrasonical inspe.t.d during.

h of the first three refueling eutages using the same ultrasonic inspectien procoduros and equipment used to generate the baseline data.

These inservice ultrasonic inspections shall verify that the r.a..s.bs.rvd in the stainless steel eladding prior to

.p.r.ti.n honVt not propagatod into the base material. The stain-less steel elad suirfacos of the steamn gonorator channol heads shall also be visually inspectod during the obove outogos.

This may be aeeomplishod by diroot visual examination, r by rooto fm.ons su.h as toeovisien camlea.

if the visnua e3xamination, either diroct or romoete, reveals dotoctoble cladding indieations, a reeord shall be m~ado by means of a 3vidoo tape reoording or nhn~o~ih~f~r -rni~~

nllrpee.

THIS PAGE INTENTIONALLY BLANK SALEM - UNIT 1 3/4 4-33 Amendment No. 2-6--5

ADMINISTRATIVE CONTROLS Note: A proposed change adding 6.8.4.j (and Page 6-19e) is pending via PSEG LAR S09-06, submitted September 23, 2009.

6.8.4.k Reactor Coolant Pump Flywheel Inspection Program In addition to the requirements of the ISI Program, 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(1) 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.

SALEM -

UNIT 1 6-19e Amendment No.

LIMITING SECTION 3/4.4 3/4.4.1 CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS PAGE REACTOR COOLANT SYSTEM REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup and Power Operation............

Hot Standby............................

Hot Shutdown...........................

Cold Shutdown..........................

3/4.4.2 SAFETY VALVES -

SHUTDOWN 3/4.4.3 SAFETY VALVES -

OPERATING 3/4.4.4 PRESSURIZER 3/4.4.5 RELIEF VALVES 3/4.4.6 STEAM GENERATOR (SG)

TUBE INTEGRITY.....

3/4.4.7 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection System...............

Operational Leakage....................

3/4.4.8 DELETED 3/4.4.9 SPECIFIC ACTIVITY......................

3/4.4.10 PRESSURE/TEMPERATURE LIMITS Reactor Coolant System.................

Pressurizer............................

Overpressure Protection Systems........

3/4.4.11 DELETED STRUCTURAL INTECRITY ASME CEdA V

lass 1, 2, and 3 C..

.p.n.nts 3/4.4.12 HEAD VENTS................................

3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 4-1 4-2 4-3 4-4a 4-5 4-6 4-7 4-8 4-9 3/4 4-16 3/4 4-17 3/4 4-23 3/4 3/4 3/4 3/4 3/4 4-27 4-30 4-31 4-33 4-34 SALEM -

UNIT 2 V

Amendment No.

2-6%

INDEX BASES SECTION PAGE 3/4.3 INSTRUMENTATION 3/4.3.1 PROTECTIVE AND 3/4.3.2 ENGINEERED SAFETY FEATURES (ESF)

INSTRUMENTATION B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION B 3/4 3-la 3/4.4 3/4.4.1 3/4.4.2 and 3/4.4.3 3/4.4.4 3/4.4.5 3/4.4.6 3/4.4.7 3/4.4.8 3/4.4.9 3/4.4.10 3/4.4.11 REACTOR COOLANT SYSTEM REACTOR COOLANT LOOPS AND COOLANT CIRCULATION B 3/4 4-1 SAFETY VALVES B 3/4 4-2 PRESSURIZER RELIEF VALVES STEAM GENERATOR (SG)

TUBE INTEGRITY.....

REACTOR COOLANT SYSTEM LEAKAGE DELETED SPECIFIC ACTIVITY.......................

PRESSURE/TEMPERATURE LIMITS DELETED STRUCTURAL INTEGRITY.............

B B

B 3/4 3/4 3/4 3/4 4-2 4-2 4-3 4-4 4-6 4-7 4-18 B 3/4 B 3/4 B 3/4 I

3/4.4.12 REACTOR VESSEL HEAD VENTS

.B 3/4 4-18 SALEM - UNIT 2 XII Amendment No.

262

REACTOR COOLANT SYSTEM 3.4.11 DELETED'STRUCTUTRL* INTEGRITY AAA=

CflODE P.LASSq

1. P Ang :; PQ4PANETW LIP4ITING CONDITION FOR GPEEAT4ION 3.4.11.1 The structuaral integrity of ASME Code Class 1, 2 and 3 eempenents shall be maintained in accordanca with Gpeeificat-ion 4411 APPLICAPILIT' ACTIO9N-:

i!ALL P4ODES.

a.

With the struetural integrity of any A.ME C.d

.Class 1: comp.n.nt not conforming to the ab.v. requirements, restere the structural integrity.f the.af.

fterd mpn.nt(s) t. within its limit or isolate the affected compenent(s) prier to increasing the Reactor Coolant System temperature mere than 50OF abevc the minimum tempearture required by NDT consideartions.

b.

With the structural integrity of any ASME Code Class 2 compenent(s)-

+

not conforming to the ab.v. requirements, re.t.r-the stru..tural integrity of the affected e.mpon.nt(s) to within its limit or icolate theaof fected component(s) prior to increasing the Reactor Coolant System temperature abeve 2000 P.

e.With the structural integrity of any ASME Coda Class 3 eempenent(s) not conforming to the above requirements, ractora the structural integrity of the affeet.d mp.n.nt

) to within its limit or isolate the affected cornpenent(s) fromn seviee.

Relocated to TS 6.8.4.k SCUVBhTEI LN3ACE REQUIREMENTS 4.n.1.

laddition to tha requirements of Cpacifi*

SCoolant Pump flywheel shall be inspected per the recomm Politimn e

.4.b o

f Regulteory Guide 1.14, Rtviioen 1, Au Position C.4.b(i) and C.4.b(2), a qualified in ploae UT 3vlfefrom the inner bora of the flywheel to the circi p 1= i ýz endat gust eiEafi e eBE n 4.0.5, eaah Reactor tions of Regulatory 1975.

In lieu of

~inatien ever the half of the out-ry radius or a curfaca a2xaminatien (qT-and/oer PT-) of axpocad cuirfaeac o~f the ramevad fl~~aal ma bacondcta at20 aarintervals.

1.4.11.2 Augmented inserviee inspactioen Programn for Steam Canarator Channal Heade The No. 21: Steam Caneartor channal head shall be ultraconicolly insipacted in a calactad area during eaeh of the first three rafuaeling outagec using the coa~i ultrasonic inspection procadurac and equipment used to generate the baceline data These incarvica ult-rasonic inspections shall verify that the croaks observad in the stainless steel cladding prior to epeartien hv not propagatad into the boas material.

/

SALEM - UNIT 2 3/4 4-33 Amendment No. 258

ADMINISTRATIVE CONTROLS Note: A proposed change adding 6.8.4.j (and Page 6-19g) is pending via PSEG LAR S09-06, submitted September 23, 2009.

6.8.4.k Reactor Coolant Pump Flywheel Inspection Program In addition to the requirements of the ISI Program, 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(1) 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.

SALEM -

UNIT 2 6-19g Amendment No.

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4.6 PRESSURE/TEMPERATURE LIMITS Reactor Coolant System..................................

3/4 4-21 Figure 3.4.6.1-1 Hydrostatic Pressure and Leak Tests Pressure/Temperature Limits -

Curve A 3/4 4-23 Figure 3.4.6.1-2 Non-Nuclear Heatup and Cooldown Pressure/Temperature Limits -

Curve B 3/4 4-23a Figure 3.4.6.1-3 Core Critical Heatup and Cooldown Pressure/Temperature Limits -

Curve C 3/4 4-23b Table 4.4.6.1.3-1 (Deleted)...........................

3/4 4-24 Reactor Steam Dome......................................

3/4 4-25 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES........................

3/4 4-26 3/4.4.8 DELETED STRUCTURAL INTECRITY............................

3/4 4-27 3/4.4.9 RESIDUAL HEAT REMOVAL Hot Shutdown............................................

3/4 4-28 Cold Shutdown...........................................

3/4 4-29 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ECCS -

OPERATING........................................

3/4 5-1 3/4.5.2 ECCS -

SHUTDOWN.........................................

3/4 5-6 3/4.5.3 SUPPRESSION CHAMBER.....................................

3/4 5-8 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Primary Containment Integrity...........................

3/4 6-1 Primary Containment Leakage.............................

3/4 6-2 Primary Containment Air Locks...........................

3/4 6-5 Primary Containment Structural Integrity...................

3/4 6-8 Drywell and Suppression Chamber Internal Pressure.......

3/4 6-9 HOPE CREEK xi Amendment No.

1-3-4

INDEX BASES SECTION 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES...........

3/4.4.8 DELETED STRUCTURAL INTEC-RITY...............

3/4.4.9 RESIDUAL HEAT REMOVAL......................

3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1/2 ECCS - OPERATING and SHUTDOWN..............

3/4.5.3 SUPPRESSION CHAMBER........................

3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Primary Containment Integrity..............

Primary Containment Leakage................

Primary Containment Air Locks..............

Primary Containment Structural Integrity...

Drywell and Suppression Chamber Internal Pressure Drywell Average Air Temperature............

Drywell and Suppression Chamber Purge System 3/4.6.2 DEPRESSURIZATION SYSTEMS...................

3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES.......

3/4.6.4 VACUUM RELIEF..............................

3/4.6.5 SECONDARY CONTAINMENT......................

3/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROL.....

PAGE B 3/4 B 3/4 B 3/4 4-6 4-6 4-6 B 3/4 5-1 B 3/4 5-3 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 6-1 6-1 6-1 6-2 6-2 6-2 6-2 6-3 6-5 6-5 6-13 6-14 Amendment No.

34 HOPE CREEK xix

REACTOR COOLANT SYSTEM 3/4.4.8 DELETED STRUCTURAL INTECRITY LIMI1TING CONDIITION FOR OPERATIO4N 3.4.8 The struetural integrity of AGME Cede Class 1, 2 and 3 eemponents shall be maintained in aeeordaneo with Speeifieation 4.4.8.

APPLIGABILITY. OPERATIONAL CONDITIONS 1, 2, 3, 4 and S.

a.

With the "tru.tural integrity vf any A'ME Cede Glass llemponent(,)

noi=etscn-forming to the above roguirefmonts, restere the struetural integrity of the affected component(.)

t. within its limnit or i.lat.

the af feeted eemponent(s) prier to inereasing the Reaeter Coolant System temperature mare than 50 0 F above the minimum temperature required by NDT eonsiderations.

b.

With the struetural integrity of any A-ME CGd* Class 2

-,mpen.nt(s) net eonforming to the aboro requirements, rootere the struetural integrity of the affoctod eempenent(s) to within its limnit or isolate the affoatod eefmpenent(s) prior to inereaaing the Roaetor Coolant System temperature abevo 2000 F-.

e. With the struetural integrity of any ASME Coda Class 3 eempenent(s) not conorm+ing to the above roquirmoRnt integrity of the aff..t.d eomponont(s) the affected component(.)

from.....

s, rootoro the str-ueturai-to within its limit or isolato SURZVEILLANCE REQUIREMENTS

• '*TTT')ST'*TT T 7*TI"I*

T)'*/*TTT*[*'*?.T'*&Tri*('* -- L 1.

.o ti oj reguiromento ether-tanon bpOOL100QL HOPE CREEK 3/4 4-27 Amendment No.----

.LR-N10-0032 TECHNICAL SPECIFICATION BASES PAGES WITH PROPOSED CHANGES (Provided for INFORMATION ONLY)

REACTOR COOLANT SYSTEM BASES 3/4.4.10 DELETED STRUCTURAL INTECRITY Tha insptatien pr.gramzs f-r A.ME Cod. Glass 1, 2 anad 3 T

empenent ansaure that the struetural integrity of these eempenents will be mRaintained at an aeeeptable level througheut the life af the plant. Te the extent applieable, the inspeetien program for these eempenents is in eomplianee with Saption X!

Af tFht

?APMRP2,i 1,:r-Anci 1TFrmi

-' \\h qi1 Pnfit'-

3/4.4.11 THIS SECTION INTENTIONALLY BLANK 3/4.4.12 REACTOR VESSEL HEAD VENTS Reactor Coolant System vents are provided to exhaust noncondensible gases and/or steam from the Reactor Coolant System that could inhibit natural circulation core cooling. The OPERABILITY of a reactor vessel head vent path ensures the capability exists to perform this function.

The valve redundancy of the Reactor Coolant System vent paths serves to minimize the probability of inadvertent or irreversible actuation while ensuring that a single failure in a vent valve power supply or control system does not prevent isolation of the vent path.

The function, capabilities, and testing requirements of the.Reactor Coolant System Vent Systems are consistent with the requirements of Item II.BI of NUREG-0737, "Clarification of TMI Action Plan Requirements,"

November 1980.

Correction lettr dated February 15, 1990, to Amendment 108 dated January 29, 1990.

SALEM UNIT 1 B 3/4 4-17 Amendment No.

-08

REACTOR COOLANT SYSTEM BASES 3/4.4.11 DELETED I The i.n..rv..i insp

.ti.n and testing prgraf frr A*MnE Cede Class 1, 2 and 3 eempenents ensure that the struettural integrity and eperatienal readinbop of these

.mp.n.nts will be maintained at an accaptable lev.l through the life of the plant. These programas are in aeeerdaiiee with Saction X! of the ASME Beollr and Pressure Vessel Cede and applieable Addenda as required by 10 CFR Part 50.55a(g) axcept whe.r sp.. ifi. written relief has been granted by the Comm...issin pursuant to 10 CFR

• ....a(g)()(i)c 3/4.4.12 REACTOR VESSEL HEAD VENTS Reactor Coolant System vents are provided to exhaust noncondensible gases and/or steam from the Reactor Coolant System that could inhibit natural circulation core cooling. The OPERABILITY of a reactor vessel head vent path ensures the capability exists to perform this function.

The valve redundancy of the Reactor Coolant System vent paths serves to minimize the probability of inadvertent or irreversible actuation while ensuring that a single failure vent in a valve power supply or control system does not prevent isolation of the vent path.

The function, capabilities, and testing requirements of the Reactor Coolant System Vent Systems are consistent with the requirements of Item II.B.I of NUREG-0737, "Clarification of TMI Action Plant Requirements,"

November 1980.

Correction letter dated February 15, 1990, to Amendment 86 dated January 29, 1990.

SALEM UNIT 2 B 3/4 4-18 Amendment No.843

REACTOR COOLANT SYSTEM BASES PRESSURE/TEMPERATURE LIMITS (Continued)

The pressure-temperature limit lines shown in Figures 3.4.6.1-1 and 3.4.6.1-3, curves for inservice leak and hydrostatic testing and reactor criticality have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR Part 50 for reactor criticality and for inservice leak and hydrostatic testing.

The number of reactor vessel irradiation surveillance capsules and the frequencies for removing and testing the specimens in these capsules are provided in UFSAR Section 5.3 and Appendix 5A.

3/4.4.7 MAIN STEAM LINE ISOLATION VALVES Double isolation valves are provided on each of the main steam lines to minimize the potential leakage paths from the containment in case of a line break.

Only one valve in each line is required to maintain the integrity of the containment, however, single failure considerations require that two valves be OPERABLE.

The surveillance requirements are based on the operating history of this type valve.

The maximum closure time has been selected to contain fission products and to ensure the core is not uncovered following line breaks.

The minimum closure time is consistent with the assumptions in the safety analyses to prevent pressure surges.

3/4.4.8 DELETED STRUCTURAL INTECRITY The inspcction programo for ASME Cde CGlacs !,

2 and 3 componants ensure that tha

.tru.tural integrity of these e.mpn.ant. will be maintained at an aeeaptable level througheut the life of the plant.

Compenents of the reacter eoolant system were designed to previde aeeaoo to permit incarviea inspactiens in aeeordanee with Sectien X! of the ASME Be-iler and Pressure Vessel Coda 1977 Editio n and Addenda through Summer 1978.

The insarvica inopactien programn for ASME Cede Class 1, 2 and 3 eempenents will be parformad in accordanca with Sactien X! of the ASMS Beiler and Pressure Vessel Cede and applieabla addenda as reauired by 10 CER Part S0.55a(g) e)Eeept where speeific written relief has been granted by the NTRC pursuant to 10 CFR Part 50.55a(g) (6) (i:)

3/4.4.9 RESIDUAL HEAT REMOVAL A single shutdown cooling mode loop provides sufficient heat removal capability for removing core decay heat and mixing to assure accurate temperature indication, however, single failure considerations require that two loops be OPERABLE or that alternate methods capable of decay heat removal be demonstrated and that an alternate method of coolant mixing be in operation.

HOPE CREEK B 3/4 4-6 Amendment No.

-&&