Text

License Amendment Request to Revise Technical Specifications to Adopt TSTF-545, Revision 3, TS Lnservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing Using the Consolidated Line .
	ML16243A233
Person / Time
Site:	Salem
Issue date:	08/30/2016
From:	Carr E; Public Service Enterprise Group
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	LAR S16-05, LR-N16-0114
	Download: ML16243A233 (45)
	v • d • e

PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, NJ 08038-0236 AUG 3 0 2016 PSEG NuclearLLC 10 CFR 50.90 10 CFR 50.55a LR-N16-0114 LAR S16-05 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Salem Generating Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-70 and DPR-75 NRC Docket Nos. 50-272 and 50-311

Subject:

License Amendment Request to Revise Technical Specifications to Adopt TSTF-545, Revision 3, "TS lnservice n~sting Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing" Using the Consolidated Line Item Improvement Process and to Request an Alternative to the ASME Code Pursuant to 10 CFR 50.90, PSEG Nuclear LLC (PSEG) is submitting a request for an amendment to the Technical Specifications (TS) to the renewed facility operating licenses listed above.

The proposed change revises the TS to eliminate TS 6.8.4.j, "lnservice Testing Program." A new defined term, "lnservice Testing Program (1ST)," is added toTS Definitions section. This request is consistent with TSTF-545, Revision 3, "TS lnservice Testing Program Removal &

Clarify SR (Surveillance Requirement) Usage Rule Application to Section 5.5 Testing."

Pursuant to 10 CFR 50.55a(z), the application also proposes an alternative to the testing frequencies in the American Society of Mechanical Engineers (ASME) Operation and Maintenance (OM) Code, by adoption of approved Code Case OMN-20, "lnservice Test

P:reEjt;~eRey-;-feF-tF!e~Gt;~FFeRt-1Q~year-IRserviee-l=estiA§~tiS=r-)~IAterv-al1-.- - - - - - - - - - - - - - - - - - - ' - -

Attachment 1 provides a description and assessment of the proposed TS changes.

Attachment 2 provides the existing TS pages marked up to show the proposed changes.

Attachment 3 provides the Relief Request G-01 for an alternative to the A'SME Code.

Approval of the proposed amendment and relief request is requested within one year of submittal. Once approved, the amendment shall be implemented within 60 days.

AUG 3 0 2016 LR-N16-0114 10 CFR 50.90 Page 2 10 CFR 50.55a In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated State of New Jersey Official.

There are no regulatory commitments contained in this letter.

If you have any questions regarding this submittal, please contact Ms. Tanya Timberman at 856-339-1426.

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

Executed on _ _ 2?-..~--:/_;;7_(}:::..L_0-1./_.::6:::.___

(Date)

Respectfully,

-*~

Eric Carr

/~-----**

-~-

Acting Site Vice President Salem Generating Station Attachments:

1. Description and Assessment of Technical Specification Changes

2. Proposed Technical Specification Changes (Mark-Up)

3. Description and Assessment of the Proposed Alternative to the ASME Code cc: Mr. D. Dorman, Administrator, Region I, NRC Ms. C. Parker, Project Manager, NRC NRC Senior Resident Inspector, Salem Mr. P. Mulligan, Chief, NJBNE PSEG Corporate Commitment Tracking Coordinator Salem Commitment Tracking Coordinator

LR-N16-0114 LAR S16-05 Attachment 1 Description and Assessment of Technical Specification Changes Table of Contents

1.0 DESCRIPTION

................................................................................................................1 2.0 ASSESSMENT ................................................................................................................1 2.1 Applicability of Published Safety Evaluation .............................................................1 2.2 Variations ................................................................................................................1

3.0 REGULATORY ANALYSIS

.............................................................................................3 3.1 No Significant Hazards Consideration Analysis .......................................................3 4.0 ENVIRONMENTAL EVALUATION ..................................................................................4

LR-N16-0114 LAR S16-05

1.0 DESCRIPTION

The proposed change eliminates the Technical Specifications (TS), Section 6.8.4.j, "Inservice Test (IST) Program," to remove requirements duplicated in American Society of Mechanical Engineers (ASME) Code for Operations and Maintenance of Nuclear Power Plants (OM Code),

Case OMN-20, "Inservice Test Frequency." A new defined term, "Inservice Testing Program,"

is added to TS Section 1.0, "Definitions. The proposed change to the TS is consistent with TSTF-545, Revision 3, "TS Inservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing."

2.0 ASSESSMENT 2.1 Applicability of Published Safety Evaluation PSEG has reviewed the model safety evaluation provided to the Technical Specifications Task Force in a letter dated December 11, 2015 (NRC ADAMS Accession No. ML15314A365). This review included a review of the NRC staffs evaluation, as well as the information provided in TSTF-545. PSEG concluded that the justifications presented in TSTF-545, and the model safety evaluation prepared by the NRC staff are applicable to Salem Generating Station Units 1 and 2 (Salem Units 1 and 2) and justify this amendment for the incorporation of the changes to the Salem Units 1 and 2 TS.

Salem Unit 1 was issued a construction permit, CPPR-52, on September 25, 1968 and the provisions of 10 CFR 50.55a(f)(1) are applicable.

Salem Unit 2 was issued a construction permit, CPPR-53, on September 25, 1968 and the provisions of 10 CFR 50.55a(f)(1) are applicable.

2.2 Variations PSEG is not proposing any variations or deviations from the TS changes described in the TSTF-545 or the applicable parts of the NRC staffs model safety evaluation dated December 11, 2015.

The Salem Units 1 and 2 TS utilize different numbering than the Standard Technical Specifications on which TSTF-545 was based. Specifically, the "Inservice Testing Program" in the Salem Units 1 and 2 TS is numbered 6.8.4.j rather than 5.5.7. The "Definitions" in the Salem Units 1 and 2 TS is numbered 1.0 rather than 1.1. Also, SR 3.0.2 and SR 3.0.3 are TS 4.0.2 and TS 4.0.3 in the Salem Units 1 and 2 TS.

As noted in Section 2.2.1 of the TSTF Traveler, the phrase "Inservice Testing Program" may appear in different locations in plant-specific TS. Revising this phrase to be capitalized wherever it may appear is within the scope of this proposed change. The SR listed in the Standard Technical Specifications differs from the SR listed in the Salem Units 1 and 2 TS as follows:

1

LR-N16-0114 LAR S16-05 Standard Technical Specifications Salem Unit 1 TS --- 4.1.2.3, Charging Pump - Shutdown

--- 4.1.2.4, Charging Pump - Operating 3.4.10.1, Pressurizer Safety Valves 4.4.2.1, Safety Valves - Shutdown 3.4.14.1, RCS PIV Leakage ---

--- 4.4.2.2, Safety Valves - Operating

--- 4.4.3.1, Relief Valves

--- 4.4.6.3, Primary Coolant System Pressure Isolation Valves

--- 4.4.9.3.1.d, Overpressure Protection Systems 3.5.2.4, ECCS Operating 4.5.2.f, ECCS Subsystems 3.6.3.5, Containment Isolation Valves 4.6.3.1.4, Containment Isolation Valves 3.6.6A.4, Containment Spray and Cooling Systems 4.6.2.1.b, Containment Spray System 3.6.12, Vacuum Relief Valves ---

3.7.1.1, MSSVs 4.7.1.1, Turbine Cycle, Safety Valves 3.7.2.1, MSIVs 4.7.1.5, Main Steam Line Isolation Valves 3.7.3.1, MFIVs and MFRVs ---

3.7.5.2, AFW System ---

--- 4.9.8.2, Low Water Level Standard Technical Specifications Salem Unit 2 TS --- 4.1.2.3, Charging Pump - Shutdown

--- 4.1.2.4, Charging Pump - Operating 3.4.10.1, Pressurizer Safety Valves 4.4.2.1, Safety Valves - Shutdown 3.4.14.1, RCS PIV Leakage ---

--- 4.4.3, Safety Valves - Operating

--- 4.4.5.1, Relief Valves

--- 4.4.7.2.2, Primary Coolant System Pressure Isolation Valves

--- 4.4.10.3.1.d, Overpressure Protection Systems 3.5.2.4, ECCS Operating 4.5.2.f, ECCS Subsystems 3.6.3.5, Containment Isolation Valves 4.6.3.4, Containment Isolation Valves 3.6.6A.4, Containment Spray and Cooling Systems 4.6.2.1.b, Containment Spray System 3.6.12.1, Vacuum Relief Valves ---

3.7.1.1, MSSVs 4.7.1.1, Turbine Cycle, Safety Valves 3.7.2.1, MSIVs 4.7.1.5, Main Steam Line Isolation Valves 3.7.3.1, MFIVs and MFRVs ---

3.7.5.2, AFW System ---

--- 4.9.8.2, Low Water Level Finally, TSTF-545 renumbers the Standard Technical Specifications sections to reflect elimination of the program and references to the renumbered are revised, whereas PSEG is keeping the existing numbering after deleting TS section 6.8.4.j.

These differences are administrative and do not affect the applicability of TSTF-545 to the Salem Units 1 and 2 TS.

2

LR-N16-0114 LAR S16-05

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration PSEG requests adoption of the Technical Specification (TS) changes described in TSTF-545, "TS Inservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing," which is an approved change to the Improved Standard Technical Specifications (ISTS), into the Salem Units 1 and 2 TS. The proposed change revises the TS Chapter 6, "Administrative Controls," Section 6.8, "Procedures and Programs," to delete the "Inservice Testing (IST) Program" specification. Requirements in the IST Program are removed, as they are duplicative of requirements in the American Society of Mechanical Engineers (ASME)

Operations and Maintenance (OM) Code, as clarified by Code Case OMN-20, "Inservice Test Frequency." Other requirements in Section 6.8 are eliminated because the Nuclear Regulatory Commission (NRC) has determined their appearance in the TS is contrary to regulations. A new defined term, "Inservice Testing Program," is added, which references the requirements of Title 10 of the Code of Federal Regulations (10 CFR), Part 50, paragraph 50.55a(f). 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 change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed change revises TS Chapter 6, "Administrative Controls," Section 6.8, "Procedures and Programs," by eliminating the "Inservice Testing Program" specification.

Most requirements in the Inservice Testing Program are removed, as they are duplicative of requirements in the ASME OM Code, as clarified by Code Case OMN-20, "Inservice Test Frequency." The remaining requirements in the Section 6.8 IST Program are eliminated because the NRC has determined their inclusion in the TS is contrary to regulations. A new defined term, "Inservice Testing Program," is added to the TS, which references the requirements of 10 CFR 50.55a(f).

Performance of inservice testing is not an initiator to any accident previously evaluated. As a result, the probability of occurrence of an accident is not significantly affected by the proposed change. Inservice test frequencies under Code Case OMN-20 are equivalent to the current testing period allowed by the TS with the exception that testing frequencies greater than 2 years may be extended by up to 6 months to facilitate test scheduling and consideration of plant operating conditions that may not be suitable for performance of the required testing. The testing frequency extension will not affect the ability of the components to mitigate any accident previously evaluated as the components are required to be operable during the testing period extension. Performance of inservice tests utilizing the allowances in OMN-20 will not significantly affect the reliability of the tested components.

As a result, the availability of the affected components, as well as their ability to mitigate the consequences of accidents previously evaluated, is not affected.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

3

LR-N16-0114 LAR S16-05

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 change does not alter the design or configuration of the plant. The proposed change does not involve a physical alteration of the plant; no new or different kind of equipment will be installed. The proposed change does not alter the types of inservice testing performed. In most cases, the frequency of inservice testing is unchanged.

However, the frequency of testing would not result in a new or different kind of accident from any previously evaluated since the testing methods are not altered.

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

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

Response: No The proposed change eliminates some requirements from the TS in lieu of requirements in the ASME Code, as modified by use of Code Case OMN-20. Compliance with the ASME Code is required by 10 CFR 50.55a. The proposed change also allows inservice tests with frequencies greater than 2 years to be extended by 6 months to facilitate test scheduling and consideration of plant operating conditions that may not be suitable for performance of the required testing. The testing frequency extension will not affect the ability of the components to respond to an accident as the components are required to be operable during the testing period extension. The proposed change will eliminate the existing TS 4.0.3 allowance to defer performance of missed inservice tests up to the duration of the specified testing frequency, and instead will require an assessment of the missed test on equipment operability. This assessment will consider the effect on a margin of safety (equipment operability). Should the component be inoperable, the TS provide actions to ensure that the margin of safety is protected. The proposed change also eliminates a statement that nothing in the ASME Code should be construed to supersede the requirements of any TS. The NRC has determined that statement to be incorrect. However, elimination of the statement will have no effect on plant operation or safety.

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

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

4.0 ENVIRONMENTAL EVALUATION The proposed change 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 change does not involve (i) a 4

LR-N16-0114 LAR S16-05 significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change 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 change.

5

LR-N16-0114 LAR S16-05 Attachment 2 Mark-up of Proposed Technical Specification Pages The following Technical Specifications pages for Renewed Facility Operating License DPR-70 are affected by this change request:

Technical Specification Page Index I 1.0, Definitions 1-4 4.1.2.3, Charging Pump - Shutdown 3/4 1-10 4.1.2.4, Charging Pump - Operating 3/4 1-11 4.4.2.1, Safety Valves - Shutdown 3/4 4-4 4.4.2.2, Safety Valves - Operating 3/4 4-4a 4.4.3.1, Relief Valves 3/4 4-5a 4.4.6.3, Primary Coolant System Pressure Isolation Valves 3/4 4-16a 4.4.9.3.1.d, Overpressure Protection Systems 3/4 4-31 4.5.2.f, ECCS Subsystems 3/4 5-5a 4.6.2.1.b, Containment Spray System 3/4 6-9 4.6.3.1.4, Containment Isolation Valves 3/4 6-13 4.7.1.1, Turbine Cycle, Safety Valves 3/4 7-1 4.7.1.5, Main Steam Line Isolation Valves 3/4 7-10 4.9.8.2, Low Water Level 3/4 9-8a 6.8.4.j, Inservice Testing Program 6-19e The following Technical Specifications pages for Renewed Facility Operating License DPR-75 are affected by this change request:

Technical Specification Page Index I 1.0, Definitions 1-4 4.1.2.3, Charging Pump - Shutdown 3/4 1-9 4.1.2.4, Charging Pump - Operating 3/4 1-10 4.4.2.1, Safety Valves - Shutdown 3/4 4-5 4.4.3, Safety Valves - Operating 3/4 4-6 4.4.5.1, Relief Valves 3/4 4-8a 4.4.7.2.2, Primary Coolant System Pressure Isolation Valves 3/4 4-18 4.4.10.3.1.d, Overpressure Protection Systems 3/4 4-32 4.5.2.f, ECCS Subsystems 3/4 5-6 4.6.2.1.b, Containment Spray System 3/4 6-10 4.6.3.4, Containment Isolation Valves 3/4 6-15 4.7.1.1, Turbine Cycle, Safety Valves 3/4 7-1 4.7.1.5, Main Steam Line Isolation Valves 3/4 7-10 4.9.8.2, Low Water Level 3/4 9-9 6.8.4.j, Inservice Testing Program 6-19f

DEFINITIONS SECTION 1.0 DEFINITIONS DEFINED TERMS ........*.. , . . . . . . . . . . . . . . . . . . . , . , , .. , .. , ... , , . . . . . l-1 ACTION . . . . . . . . . . . . . . . . . . . . . . . . . * . . . . . . . . . . , .....*. , . * . . . . . . . . . . . l-l AXIAL FLUX DIFFERENCE . . . . * . . . . . . . . . * . . . . . . . . * . . . * . . . . . . . . . . . . . . . 1-1 CHANNEL CALIBRATION . . . . . . . . . . . . . . . . . * . . . . . . . . . * . * . . * . * . . * . . . . . . . 1-1 CHANNEL CHECK . . . . . . . . . . . . . . . . * . . . . . . . . * . . , .*.... , ..... , ... , , . . . . 1-1 CHANNEL FUNCTIONAL TEST......................................... 1-1 CONTAINMENT INTEGRITY . . . . . . . . . . . . . . . * * . . . . . . * . * . * . . . . . . . . . . . . . . . 1-2 CORE ALTERATION . . . . . . . . . . . . . . . * . . . * . . . . . . . . . * . . * . . . . . . . . * . . . . . . . 1-2 CORE OPERATING LIMITS REPORT .....****......**... **.*... ..*..... 1-2 QOSE EQUIVALENT I-131 ..............*.***. , ..... , ...*.. , . , ..... , . 1-2 E-AVERAGE DISINTEGRATION ENERGY................................. 1-3 ENGINEERED SAFETY FEATURE RESPONSE TIME......................... 1-3 FREQUENCY NOTATION . . . . * . . . . . . . . . . . . * * * * . . . * . * . * . * . . * . . . . . . * . . . . . 1-3 FULLY WITHDRAWN * * . * * * * * * * * . * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * *

1-3 GASEOUS RADWASTE TREATMENT SYSTEM............................... 1-3 IDENTIFIED LEAKAGE * * . * * . . * . * * * * * * * * * * * * . . . * * * . . . * * * . * * . * * * * . * . *

1-3 MEMBER (S) OF THE PUBLIC . . . . . * * * . . . . . * . . . . . . * * . . * * . . . . . . . . . * . . . . . 1-4 OFFSITE DOSE CALCULATION MANUAL (ODCM) ......*.... . . . . . . . . . . . . . . . 1-4 OPERABLE - OPERABILITY . . . . . . . . . . * . * * . . . . . . . . . . . . . . . * . . . . . . * . . . . . 1-4 OPERATIONAL MODE - MODE . . . . * . . . . . . . . . * . . . . . . * . . . . . . . . . * . . . . . . . . . 1-4 PHYSICS TESTS . . . . . . . . . . . . . . . . . . . * . . . . . . . . . . . . . * . . . * . . . . . . . . . . . . . 1-5 PRESSURE BOUNDARY LEAKAGE....................................... 1-5 PROCESS CONTROL PROGRAM (PCP)................................... 1-5 PURGE-PURGING................................................... 1-5 QUADRANT POWER TILT RATIO . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 RATED THERMAL POWER............................................. 1-5 REACTOR TRIP SYSTEM RESPONSE TIME . . . . * . . . . . . . . . . . . . . . . . . . . . . . . . . 1-6 REPORTABLE EVENT . . . . . . . . . . . . . . . . . . . * . . . . . . . * . . . . . . . . . . . . . . . . . . . . 1-6 SHUTDOWN MARGIN................................................. 1-6 SITE BOUNDARY . . . . . . . . . . . . . . . . . . . * . . * . . . . . . . . . . . . . . * . . . . . . . . . . . . . 1-6 SOLIDIFICATION.................................................. l-6 SOURCE CHECK . . . . . . . . . . . . . . . . . . * . * . . . . . . . . . . . . . . . . . . * . . . . . . . . . . . . 1-6 STAGGERED TEST BASIS . . . . . . . . . . . . . . . . . * . . * . . . . . . . . . . * . . . . . . * . . . . . l-6 THERMAL POWER . . . * . . . . . * . . . * . . . . . . . * . . * * * . . . . . . . * . . . . * . . . . . . . . * . . 1-7 UNIDENTIFIED LEAKAGE ****..*..*.*.*********.*.**.*.*********.*...

1-7 UNRESTRICTED AREA . . * . * * . . * * . . * * * . * * * * * * * * * * * . * * * * * * * * * * . . * * * * * . . 1-7 VENTILATION EXHAUST TREATMENT SYSTEM............................ 1-7 VENTING . . . . . . . . . . . * . . . * . . * * . . . . . . . . . . . . . . . . . . . . * . . * . . . . . . . . . . . . . 1-7 INSERVICE TESTING PROGRAM .............................................................. 1-4 SALEM - UNIT 1 I Amendment No.225

INSERVICE TESTING PROGRAM 1.15.1 The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f).

tlEFINITIONS

b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE, or

c. Reactor coolant system leakage through a steam generator to the secondary system (primary-to-secondary leakage) .

MEMBER(S) OF THE PUBLIC 1.16 MEMBER(S) OF THE PUBLIC shall be all those persons who are not occupationally associated with the plant. This category does not include employees of PSE&G, its contractors, or vendors. Also excluded from this category are perSons Who enter the site to service equipment or to make deliveries. Thi~ category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.

OFFSITE DOSE CALCULATION MANUAL (ODCM) 1.17 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm/Trip setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The'ODCM shall also contain (1) the Radioactive Effluent controls and Radiological Environmental Monitoring programs required by Section 6.8.4 and (2) descriptions of the information that .should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Specifications 6.9.1.7 and 6.9.1.8 respectively.

OPERABLE - OPERABILITY 1.18 A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency eleCtrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

OPERATIONAL MODE - MODE 1.19 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.1.

SALEM - UNIT 1 1-4 Amendment No. 268

REACTIVITY CONTROL SYSTEMS CHARGING PUMP- SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.3 At least one charging pump in the boron injection flow path required by Specification 3.1.2.1 shall be OPERABLE.#

APPLICABILITY: MODES 4, 5 and 6.

ACTION:

With no charging pump OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until one charging pump is restored to OPERABLE status.

SURVEILLANCE REQUIREMENTS 4.1.2.3 No additional Surveillance Requirements other than those required by the lnservice Testing Program. INSERVICE TESTING PROGRAM

A maximum of one centrifugal charging pump shall be OPERABLE while in MODE 4 when the temperature of one or more of the RCS cold legs is less than or equal to 312°F, MODE 5, or MODE 6 when the head is on the reactor vessel.

SALEM- UNIT 1 3/41-10 Amendment No. 297

REACTIVITY CONTROL SYSTEMS CHARGING PUMPS- OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two charging pumps shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

With only one charging pump OPERABLE, restore at least two charging pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1% bk/k at 200°F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two charging pumps to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.1.2.4 No additional Surveillance Requirements other than those required by the lnservice Testing Program. INSERVICE TESTING PROGRAM SALEM - UNIT 1 3/4 1-11 Amendment No. 297.

REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY VALVES SAFETY VALVES- SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.2.1 A minimum of one pressurizer code safety valve shall be OPERABLE* with a lift setting of 2485 psig +/- 3%.**,***

APPLICABILITY: MODE 4 and 5 ACTION:

With no pressurizer code safety valve OPERABLE, immediately suspend all operations involving positive reactivity changes and place an OPERABLE RHR loop into operation in the shutdown cooling mode.

SURVEILLANCE REQUIREMENTS 4.4.2.1 No additional Surveillance Requirements other than those required by the lnservice Testing Program. INSERVICE TESTING PROGRAM

While in Mode 5, an e"quivalent size vent pathway may be used provided that the vent pathway is not isolated or sealed.

- The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

- - Following testing the lift setting shall be reset to within +/- 1%.

SALEM - UNIT 1 3/4 4-4 Amendment No. 297

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY VALVES SAFETY VALVES- OPERATING LIMITING CONDITION FOR OPERATION 3.4.2.2 All pressurizer code safety valve shall be OPERABLE with a lift setting of 2485 psig +/- 3%.*,**

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

With one pressurizer code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes, or be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS 4.4.2.2 No additional Surveillance Requirements other than those required by the lnservice Testing Program. INSERVICE TESTING PROGRAM

The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure. 1

- Following testing the lift setting shall be reset to within +/- 1%.

SALEM- UNIT 1 3/4 4-4a Amendment No. 297

REACTOR COOLANT SYSTEM 3/4.4.3 RELIEF VALVES SURVEILLANCE REQUIREMENTS INSERVICE TESTING PROGRAM 4.4.3.1 In addition to the requirements of the lnservice Testing Program, each PORV shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by:

a. Operating the PORV through one complete cycle of full travel during MODES 3 or 4, and

b. Operating solenoid valves, air control valves, and check valves on associated air accumulators in PORV control systems through one complete cycle of full travel, and

c. Performing a CHANNEL CALIBRATION of the actuation instrumentation.

4.4.3.2 Each block valve shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by operating the valve through one complete cycle of full travel unless the block valve is closed in order to meet the requirements of ACTION b, or c in Specification 3.4.3.

SALEM- UNIT 1 3/4 4-5a Amendment No. 299

REACTOR COOLANT SYSTEM PRIMARY COOLANT SYSTEM PRESSURE ISOLATION VALVES LIMITING CONDITION FQR OPERATION 3.4.6.3 Reactor Coolant System Pressure Isolation Valves specified in table 4.4-3 shall be OPERABLE.

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

ACTION:

a. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the specified limit in Table 4.4-3, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least two closed manual or deactivated automatic valves, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 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 4.4.6.3 Each Reactor Coolant System Pressure Isolation Valve specified in Table 4.4-3 shall be demonstrated OPERABLE pursuant to the lnservice Testing Program, except that in lieu of any leakage testing required by the lnservice Testing Program, each valve shall be demonstrated OPERABLE by verifying leakage to be within its limit: INSERVICE TESTING INSERVICE PROGRAM

a. In accordance with the Surveillance Frequency Control Program.

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

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

d. For the Residual Heat Removal and Safety Injection Systems hot and cold leg injection valves and accumulator valves listed in Table 4.4-3 the testing will be done within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or flow through the valve. For all other systems testing will be done once per refueling.

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

SALEM- UNIT 1 3/4 4-16a Amendment No. 299

REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS SURVEILLANCE REQUIREMENTS 4.4.9.3.1 Each POPS shall be demonstrated OPERABLE by:

a. Performance of a CHANNEL FUNCTIONAL TEST on the POPS actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which the POPS is required OPERABLE, and in accordance with the Surveillance Frequency Control Program thereafter when the POPS is required OPERABLE.

b. Performance of a CHANNEL CALIBRATION on the POPS actuation channel in accordance with the Surveillance Frequency Control Program.

c. Verifying the POPS isolation valve is open in accordance with the Surveillance Frequency Control Program when the POPS is being used for overpressure protection.

d. Testing pursuant to the lnservice Testing Program. INSERVICE TESTING PROGRAM 4.4.9.3.2 The RCS vent(s) shall be verified to be open in accordance with the Surveillance Frequency Control Program* when the vents(s) is being used for overpressure protection.

Except when the vent pathway is provided with a valve which is locked, sealed, or otherwise secured in the open position, then verify these valves open in accordance with the Surveillance Frequency Control Program.

SALEM - UNIT 1 3/4 4-31 Amendment No. 299

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

f. By verifying that each of the following pumps develops the indicated Total Dynamic Head (TDH) when tested at the test flow point pursuant to the Inservice Testing Program: INSERVICE TESTING PROGRAM

1. Centrifugal charging pump ~ 2338 psi TDH

2. Safety Injection Pump ~ 1369 psi TDH

3. Residual heat removal pump ~ 165 psi TDH

g. By verifying the correct position of each of the following ECCS throttle valves:

1. Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following completion of each valve stroking operation or maintenance on the valve when the ECCS subsystems are required to be OPERABLE.

2. In accordance with the Surveillance Frequency Control Program.

HPSISYSTEM LPSI SYSTEM VALVE NUMBER VALVE NUMBER 11 SJ 16 11 SJ 138 12 SJ 16 12 SJ 138 13 SJ 16 13 SJ 138 14 SJ 16 14 SJ 138 11 SJ 143 12 SJ 143 13 SJ 143 14 SJ 143

h. By performing a flow balance test, during shutdown, following completion of modifications to the ECCS subsystems that alter the subsystem flow characteristics and verifying that:

1. For Safety Injection pumps, with a single pump running:

a) The sum of the injection line flow rates, excluding the highest flow rate, is ~ 453 gpm; and b) The total flow rate through all four injection lines is s 647 gpm, and c) The difference between any pair of injection line flow rates is s 12.0 gpm, and d) The total pump flow rate is ::; 664 gpm in the cold leg alignment, and e) The total pump flow rate is s 654 gpm in the hot leg alignment.

SALEM - UNIT 1 3/45-5a Amendment No. 299

CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM LIMITING CONDITION FOR.~ObP~EldRA~Tddl~O~N~================

3.6.2.1 Two independent containment spray systems shall be OPERABLE with each spray system capable of taking suction from the RWST and transferring suction to the RHR pump discharge.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With one containment spray system inoperable, restore the inoperable spray system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore the inoperable spray system to OPERABLE status within the next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.2.1 Each containment spray system shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

b. By verifying, that on recirculation flow, each pump develops a differential pressure of greater than or equal to 204 psid when tested pursuant to the Inservice Testing Program, INSERVICE TESTING PROGRAM

c. In accordance with the Surveillance Frequency Control Program during shutdown, by:

1. Verifying that each automatic valve in the flow path actuates to its correct position on a Containment High-High pressure test signal.

2. Verifying that each spray pump starts automatically on a Containment High-High pressure test signal.

d. Following activities that could result in nozzle blockage, either evaluate the work performed to determine the impact to the containment spray system, or perform an air or smoke flow test through each spray header and verifying each spray nozzle is unobstructed.

SALEM - UNIT 1 3/4 6-9 Amendment No. 299

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4.6.3.1.2 Each containment isolation valve shall be demonstrated OPERABLE during the COLD SHUTDOWN or REFUELING MODE in accordance with the Surveillance Frequency Control Program by:

a. Verifying that on a Phase A containment isolation test signal, each Phase A isolation valve actuates to its isolation position.

b. Verifying that on a Phase B containment isolation test signal, each Phase B isolation valve actuates to its isolation position.

c. Not used.

d. Verifying that on a Containment Purge and Pressure-Vacuum Relief isolation test signal, each required Purge and each Pressure-Vacuum Relief valve actuates to its isolation position.

e. Verifying that the Containment Pressure-Vacuum Relief Isolation valves are limited to s 60% opening angle.

4.6.3.1.3 In accordance with the Surveillance Frequency Control Program, verify that on a main steam isolation test signal, each main steam isolation valve actuates to its isolation position.

4.6.3.1.4 The isolation time of each power operated or automatic containment isolation valve shall be determined to be within its limit when tested pursuant to the Inservice Testing Program.

INSERVICE TESTING PROGRAM 4.6.3.1.5 Each required containment purge isolation valve shall be demonstrated OPERABLE within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after each closing of the valve, except when the valve is being used for multiple cyclings, then in accordance with the Surveillance Frequency Control Program, by verifying that when the measured leakage rate is added to the leakage rates determined pursuant to Specification 4.6.1.2.b for all other Type Band C penetrations, the combined leakage rate is less than or equal to 0.60La.

4.6.3.1.6 A pressure drop test to identify excessive degradation of resilient valve seals shall be conducted on the:

a. Required Containment Purge Supply and Exhaust Isolation Valves in accordance with the Surveillance Frequency Control Program.

b. Deleted.

4.6.3.1.7 The required containment purge supply and exhaust isolation valves shall be determined closed in accordance with the Surveillance Frequency Control Program.

SALEM - UNIT 1 3/46-13 Amendment No. 299

3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.7.1.1 All main steam line code safety valves (MSSVs) associated with each steam generator shall be OPERABLE.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

a. With one or two main steam line code safety valves inoperable in one or more steam generators, operation in Modes 1, 2 and 3 may proceed provided, that within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , either the inoperable valve is restored to OPERABLE status or reduce power to less than or equal to the applicable percent of RATED THERMAL POWER per Table 3.7-1; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With three main steam line code safety valves inoperable in one or more steam generators, operation in Modes 1, 2 and 3 may proceed provided, that within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , either the inoperable valves are restored to OPERABLE status or reduce power to less than or equal to the applicable percent of RATED THERMAL POWER per Table 3.7-1 and within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months , reduce the Power Range Neutron Flux High trip setpoint to less than or equal to the RATED THERMAL POWER per Table 3.7-1; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.7.1.1 Verify each required MSSV lift setpoint per Table 4.7-1. No additional Surveillance Requirements other than those required by the Inservice Testing Program.

INSERVICE TESTING PROGRAM SALEM - UNIT 1 3/47-1 Amendment No. 297

PLANT SYSTEMS MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.7.1.5 Each main steam line isolation valve shall be OPERABLE.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

MODES 1 - With one main steam line isolation valve inoperable, POWER OPERATION may continue provided the inoperable valve is either restored to OPERABLE status or closed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; otherwise, be in MODE 2 within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

MODES 2- With one or more main steam line isolation valve(s) inoperable, subsequent and 3 operation in MODES 2 or 3 may proceed provided;

a. The isolation valve(s) is (are) maintained closed, and

b. The isolation valve(s) is (are) verified closed once per 7 days.

Otherwise, be in MODE 3, HOT STANDBY, within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and MODE 4, HOT SHUTDOWN, within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS

4. 7.1.5 Each main steam line isolation valve shall be demonstrated OPERABLE by verifying full closure within 5 seconds when tested pursuant to the lnservice Testing Program. The provisions of Specification 4.0.4 are not applicable. INSERVICE TESTING PROGRAM SALEM- UNIT 1 3/4 7-10 Amendment No. 297

REFUELING OPERATIONS LOW WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.8.2 Two independent Residual Heat Removal (RHR) loops shall be OPERABLE.*

APPLICABILITY: MODE 6 when water level above the top of the reactor pressure vessel flange is less than 23 feet.

ACTION:

a. With less than the required RHR loops operable, immediately initiate corrective action to return the required RHR loops to OPERABLE status as soon as possible.

b. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.8.2 The required Residual Heat Removal loops shall be determined OPERABLE per the lnservice Testing Program.

INSERVICE TESTING PROGRAM

Systems supporting RHR loop operability may be excepted as follows:

a. The normal or emergency power source may be inoperable.

SALEM- UNIT 1 3/4 9-8a Amendment No. 297

ADMINISTRATIVE CONTROLS

3. If crack indications are found in portions of the SG tube not excluded above, then the next inspection for each affected and potentially affected SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever results in more frequent inspections). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

e. Provisions for monitoring operational primary-to-secondary leakage.

6.8.4.j lnservice Testing Program Deleted This Program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components. The program shall include the following:

a. Testing frequencies applicable to the ASME Code for Operations and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as follows:

ASME OM Code and applicable Required Frequencies for Addenda terminology for performing inservice inservice testing activities testing activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days

b. The provisions of Specification 4.0.2 are applicable to the above required frequencies and to other normal and accelerated frequencies specified as 2 years or less in the lnservice Testing Program for performing inservice testing activities,

c. The provisions of Specification 4.0.3 are applicable to inservice testing activities, and

d. Nothing in the ASME OM Code shall be construed to supersede the requirements of any Technical Specification.

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. 309

INDEX DEFINITIONS SECTION 1.0 DEFINITIONS DEFINED TERMS .........*...*.*.....**...******....*.*.******...*.. 1-1 ACTION **..****.****..****.*..**. ; **.*.*..*...**.***...*.*..*...*. 1-1 AXIAL FLUX DIFFERENCE ......**...*.*..**.*.***.*****.*..*.*.*.*.** 1-1 CHANNEL CALIBRATION *.***..*.****......**.**.*...**.*...****.*.*.. 1-1 CHANNEL CHECK *****..***.**.*.**.***..*****.*.*.******.*...*****.* 1-1 CHANNEL FUNCTIONAL TEST **.***.**..**.******.*.**.*************.*. 1~1 CONTAINMENT INTEGRITY *******.******.*.**.*.***.**.****.*.****.*.. 1-2 CORE ALTERATION **.***.*.**.********.********.************.**.***. 1-2 CORE OPERATING LIMITS REPORT .*..** ; **.********..***********.****** 1-2 DOSE EQUIVALENT I-131 .*.*****.*******.****.***..********.**...*** 1-2 E-AVERAGE DISINTEGRATION ENERGY .**..****.********.******.***..*** 1-3 ENGINEERED SAFETY FEATURE RESPONSE TIME ******..****.**.****.****** 1-3 FREQUENCY NOTATION ****************.** .- **.**.***.***************** 1-3 FULLY WITHDRAWN **************.*****************.***************.* 1-3 GASEOUS RADWASTE TREATMENT SYSTEM ******************.*****.*....** 1-3 IDENTIFIED LEAKAGE *****.***********.**************.*********.*..* 1-3 MEMBER (S) OF THE PUBLIC ********.**********.***.****************** 1-4 OFFSITE DOSE CALCULATION MANUAL (ODCM) *****************.***.***.* 1-4 OPERABL!i - OPERABILITY 0 0 ** 0 ******** 0 ' ** 0 0 0 ********* I ******** 0 ** 0

1-4 OPERATIONAL MODE - MODE *****..**.********.**.***.**.***.*****.*.** 1-4 PHYSICS TESTS **.*********..**.**..**.****.**.***.**************** 1-5 PRESSURE BOUNDARY LEAKAGE .***.******.**********************.*.*.* 1-5 PROCESS CONTROL PROGRAM (PCP) ***.***.****.*.*.*..**********.*...* 1-5 PURGE-PURGING ...**...*...*.*.*******....*.***.**.****...*.*.*.*.. 1-5 QUADRANT POWER TILT RATIO **************.*.**..************.****** 1-5 RATED THERMAL POWER **..***.*.**..*..***..**.***********.*..**.**. 1-5 REACTOR TRIP SYSTEM RESPONSE TIME *.**********.*****.******.**.*.. 1-6 REPORTABLE EVENT ..**.****...**...**.***.*****.**.**.*.*.*****.*.* 1-6 SHUTDOWN MARGIN *.******.***.*.***...*****.*********.*********.*.* 1-6 SITE BOUNDARY ***************..********.***********.*******.*..*.* 1-6 SOLIDIFICATION .************.*.*****.**.********.****.****..*.*** ; 1-6 SOURCE CHECK ***********.***.*.*****.*..***.**......***.**..****.* 1-6 STAGGERED TEST BASIS *******..*******.*****..**************.****.* 1-6 THERMAL POWER . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-7 UNIDENTIFIED LEAKAGE **.*..*.***.****.*********.***.**.*****...... 1-7 UNRESTRICTED AREA .*************.****..*****.***.****.*****.*****. 1-7 VENTILATION EXHAUST TREATMENT SYSTEM **..********...****.**..****.. 1-7 VENTING .***********.****.**..**.*****.*****.*.********.**.***.*** 1-7 INSERVICE TESTING PROGRAM ......................................................... 1-4 SALEM - UNIT 2 I Amendment No. 206

INSERVICE TESTING PROGRAM 1.15.1 The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f).

DEFINITIONS

b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE, or

c. Reactor coolant system leakage through a steam generator to the secondary system (primary-to-secondary leakage) .

MEMBER(S) OF THE PUBLIC 1.16 MEMBER(S) OF THE PUBLIC shall be all those persons who are not occupationally associated with the plant. This category does not include employees of PSE&G, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.

OFFSITE DOSE CALCULATION MANUAL (ODCM) 1.17 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm/Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent controls and Radiological Environmental Monitoring programs required by Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Specifications 6.9.1.7 and 6.9.1.8 respectively.

OPERABLE - OPERABILITY 1.18 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s),

and when all necessary attendant instrumentation, controls, normal or emergency electrical power source, cooling and seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its specified safety function(s) are also capable of performing their related support function(s).

OPERATIONAL MODE - MODE 1.19 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.1.

SALEM - UNIT 2 1-4 Amendment No. 262

REACTIVITY CONTROL SYSTEMS CHARGING PUMP- SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.3 At least one charging pump in the boron injection flow path required by Specification 3.1.2.1 shall be OPERABLE.#

APPLICABILITY: MODES 4, 5 and 6.

ACTION:

With no charging pump OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until one charging pump is restored to OPERABLE status.

SURVEILLANCE REQUIREMENTS 4.1.2.3 No additional Surveillance Requirements other than those required by the lnservice Testing Program. INSERVICE TESTING PROGRAM

A maximum of one centrifugal charging pump shall be OPERABLE while in MODE 4 when the temperature of one or more of the RCS cold legs is less than or equal to 312°F, MODE 5, or MODE 6 when the head is on the reactor vessel.

SALEM - UNIT 2 3/4 1-9 Amendment No. 279

REACTIVITY CONTROL SYSTEMS CHARGING PUMPS- OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two charging pumps shall be OPERABLE.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

With only one charging pump OPERABLE, restore at least two charging pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1% delta klk at 200°F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two charging pumps to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.1.2.4 No additional Surveillance Requirements other than those required by the lnservice Testing Program. INSERVICE TESTING PROGRAM SALEM - UNIT 2 3/4 1-10 Amendment No. 279

REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY VALVES SAFETY VALVES- SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.2 A minimum of one pressurizer code safety valve shall be OPERABLE* with a lift setting of 2485 psig +/- 3%.**,***

APPLICABILITY: Mode 4 and 5 ACTION:

With no pressurizer code safety valve OPERABLE, immediately suspend all operations involving positive reactivity changes and place an OPERABLE RHR loop into operation in the shutdown cooling mode.

SURVEILLANCE REQUIREMENTS 4.4.2.1 No additional Surveillance Requirements other than those required by the lnservice Testing Program. INSERVICE TESTING PROGRAM

While in Mode 5, an equivalent size vent pathway may be used provided th~t the vent pathway is not isolated or sealed.

- The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure. *

- - Following testing the lift setting shall be reset to within +/- 1%.

SALEM - UNIT 2 3/4 4-5 Amendment No. 279

REACTOR COOLANT SYSTEM 3/4.4.3 SAFETY VALVES- OPERATING LIMITING CONDITION FOR OPERATION 3.4.3 All pressurizer code safety valves shall be OPERABLE with a lift setting of 2485 psig +/- 3%.*,**

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

With one pressurizer code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS 4.4.3 No additional Surveillance Requirements other than those required by the lnservice Testing Program. INSERVICE TESTING PROGRAM

The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

- Following testing the lift setting shall be reset to +/- 1%.

SALEM- UNIT 2 3/4 4-6 Amendment No. 279

REACTOR COOLANT SYSTEM 3/4.4.5 RELIEF VALVES INSERVICE TESTING PROGRAM 4.4.5.1 In addition to the requirements of the Inservice Testing Program, each PORV shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by:

a. Operating the PORV through one complete cycle of full travel during MODES 3 or 4, and

b. Operating solenoid valves, air control valves, and check valves on associated air accumulators in PORV control systems through one complete cycle of full travel, and

c. Performing a CHANNEL CALIBRATION of the actuation instrumentation.

4.4.5.2 Each block valve shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by operating the valve through one complete cycle of full travel unless the block valve is closed in order to meet the requirements of ACTION b, or c in Specification 3.4.5.

SALEM - UNIT 2 3/44-8a Amendment No. 282

SURVEILLANCE REQUIREMENTS (Continued) c*. Verifying primary-to-secondary leakage is s; 150 gallons per day through anyone steam generator in accordance with the Surveillance Frequency Control Program during steady state operation, d*. Performance of a Reactor Coolant System water inventory balance** in accordance with the Surveillance Frequency Control Program. The water inventory balance shall be performed with the plant at steady state conditions.

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

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

4.4.7.2.2 Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1 shall be demonstrated OPERABLE pursuant to the Inservice Testing Program, except that in lieu of any leakage testing required by the Inservice Testing Program, each valve shall be demonstrated OPERABLE by verifying leakage to be within its limit: INSERVICE TESTING PROGRAM INSERVICE a. In accordance with the Surveillance Frequency Control Program.

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

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

d. For the Residual Heat Removal and Safety Injection Systems hot and cold leg injection valves and accumulator valves listed in Table 3.4-1 the testing will be done within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or flow through the valve. For all other systems testing will be done once per refueling.

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

Not required to be completed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

- Not applicable to primary-to-secondary leakage.

SALEM - UNIT 2 3/44-18 Amendment No. 282

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

a. Performance of a CHANNEL FUNCTIONAL TEST on the POPS actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which the POPS is required OPERABLE and in accordance with the Surveillance Frequency Control Program thereafter when the POPS is required OPERABLE.

b. Performance of a CHANNEL CALIBRATION on the POPS actuation channel in accordance with the Surveillance Frequency Control Program.

c. Verifying the POPS isolation valve is open in accordance with the Surveillance Frequency Control Program when the POPS is being used for overpressure protection.

d. Testing pursuant to the Inservice Testing Program. INSERVICE TESTING PROGRAM 4.4.10.3.2 The RCS vent(s) shall be verified to be open in accordance with the Surveillance Frequency Control Program* when the vent(s) is being used for overpressure protection.

Except when the vent pathway is provided with a valve which is locked, sealed, or otherwise secured in the open position, then verify these valves open in accordance with the Surveillance Frequency Control Program.

SALEM - UNIT 2 3/44-32 Amendment No. 282

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

f. By verifying that each of the following pumps develops the indicated Total Dynamic Head (TDH) when tested at the test flow point pursuant to the Inservice Testing Program: INSERVICE TESTING PROGRAM

1. Centrifugal Charging pump ~ 2338 psi TDH

2. Safety Injection pump ~ 1369 psi TDH

3. Residual Heat Removal pump ~ 165 psi TDH

g. By verifying the correct position of each of the following ECCS throttle valves:

1. Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following completion of each valve stroking operation or maintenance on the valve when the ECCS subsystems are required to be OPERABLE.

2. In accordance with the Surveillance Frequency Control Program.

HPSI System LPSI System Valve Number Valve Number 21 SJ 16 21 SJ 138 22 SJ 16 22 SJ 138 23 SJ 16 23 SJ 138 24 SJ 16 24 SJ 138 21 SJ 143 22 SJ 143 23 SJ 143 24 SJ 143

h. By performing a flow balance test, during shutdown, following completion of modifications to the ECCS subsystems that alter the subsystem flow characteristics and verifying that:

1. For Safety Injection pumps, with a single pump running:

a) The sum of the injection line flow rates, excluding the highest flow rate, is ~ 453 gpm, and b) The total flow rate through all four injection lines is ::; 647 gpm, and c) The difference between any pair of injection line flow rates is

12.0 gpm, and d) The total pump flow rate is: :; 664 gpm in the cold leg alignment, and e) The total pump flow rate is ::; 654 gpm in the hot leg alignment.

SALEM - UNIT 2 3/45-6 Amendment No. 282

3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM 3.6.2.1 Two independent containment spray systems shall be OPERABLE with each spray system capable of taking suction from the RWST and transferring suction to the RHR pump discharge.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With one containment spray system inoperable, restore the inoperable spray system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore the inoperable spray system to OPERABLE status within the next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

4.6.2.1 Each containment spray system shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

b. By verifying, that on recirculation flow, each pump develops a differential pressure of greater than or equal to 204 psid when tested pursuant to the Inservice Testing Program. INSERVICE TESTING PROGRAM

c. In accordance with the Surveillance Frequency Control Program during shutdown, by:

1. Verifying that each automatic valve in the flow path actuates to its correct position on a Containment High-High pressure test signal.

2. Verifying each spray pump starts automatically on a Containment High-High pressure test signal.

d. Following activities that could result in nozzle blockage, either evaluate the work performed to determine the impact to the containment spray system, or perform an air or smoke flow test through each spray header and verifying each spray nozzle is unobstructed.

SALEM - UNIT 2 3/46-10 Amendment No. 282

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4.6.3.2 Each containment isolation valve shall be demonstrated OPERABLE during the COLD SHUTDOWN or REFUELING MODE in accordance with the Surveillance Frequency Control Program by:

a. Verifying that on a Phase A containment isolation test signal, each Phase A isolation valve actuates to its isolation position.

b. Verifying that on a Phase B containment isolation test signal, each Phase B isolation valve actuates to its isolation position.

c. NOT USED

d. Verifying that on a Containment Purge and Pressure-Vacuum Relief isolation test signal, each required Purge and each Pressure-Vacuum Relief valve actuates to its isolation position.

e. Verifying that the Containment Pressure-Vacuum Relief Isolation valves are limited to ::; 60° opening angle.

4.6.3.3 In accordance with the Surveillance Frequency Control Program, verify that on a main steam isolation test signal, each main steam isolation valve actuates to its isolation position.

4.6.3.4 The isolation time of each power operated or automatic containment isolation valve shall be determined to be within its limit when tested pursuant to the Inservice Testing Program.

INSERVICE TESTING PROGRAM 4.6.3.5 Each required containment purge isolation valve shall be demonstrated OPERABLE within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after each closing of the valve, except when the valve is being used for multiple cyclings, then in accordance with the Surveillance Frequency Control Program, by verifying that when the measured leakage rate is added to the leakage rates determined pursuant to Specification 4.6.1.2.b for all other Type Band C penetrations, the combined leakage rate is less than or equal to 0.60La.

4.6.3.6 A pressure drop test to identify excessive degradation of resilient valve seals shall be conducted on the:

a. Required Containment Purge Supply and Exhaust Isolation Valves in accordance with the Surveillance Frequency Control Program.

b. Deleted.

4.6.3.7 The required containment purge supply and exhaust isolation valves shall be determined closed in accordance with the Surveillance Frequency Control Program.

SALEM - UNIT 2 3/46-15 Amendment No. 282

3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.7.1.1 All main steam line code safety valves (MSSVs) associated with each steam generator shall be OPERABLE with lift settings as specified in Table 3.7-4.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

a. With one or two main steam line code safety valves inoperable in one or more steam generators, operation in Modes 1, 2 and 3 may proceed provided, that within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , either the inoperable valve is restored to OPERABLE status or reduce power to less than or equal to the applicable percent of RATED THERMAL POWER per Table 3.7-1; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With three main steam line code safety valves inoperable in one or more steam generators, operation in Modes 1, 2 and 3 may proceed provided, that within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , either the inoperable valves are restored to OPERABLE status or reduce power to less than or equal to the applicable percent of RATED THERMAL POWER per Table 3.7-1 and within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months , reduce the Power Range Neutron Flux High trip setpoint to less than or equal to the RATED THERMAL POWER per Table 3.7-1; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS

4. 7.1.1 Verify each required MSSV lift setpoint per Table 3. 7-4. No additional Surveillance Requirements other than those required by the lnservice Testing Program.

INSERVICE TESTING PROGRAM SALEM - UNIT 2 3/4 7-1 Amendment No. 279

PLANT SYSTEMS MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.7.1.5 Each main steam line isolation valve shall be OPERABLE.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

MODES 1 - With one main steam line isolation valve inoperable, POWER OPERATION may continue provided the inoperable valve is either restored to OPERABLE status or closed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; Otherwise, be in MODE 2 within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

MODES 2- With one or more main steam line isolation valve(s) inoperable, subsequent and 3 operation in MODES 2 or 3 may proceed provided;

a. The isolation valve(s) is (are) maintained closed, and

b. The isolation valve(s) is (are) verified closed once per 7 days.

Otherwise, be in MODE 3, HOT STANDBY, within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and MODE 4, HOT SHUTDOWN, within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.7.1.5 Each main steam line isolation valve shall be demonstrated OPERABLE by verifying full closure within 5 seconds when tested pursuant to the lnservice Testing Program. The provisions of Specification 4.0.4 are not applicable. INSERVICE TESTING PROGRAM SALEM - UNIT 2 3/4 7-10 Amendment No. 279

REFUELING OPERATIONS LOW WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.8.2 Two independent Residual Heat Removal (RHR) loops shall be OPERABLE.*

APPLICABILITY: MODE 6 when water level above the top of the reactor pressure vessel flange is less than 23 feet.

ACTION:

a. With less than the required RHR loops operable, immediately initiate corrective*

action to return the required RHR loops to OPERABLE status as soon as possible.

b. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.8.2 The required Residual Heat Removal loops shall be determined OPERABLE per the lnservice Testing Program. INSERVICE TESTING PROGRAM

Systems supporting RHR loop operability may be excepted as follows:

I

a. I The normal or emergency power source may be inoperable.

SALEM- UNIT 2 3/4 9-9 Amendment No. 279

ADMINISTRATIVE CONTROLS 6.8.4.j lnservice Testing Program Deleted This Program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components. The program shall include the following:

a. Testing frequencies applicable to the ASME Code for Operations and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as follows:

ASME OM Code and applicable Required Frequencies for Addenda terminology for performing inservice inservice testing activities testing activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days

b. The provisions of Specification 4.0.2 are applicable to the above required frequencies and to other normal and accelerated frequencies specified as 2 years or less in the lnservice Testing Program for performing inservice testing activities,

c. The provisions of Specification 4.0.3 are applicable to inservice testing activities, and

d. Nothing in the ASME OM Code shall be construed to supersede the requirements of any Technical Specification.

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 6-19f Amendment No. 291 I

LR-N16-0114 LAR S16-05 Attachment 3 Description and Assessment of the Proposed Alternative to the ASME Code Request in Accordance with 10 CFR 50.55.a(z)(2)

Alternative Due to Hardship Without a Compensating Increase in Quality and Safety

LR-N16-0114 LAR S16-05 Request in Accordance with 10 CFR 50.55.a(z)(2)

Alternative Due to Hardship Without a Compensating Increase in Quality and Safety Relief Request G-01

1.0 DESCRIPTION

The request is to adopt a proposed alternative to the American Society of Mechanical Engineers (ASME) Operation and Maintenance (OM) Code by adoption of approved Code Case OMN-20, "Inservice Test Frequency."

2.0 ASSESSMENT Technical Evaluation of the Proposed Alternative to the OM Code Section IST of Division 1 of the OM Code, which is incorporated by reference in 10 CFR 50.55a(a), specifies component test frequencies based either on elapsed time periods (e.g., quarterly, 2 years) or on the occurrence of a plant condition or event (e.g., cold shutdown, refueling outage).

ASME Code Case OMN-20, "Inservice Test Frequency," has been approved for use by the ASME OM committee as an alternative to the test frequencies for pumps and valves specified in ASME OM Division: 1 Section IST 2009 Edition through OMa-2011 Addenda, and all earlier editions and addenda of ASME OM Code.

Code Case OMN-20 is not referenced in the latest revision of Regulatory Guide 1.192 (August 2014) as an acceptable OM Code Case to comply with 10 CFR 50.55a(f) requirements as allowed by 10 CFR 50.55a(b)(6). The proposed alternative is to use Code Case OMN-20 to extend or reduce the IST frequency requirements for the Salem Units 1 and 2 fourth 10 year IST interval or until OMN-20 is incorporated into the next revision of Regulatory Guide 1.192.

ASME Code Components Affected The Code Case applies to pumps and valves specified in ASME OM Division: 1 Section IST 2009 Edition through OMa-2011 Addenda and all earlier editions and addenda of ASME OM Code. Frequency extensions may also be applied to accelerated test frequencies (e.g., pumps in Alert Range) as specified in OMN-20.

For pumps and valves with test periods of 2 years of less, the test frequency allowed by OMN-20 and the current TS Inservice Testing Program (as modified by TS 4.0.2 and EGM 2012-001) are the same. For pumps and valves with test frequencies greater than 2 years, OMN-20 allows the test frequency to be extended by 6 months. The current TS Inservice Testing Program does not allow extension of test frequencies that are greater than 2 years.

1

LR-N16-0114 LAR S16-05

Applicable Code Edition and Addenda

ASME Code Case OMN-20 applies to ASME OM Division: 1 Section IST 2009 Edition through OMa-2011 Addenda and all earlier editions and addenda of ASME OM Code.

The Salem Units 1 and 2 Code Edition and Addenda that are applicable to the program interval are the ASME OM Code-2001 through the ASME OMb Code-2003 Addenda. The Salem Units 1 and 2 current interval ends August 30, 2019.

Applicable Code Requirement

This request is made in accordance with 10 CFR 50.55a(z)(2), and proposes an alternative to the requirements of 10 CFR 50.55a(f), which requires pumps and valves to meet the test requirements set forth in specific documents incorporated by reference in 10 CFR 50.55a(a).

ASME Code Case OMN-20 applies to Division 1, Section IST of the ASME OM Code and associated addenda incorporated by reference in 10 CFR 50.55a(a).

Reason for Request

The IST Program controls specified in Section 6.8.4.j of TS provide: a) a table specifying certain IST frequencies; b) an allowance to apply TS 4.0.2 to inservice tests required by the OM Code and with frequencies of two years or less; c) an allowance to apply TS 4.0.3 to inservice tests required by the OM Code; and d) a statement that, "Nothing in the ASME OM Code shall be construed to supersede the requirements of any TS." In Regulatory Issue Summary (RIS) 2012-10, "NRC Staff Position on Applying Surveillance Requirement 3.0.2 and 3.0.3 to Administrative Controls Program Tests," and Enforcement Guidance Memorandum (EGM) 2012-001, "Dispositioning Noncompliance with Administrative Controls Technical Specifications Programmatic Requirements that Extend Test Frequencies and Allow Performance of Missed Tests," the NRC stated that items b, c, and d of the TS IST Program were inappropriately added to the TS and may not be applied (although the EGM allows licensees to continue to apply those paragraphs pending a generic resolution of the issue).

In RIS 2012-10 and EGM 2012-001, the NRC stated that the current TS allowance to apply TS 4.0.2 and TS 4.0.3 to the Inservice Testing Program would no longer be permitted. In response, OMN-20, which provides allowances similar to TS 4.0.2, was approved and is proposed to be used as an alternative to the test periods specified in the OM code. The proposed alternative substitutes an approved Code Case for the existing TS requirements that the NRC has determined are not legally acceptable as a TS allowance. This proposed alternative provides an equivalent level of safety as the existing TS allowance, while maintaining consistency with 10 CFR 50.55a and the ASME OM Code.

Proposed Alternative and Basis for Use The proposed alternative is OMN-20, "Inservice Test Frequency," which addresses testing periods for pumps and valves specified in ASME OM Division 1, Section IST, 2009 Edition through OMa-2011 Addenda, and all earlier editions and addenda of ASME OM Code.

This request is being made in accordance with 10 CFR 50.55a(z)(2), in that the existing requirements are considered a hardship without a compensating increase in quality and safety for the following reasons:

2

LR-N16-0114 LAR S16-05

1) For IST testing periods up to and including 2 years, Code Case OMN-20 provides an allowance to extend the IST testing periods by up to 25%. The period extension is to facilitate test scheduling and considers plant operating conditions that may not be suitable for performance of the required testing (e.g., performance of the test would cause an unacceptable increase in the plant risk profile due to transient conditions or other ongoing surveillance, test or maintenance activities). Period extensions are not intended to be used repeatedly merely as an operational convenience to extend test intervals beyond those specified. The test period extension and the statements regarding the appropriate use of the period extension are equivalent to the existing TS 4.0.2 allowance and the statements regarding its use in the TS 4.0.2 Bases. Use of the TS 4.0.2 period extension has been a practice in the nuclear industry for many decades and elimination of this allowance would place a hardship on PSEG when there is no evidence that the period extensions affect component reliability.

2) For IST testing periods of greater than 2 years, OMN-20 allows an extension of up to 6 months. The ASME OM Committee determined that such an extension is appropriate.

The 6-month extension will have a minimal impact on component reliability considering that the most probable result of performing any inservice test is satisfactory verification of the test acceptance criteria. As such, pumps and valves will continue to be adequately assessed for operational readiness when tested in accordance with the requirements specified in 10 CFR 50.55a(f) with the frequency extensions allowed by Code Case OMN-20.

3) As stated in EGM 2012-001, if an Inservice Test is not performed within its frequency, TS 4.0.3 will not be applied. The effect of a missed Inservice Test on the Operability of TS equipment will be assessed under the licensee's Operability Determination Program.

Duration of Proposed Alternative The proposed alternative is requested for the current 10 year IST interval or until Code Case OMN-20 is incorporated into a future revision of Regulatory Guide 1.192, referenced by a future revision of 10 CFR 50.55a, whichever occurs first.

Precedents The NRC approved the use of OMN-20 for North Anna on March 27, 2014 (NRC ADAMS Accession Number ML14084A407).

3

v t e Letter Sequence Request
CAC:MF8313, (Open)
Initiation Request Acceptance... Administration Questions Answers Results Approval... Other: ML17132A005
MONTHYEAR2016-08-30 [Table View]

LR-N16-0114, License Amendment Request to Revise Technical Specifications to Adopt TSTF-545, Revision 3, TS Lnservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing Using the Consolidated Line .

Text

Subject:

1.0 DESCRIPTION

3.0 REGULATORY ANALYSIS

1.0 DESCRIPTION

3.0 REGULATORY ANALYSIS

1.0 DESCRIPTION

Applicable Code Edition and Addenda

Applicable Code Requirement

Reason for Request

Navigation menu

LR-N16-0114, License Amendment Request to Revise Technical Specifications to Adopt TSTF-545, Revision 3, TS Lnservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing Using the Consolidated Line .

Text

Subject:

1.0 DESCRIPTION

3.0 REGULATORY ANALYSIS

1.0 DESCRIPTION

3.0 REGULATORY ANALYSIS

1.0 DESCRIPTION

Applicable Code Edition and Addenda

Applicable Code Requirement

Reason for Request

Navigation menu

Search