Text

Salem Nuclear Generating Station, Unit Nos. 1 and 2, License Amendment, Issuance of Amendments Relocation of Specific Surveillance Frequencies to a Licensee-Controlled Program Based on Technical Specification Task Force (TSTF) Change TSTF-4
	ML110410691
Person / Time
Site:	Salem
Issue date:	03/21/2011
From:	Richard Ennis; Plant Licensing Branch 1
To:	Joyce T; Public Service Enterprise Group
	Ennis R B, NRR/DORL, 415-1420
References
	TAC ME3574, TAC ME3575
	Download: ML110410691 (250)
	v • d • e

UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 March 21, 2011 Mr. Thomas Joyce President and Chief Nuclear Officer PSEG Nuclear P.O. Box 236, N09 Hancocks Bridge, NJ 08038 SALEM NUCLEAR GENERATING STATION, UNIT NOS. 1 AND 2, ISSUANCE OF AMENDMENTS RE: RELOCATION OF SPECIFIC SURVEILLANCE FREQUENCIES TO A LICENSEE-CONTROLLED PROGRAM BASED ON TECHNICAL SPECIFICATION TASK FORCE (TSTF) CHANGE TSTF-425 (TAC NOS. ME3574 AND ME3575)

Dear Mr. Joyce:

The Commission has issued the enclosed Amendment Nos. 299 and 282 to Facility Operating License I\los. DPR-70 and DPR-75 for the Salem Nuclear Generating Station, Unit Nos. 1 and 2. These amendments consist of changes to the Technical Specifications (TSs) and Facility Operating Licenses in response to your application dated March 23, 2010, as supplemented by letters dated November 19, 2010, January 31, 2011, and February 23, 2011. The amendments modify the TSs by relocating specific surveillance frequencies to a controlled program. The changes are based on Nuclear Regulatory Commission-approved TS Task Force (TSTF) change TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control -RITSTF [Risk-Informed TSTF] Initiative 5b." A copy of our safety evaluation is also enclosed.

Notice of Issuance will be included in the Commission's biweekly Federal Register notice. Sincerely, Richard B. Ennis, Senior Project Manager Plant Licensing Branch 1-2 Division of Operating Reactor licensing Office of Nuclear Reactor Regulation Docket Nos. 50-272 and 50-311

Enclosures:

1. Amendment No. 299 to License No. DPR*70 2. Amendment No. 282 to License No. DPR-75 3. Safety Evaluation cc w/encls: Distribution via Listserv UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 PSEG NUCLEAR, EXELON GENERATION COMPANY, DOCKET NO. SALEM NUCLEAR GENERATING STATION, UNIT AMENDMENT TO FACILITY OPERATING Amendment No. 299 License No. DPR-70 The Nuclear Regulatory Commission (the Commission) has found that: The application for amendment filed by PSEG Nuclear LLC, acting on behalf of itself and Exelon Generation Company, LLC (the licensees) dated March 23, 2010, as supplemented by letters dated November 19, 2010, January 31, 2011, and February 23, 2011, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in Title 10 of the Code of Federal Regulations (10 CFR), Chapter I; The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-70 is hereby amended to read as follows:

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 299, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. This license amendment is effective as of its date of issuance and shall be implemented within 120 days. FOR THE NUCLEAR REGULATORY COMMISSION K. Chernoff, Chief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Facility Operating License and the Technical Specifications Date of Issuance:

March 21, 2011 ATTACHMENT TO LICENSE AMENDMENT NO. FACILITY OPERATING LICENSE NO. DOCKET NO. Replace the following page of Facility Operating License No. DPR-70 with the attached revised page as indicated.

The revised page is identified by amendment number and contains a marginal line indicating the area of change. Remove Page 4 Page Replace the following pages of the Appendix A, Technical Specifications, with the attached revised pages as indicated.

The revised pages are identified by amendment number and contain marginal lines indicating the areas of change. Remove Insert Remove Insert Remove Insert Remove 3/41-1 3/4 1-1 3/43-35 3/43-35 3/45-7 3/45-7 3/48-13 3/4 1-2 3/41-2 3/43-38 3/43-38 3/46-1 3/46-1 3/48-14 3/4 1-3 3/4 1-3 3/43-38a 3/46-5a 3/46-5a 3/48-15 3/4 1-7 3/4 1-7 3/43-46 3/43-46 3/46-6 3/46-6 3/49-1 3/41-15 3/4 1-15 3/43-57a 3/43-57a 3/46-11a 3/46-11a 3/49-10 3/4 1-16 3/4 1-16 3/43-58 3/43-58 3/46-13 3/46-13 3/49-11 3/4 1-17 3/4 1-17 3/43-60 3/43-60 3/4 7-5 3/47-5 3/49-12 3/4 1-19a 3/41-19a 3/44-1 3/44-1 3/4 7-7 3/4 7-7 3/4 10-2 3/4 2-3 3/4 2-3 3/44-6 3/44-6 3/47/17 3/47/17 3/42-12 3/42-12 3/4 4-16a. 3/44-16a 3/47-34 3/43-32 3/43-32 3/4 5-5 3/4 5-5 3/48-9a 3/41-8 3/4 1-8 3/43-48 3/43-48 3/46-7 3/46-7 3/49-2 3/4 3/41-9 3/41-9 3/43-53 3/43-53 3/4 6-9 3/4 6-9 3/4 9-4 3/4 3/4 1-14 3/4 1-14 3/43-57 3/43-57 3/46-10 3/46-10 3/4 9-8 3/4 3/4 1-18a 3/41-18a 3/43-71 3/43-71 3/4 7-6 3/4 7-6 3/4 10-1 3/4 3/4 1-21 3/4 1-21 3/44-2a 3/44-2a 3/47-9 3/4 7-9 3/4 10-3 3/4 3/4 1-22 3/4 1-22 3/44-3a 3/44-3a 3/47-14 3/47-14 3/4 10-4 3/4 3/4 1-23 3/4 1-23 3/44-3b 3/44-3b 3/47-15 3/47-15 3/4 11-7 3/4 3/4 2-2 3/4 2-2 3/44-5a 3/44-5a 3/47-16 3/47-16 3/411-15 3/4 3/4 2-6 3/42-6 3/44-14 3/44-14 3/47-20 3/42-7 3/42-7 3/44-15 3/44-15 3/47-23 3/42-10 3/42-10 3/44-16 3/44-16 3/47-26 3/42-13 3/42-13 3/44-22 3/44-22 3/47-35 3/43-1 3/43-1 3/44-25 3/44-25 3/4 8-3 3/4 3/43-11 3/4 3-11 3/44-29 3/44-29 3/4 8-4 3/4 3/43-12 3/43-12 3/44-31 3/44-31 3/4 8-5 3/4 3/43-13 3/43-13 3/44-35 3/44-35 3/48-5a 3/43-13a . 3/45-1 3/45-1 3/4 8-6 3/4 3/43-14 3/43-14 3/45-2 3/4 5-2 3/48-7 3/4 3/43-31a 3/43-31a 3/4 5-4 3/4 5-4 3/4 8-9 3/4 3/43-32a 3/43-32a 3/45-5a 3/45-5a 3/48-10 3/43-33 3/43-33 3/45-6a 3/45-6a 3/48-12 3/43-34 3/43-34 3/45-6b 3/45-6b 3/48-12a

-4 Maximum Power Level PSEG Nuclear LLC is authorized to operate the faci at a steady state reactor core power level not in excess of 3459 megawatts (one hundred percent of rated core power) . Technical Specifications The Technical Specifications contained in A and B, as revised through Amendment No. 299, are hereby incorporated in the license. The licensee shall operate the facil in accordance with the Technical Specifications. Deleted Per Amendment 22, 11 20 79 Less than Four Loop Operation PSEG Nuclear LLC shall not operate the reactor at power levels P-7 (as defined in Table 3.3 1 of fication 3.3.1.1 of Appendix A to this license) with less than four (4) reactor coolant loops in operation for less than four loop operation have been submitted by the licensees and approval for less than four loop operation at power levels above P-7 has been granted by the Commission by Amendment of this license. PSEG Nuclear LLC shall implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final safety Analysis Report, and as in the NRC Safety Evaluation Report dated November 20, 1979, and in its supplements, subject to the following sion: PSEG Nuclear LLC may make changes to the approved fire protection program without prior approval of the Commission if those changes would not affect the ability to achieve and maintain safe shutdown in the event of fire. Amendment No. 299 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN -T §Y9 > 200°F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be 1.3% APPLICABILITY:

MODES 1, 2*, 3, and With the SHUTDOWN MARGIN < 1.3% llklk, immediately initiate and continue boration 33 gpm of a solution containing 6,560 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be 1.3% llklk: a. Within one hour after detection of an inoperable control rod(s) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the rod(s) is inoperable.

If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the immovable or untrippable control rod(s). b. When in MODES 1 or 2#, in accordance with the Surveillance Frequency Control Program by verifying that control banks are within the limits in the COLR per Specification 3.1.3.5. c. When in MODE 2##, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to achieving reactor criticality by verifying that the predicted critical control rod position is within the limits in the COLR per specification 3.1.3.5.

See Special Test Exception 3.10.1 # With Keff 1.0 ## With Keff < 1.0 SALEM -UNIT 1 3/4 1-1 Amendment No. 299 REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of e below, with the control banks at the maximum insertion limit in the COLR per Specification 3.1.3.5. When in MODES 3 or 4, in accordance with the Surveillance Frequency Control Program by consideration of the following factors: Reactor coolant system boron concentration, Control rod position, Reactor coolant system average temperature, Fuel burnup based on gross thermal energy generation, Xenon concentration, and Samarium concentration.

4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within +/- 1 % Llklk in accordance with the Surveillance Frequency Control Program. This comparison shall consider at least those factors stated in Specification 4.1.1.1.1.e, above. The predicted reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 Effective Full Power Days after each fuel loading. SALEM -UNIT 3/4 1-2 Amendment No. 299 REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN -T ml < 200°F LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be 2: 1.0% APPLICABILITY:

MODE With the SHUTDOWN MARGIN < 1.0% Llklk, immediately initiate and continue boration 2: 33 gpm of a solution containing 2: 6,560 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

4.1.1.2 The SHUTDOWN MARGIN shall be determined to be 2: 1.0% Llklk: Within one hour after detection of an inoperable control rod(s) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the rod(s) is inoperable.

If the inoperable control rod is immovable or untrippable, the SHUTDOWN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the immovable or untrippable control rod(s). In accordance with the Surveillance Frequency Control Program by consideration of the following factors: Reactor coolant system boron concentration, Control rod position, Reactor coolant system average temperature, Fuel burnup based on gross thermal energy generation, Xenon concentration, and Samarium concentration.

SALEM -UNIT 3/4 1-3 Amendment No. 299 REACTIVITY CONTROL SYSTEMS 3/4.1.2 BORATION SYSTEMS FLOW PATHS -SHUTDOWN As a minimum, one of the following boron injection flow paths shall be OPERABLE: A flow path from the boric acid tanks via a boric acid transfer pump and a charging pump to the Reactor Coolant System if the boric acid storage system is OPERABLE, per Specification 3.1.2.6a while in MODE 4, or per Specification 3.1.2.Sa while in MODE S or 6, or A flow path from the refueling water storage tank via a charging purnp to the Reactor Coolant System if the refueling water storage tank is OPERABLE per Specification 3.1.2.6b while in MODE 4, or per Specification 3.1.2.Sb while in MODE S or 6. APPLICABILITY:

MODES 4, Sand 6. ACTION: With none of the above flow paths OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until at least one injection path is restored to OPERABLE status. 4.1.2.1 At least one of the above required flow paths shall be demonstrated OPERABLE: When the boric acid tank is a required water source, by verifying in accordance with the Surveillance Frequency Control Program that: The flow path from the boric acid tank to the boric acid transfer pump, the boric acid transfer pump, and the recirculation path from the boric acid transfer pump to the boriC acid tank is ?! 63°F, and The flow path between the boric acid transfer pump recirculation line to the charging pump suction line is ?! SO°F, 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.

SALEM -UNIT 3/4 1-7 Amendment No. 299 REACTIVITY CONTROL SYSTEMS FLOW PATHS -OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.2 At least two of the following three boron injection flow paths shall be OPERABLE:

a. A flow path from the boric acid tanks via a boric acid transfer pump and a charging pump to the Reactor Coolant System. b. Two flow paths from the refueling water storage tank via charging pumps to the Reactor Coolant System. APPLICABILITY:

MODES 1, 2 and 3. ACTION: With only one of the above required boron injection flow paths to the Reactor Coolant System OPERABLE, restore at least two boron injection flow paths to the Reactor Coolant 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 and borated to a SHUTDOWN MARGIN equivalent to at least 1 % delta k/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 flow paths 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 4.1.2.2 Each of the above required flow paths shall be demonstrated OPERABLE: By verifying in accordance with the Surveillance Frequency Control Program that The flow path from the boric acid tank to the boric acid transfer pump and from the recirculation line back to the boric acid tank is ;;:: 63°F, and the flow path between the boric acid tank recirculation line to the charging pump suction line is ;;:: 50°F, 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.

SALEM -UNIT 3/4 1-8 Amendment No, 299 REACTIVITY CONTROL SYSTEMS SUB\LEILLANCE REQUIREMENTS (Continued)

c. In accordance with the Surveillance Frequency Control Program during shutdown by verifying that each automatic valve in the flow path actuates to its correct position on a safety injection test signal. d. In accordance with the Surveillance Frequency Control Program by verifying that the flow path required by specification 3.1.2.2.a delivers at least 33 gpm to the Reactor Coolant System. SALEM -UNIT 1 3/4 1-9 Amendment No. 299 REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES -SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.5 As a minimum, one of the following borated water sources shall be OPERABLE: A boric acid storage system with: A minimum contained volume of 2,600 gallons, Between 6,560 and 6,990 ppm of boron, and, A minimum solution temperature of 63°F. The refueling water storage tank with: A minimum contained volume of 37,000 gallons, A minimum boron concentration of 2300 ppm, and 3. A minimum solution temperature of 35°F. APPLICABILITY:

MODES 5 and 6. ACTION: With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until at least one borated water source is restored to OPERABLE status. 4.1.2.5 The above required borated water source shall be demonstrated OPERABLE: For the boric acid storage system, when it is the source of borated water in accordance with the Surveillance Frequency Control Program by: Verifying the boron concentration of the water, Verifying the water level of the tank, and Verifying the boric acid storage tank solution temperature when it is the source of borated water. SALEM -UNIT 3/4 1-14 Amendment No. 299 REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) For the refueling water storage tank by: Verifying the boron concentration in accordance with the Surveillance Frequency Control Program, Verifying the borated water volume in accordance with the Surveillance Frequency Control Program, and Verifying the solution temperature in accordance with the Surveillance Frequency Control Program, when it is the source of borated water and the outside air temperature is less than 35°F. SALEM -UNIT 3/4 1-15 Amendment No. 299 REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES -OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.6 As a minimum, the following borated water source(s) shall be OPERABLE as required by specifications 3.1.2.1 and 3.1.2.2: A boric acid storage system with: A contained volume of borated water in accordance with figure 3.1.2, A boron concentration in accordance with figure 3.1-2, and A minimum solution temperature of 63"F. The refueling water storage tank with: A contained volume of between 364,500 and 400,000 gallons of water, A boron concentration of between 2,300 and 2,500 ppm, and 3. A minimum solution temperature of 35"F. APPLICABILITY:

MODES 1, 2, 3 and 4. ACTION: With the boric acid storage system inoperable and being used as one of the above required boration water systems, restore the storage 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 and borated to a SHUTDOWN MARGIN equivalent to at least 1% delta Klk at 200°F; restore the boric acid storage system 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 . With the refueling water storage tank inoperable, restore the tank to OPERABLE status within one hour 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 . SALEM -UNIT 3/4 1-16 Amendment No 299 REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS 4.1.2.6 Each borated water source shall be demonstrated OPERABLE: For the boric acid storage system, when it is the source of borated water in accordance with the Surveillance Frequency Control Program by: Verifying the boron concentration in each water source, Verifying the water level of each water source, and Verifying the boric acid storage system solution temperature. For the refueling water storage tank by: Verifying the boron concentration in accordance with the Surveillance Frequency Control Program, Verifying the borated water volume in accordance with the Surveillance Frequency Control Program, and Verifying the solution temperature in accordance with the Surveillance Frequency Control Program when the outside air temperature is less than 35°F. SALEM -UNIT 3/41-17 Amendment No. 299 REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) A reevaluation of each accident analysis of table 3.1-1 is performed within 5 days; this reevaluation shall confirm that the previously analyzed results of these accidents remain valid for the duration of operation under these conditions. The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . A core power distribution measurement is obtained and Fo(Z) FN llH are verified to be within their limits within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . The THERMAL POWER level is reduced to less than or equal to 75% of RATED THERMAL POWER within one hour and within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the high neutron flux trip setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER. THERMAL POWER shall be maintained less than or equal to 75% of RATED THERMAL POWER until compliance with ACTIONS 3.1.3.1.c.3.a and 3.1.3.1.c.3.c above are demonstrated.

SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The position of each full length rod shall be determined to be within the limits established in the limiting condition for operation in accordance with the Surveillance Frequency Control Program (allowing for one hour thermal soak after rod motion) except during time intervals when the Rod Position Deviation Monitor is inoperable, then verify the group positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . 4.1.3.1.2 Each full length rod not fully inserted in the core shall be determined to be OPERABLE by movement of at least 10 steps in anyone direction in accordance with the Surveillance Frequency Control Program. SALEM -UNIT 3/41-18a Amendment No. 299 REACTIVITY CONTROL LIMITING CONDITION FOR OPERATION

b. With two or more analog rod position indicators per bank inoperable, within one hour restore the inoperable rod position indicator(s) to OPERABLE status or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . A maximum of one rod position indicator per bank may remain inoperable following the hour, with Action (a) above being applicable from the original entry time into the LCO. c. With a maximum of one group demand position indicator per bank inoperable either: Verify that all analog rod position indicators for the affected bank are OPERABLE and that the most withdrawn rod and the least withdrawn rod of the bank are within a maximum of 18 steps when reactor power is s; 85% RATED THERMAL POWER or if reactor power is> 85% RATED THERMAL POWER, 12 steps of each other at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , or Reduce THERMAL POWER to less than 50% of RATED POWER within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . SURVEILLANCE REQUIREMENTS 4.1.3.2.1.1 Each analog rod position indicator shall be determined to be OPERABLE by verifying that the demand position indication system and the rod position indication system agree within 18 steps when reactor power is s; 85% RATED THERMAL POWER or if reactor power is> 85% RATED THERMAL POWER, 12 steps (allowing for one hour thermal soak after rod motion) in accordance with the Surveillance Frequency Control Program except during time intervals when the Rod Position Deviation Monitor is inoperable, then compare the demand position indication system and the rod position indication system at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . 4.1.3.2.1.2 Each of the above required rod position indicator(s) shall be determined to be OPERABLE by performance of a CHANNEL calibration in accordance with the Surveillance Frequency Control Program. SALEM -UNIT 3/41-19a Amendment No. 299 REACTIVITY CONTROL SYSTEMS ROD DROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.3 The individual full length (shutdown and control) rod drop time from 230 steps withdrawn shall be less than or equal to 2.7 seconds from beginning of decay of stationary gripper coil voltage to dashpot entry with: T avg greater than or equal to 541°F, and b. All reactor coolant pumps operating.

APPLICABILITY:

MODES 1 &2. ACTION: With the drop time of any full length rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2. SURVEILLANCE REQUIREMENTS 4.1.3.3 The rod drop time of full length rods shall be demonstrated through measurement prior to reactor criticality: For all rods following each removal of the reactor vessel head, For specifically affected individual rods following any maintenance on or modification to the control rod drive system which could affect the drop time of those specific rods, and In accordance with the Surveillance Frequency Control Program. SALEM -UNIT 3/4 1-21 Amendment No. 299 REACTIVITY CONTROL SYSTEMS SHUTDOWN ROD INSERTION LIMIT LIMITING CONDITION FOR OPERATION 3.1.3.4 All shutdown rods shall be FULLY APPLICABILITY:

MODES 1*, and With a maximum of one shutdown rod not FULLY WITHDRAWN, except for surveillance pursuant to Specification 4.1.3.1.2, within one hour a. FULLY WITHDRAW the rod, or, b. Declare the rod to be inoperable and apply Specification 3.1.3.1. 4.1.3.4 Each shutdown rod shall be determined to be FULLY WITHDRAWN by use of the group demand counters, and verified by the analog rod position indicators**:

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

and b. In accordance with the Surveillance Frequency Control Program thereafter.

See Special Test Exceptions 3.10.2 and 3.10.3 ** For power levels below 50% one hour thermal "soak time" is permitted.

During this soak time, the absolute value of rod motion is limited to six steps. # With Keff greater than or equal to 1.0 @ Surveillance 4.1.3.4.a is applicable prior to withdrawing control banks in preparation for startup (Mode 2). SALEM -UNIT 1 3/4 1-22 Amendment No. 299 REACTIVITY CONTROL SYSTEMS CONTROL ROD INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.5 The control banks shall be limited in physical insertion as specified in the OPERATING LIMITS REPORT APPLICABILITY:

MODES 1*, and With the control banks inserted beyond the above insertion limits, except for surveillance pursuant to Specification 4.1.3.1.2, a. Restore the control banks to within the limits within two hours, or b. Reduce THERMAL POWER within two hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the bank position using the insertion limits specified in the COLR, or c. Be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . SURVEILLANCE REQUIREMENTS 4.1.3.5 The position of each control bank shall be determined to be within the insertion limits in accordance with the Surveillance Frequency Control Program by use of the group demand counters and verified by the analog rod position indicators**

except during time intervals when the Rod Insertion Limit Monitor is inoperable, then verify the individual rod positions at least once per 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s**.

See Special Test Exceptions 3.10.2 and 3.10.3 ** For power levels below 50% one hour thermal "soak time" is permitted.

During this soak time, the absolute value of rod motion is limited to six steps. # With Keff greater than or equal to 1.0 SALEM -UNIT 1 3/4 1-23 Amendment No. 299 POWER DISTRIBUTION LIMITING CONDITION FOR OPERATION

b. THERMAL POWER shall not be increased above 90% of RATED THERMAL POWER unless the indicated AFD is within the limits specified in the COLR and ACTION a.2.a)1), above has been satisfied.

c. THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER unless the indicated AFD has not been outside of the limits specified in the COLR for more than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months penalty deviation cumulative during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . SURVEILLANCE REQUIREMENTS 4.2.1.1 The indicated AXIAL FLUX DIFFERENCE shall be determined to be within its limits during POWER OPERATION above 15% of RATED THERMAL POWER by: Monitoring the indicated AFD for each OPERABLE excore channel: In accordance with the Surveillance Frequency Control Program when the AFD Monitor Alarm is OPERABLE, and In accordance with the Surveillance Frequency Control Program for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after restoring the AFD Monitor Alarm to OPERABLE status. Monitoring and logging the indicated AXIAL FLUX DIFFERENCE for each OPERABLE excore channel in accordance with the Surveillance Frequency Control Program for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and in accordance with the Surveillance Frequency Control Program thereafter, when the AXIAL FLUX DIFFERENCE Monitor Alarm is inoperable.

The logged values of the indicated AXIAL FLUX DIFFERENCE shall be assumed to exist during the interval preceding each logging. 4.2.1.2 The indicated AFD shall be considered outside of its limits when at least 2 of 4 or 2 of 3 OPERABLE excore channels are indicating the AFD to be outside of the target band. Penalty deviation outside of the target band shall be accumulated on a time basis of: One minute penalty deviation for each one minute of POWER OPERATION outside of the limits at THERMAL POWER levels equal to or above 50% of RATED THERMAL POWER, and One-half minute penalty deviation for each one minute of POWER OPERATION outside of the limits at THERMAL POWER levels below 50% of RATED THERMAL POWER. SALEM -UNIT 3/4 2-2 Amendment No. 299 POWER DISTRIBUTION LIMITS REQUIREMENTS (Continued) 4.2.1.3 The target flux difference of each OPERABLE excore channel shall be determined by measurement in accordance with the Surveillance Frequency Control Program. The provisions of Specification 4.0.4 are not applicable.

4.2.1.4 The target flux difference shall be updated in accordance with the Surveillance Frequency Control Program by either determining the target flux difference pursuant to 4.2.1.3 above or by linear interpolation between the most recently measured value and 0 percent at the end of the cycle life. The provisions of Specification 4.0.4 are not applicable.

SALEM -UNIT 1 3/4 2-3 Amendment No. 299 POWER DISTRIBUTION The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 Fxy shall be evaluated to determine if Fo(Z) is within its limit by: Using the movable incore detectors to obtain a power distribution map: When THERMAL POWER is s 25%, but> 5% of RATED THERMAL POWER, or When the Power Distribution Monitoring System (PDMS) is inoperable; and increasing the Measured Fo(Z) by the applicable manufacturing and measurement uncertainties as specified in the COLR. Using the PDMS or the moveable incore detectors when THERMAL POWER is> 25% of RATED THERMAL POWER, and increasing the measured Fo(Z) by the applicable manufacturing and measurement uncertainties as specified in the COLR. Comparing the Fxy computed (FXYc) obtained in b, above to: The Fxy limits for RATED THERMAL POWER (FXyRTP) for the appropriate measured core planes given in e and f below, and The relationship:

FXyL = Fxt TP [1 + PF xy (1-P)] where Fx/ is the limit for fractional THERMAL POWER operation expressed as a function of FXyRTP, PF xy is the power factor multiplier for Fxy in the COLR, and P is the fraction of RATED THERMAL POWER at which Fxy was measured.

SALEM -UNIT 3/42-6 Amendment No. 299 POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENT&JContinued) Remeasuring Fxy according to the following schedule: When Fxyc is greater than the FxyRTP limit for the appropriate measured core plane but less than the FXyL relationship, additional core power distribution measurements shall be taken and Fxy C compared to Fxy RTP and F xy L.. Either within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding by 20% of RATED THERMAL POWER or greater, the THERMAL POWER at which Fxyc was last determined, or In accordance with the Surveillance Frequency Control Program, whichever occurs first. When the Fxyc is less than or equal to the F x/ TP limit for the appropriate measured core plane, additional core power distribution measurements shall be taken and Fxyc compared to FxyRTP and FxyL in accordance with the Surveillance Frequency Control Program. The Fxy limit for Rated Thermal Power (F x/ TP) shall be provided for all core planes containing bank "0" control rods and all unrodded core planes in the COLR per specification 6.9.1.9. The Fxy limits of e, above, are not applicable in the following core plane regions as measured in percent of core height from the bottom of the fuel: Lower core region from 0 to 15% inclusive. Upper core region from 85 to 100% inclusive. Grid plane regions at 17.8 +/- 2%, 32.1 +/- 2%,46.4 +/- 2%,60.6 +/- 2%, and 74.9 +/- 2% inclusive. Core plane regions within +/- 2% of core height (+/- 2.88 inches) about the bank demand position of the bank "0" control rods. Evaluating the effects of Fx( on Fa(Z) to determine if Fa(Z) is within its limit whenever Fxy C exceeds Fxy . SALEM -UNIT 3/4 2-7 Amendment No. 299 POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.3.1 shall be determined to be within its limit by obtaining a core power distribution measurement:

a. Prior to operation above 75% of RATED THERMAL POWER after each fuel loading, and b. In accordance with the Surveillance Frequency Control Program. c. The provisions of Specification 4.0.4 are not applicable.

4.2.3.2 The measured of 4.2.3.1 above, shall be increased by the applicable uncertainties specified in the COLR. SALEM -UNIT 1 Amendment No. 299 POWER DISTRIBUTION LIMITING CONDITION FOR OPERATION (Continued) reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High trip Setpoints to :s; 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95% or greater RATED THERMAL POWER. c. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to causes other than the misalignment of either a shutdown or control rod: Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip Setpoints to:s; 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified at 95% or greater RATED THERMAL POWER. SURVEILLANCE REQUIREMENTS 4.2.4 The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50% of RATED THERMAL POWER by: Calculating the ratio in accordance with the Surveillance Frequency Control Program when the alarm is OPERABLE. Calculating the ratio in accordance with the Surveillance Frequency Control Program during steady state operation when the alarm is inoperable. Obtaining a core power distribution measurement to determine the QUADRANT POWER TILT RATIO in accordance with the Surveillance Frequency Control Program when one Power Range Channel is inoperable and THERMAL POWER is > 75 percent of RATED THERMAL POWER. SALEM -UNIT 3/42-12 Amendment No. 299 POWER DISTRIBUTION LIMITS DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB related parameters shall be maintained within the limits shown on Table 3.2-1: a. Reactor Coolant System T avg. b. Pressurizer Pressure c. Reactor Coolant System Total Flow Rate APPLICABILITY:

MODE 1 ACTION: With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . 4.2.5.1 Each of the parameters of Table 3.2-1 shall be verified to be within their limits in accordance with the Surveillance Frequency Control Program. 4.2.5.2 The Reactor Coolant System total flow rate shall be determined to be within the limits of Table by performing a precision heat balance within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after achieving steady state conditions 2: 90% RATED THERMAL POWER in accordance with the Surveillance Frequency Control Program. The provisions of Specification 4.0.4 are not applicable.

SALEM -UNIT 1 3/42-13 Amendment No. 299 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1.1 As a minimum, the reactor trip system instrumentation channels and interlocks of 3.3-1 shall be APPLICABILITY:

As shown in Table As shown in Table SURVEILLANCE REQUIREMENTS 4.3.1.1.1 Each reactor trip system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies specified in the Surveillance Frequency Control Program unless otherwise noted in Table 4.3-1. 4.3.1.1.2 The logic for the interlocks shall be demonstrated OPERABLE prior to each reactor startup unless performed during the preceding 92 days. The total interlock function shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program during CHANNEL CALIBRATION testing of each channel affected by interlock operation.

4.3.1.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip function shall be verified to be within its limit in accordance with the Surveillance Frequency Control Program. Neutron detectors are exempt from response time testing. SALEM -UNIT 1 3/43-1 Amendment No. 299 TABLE REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE FUNCTIONAL UNIT Manual Reactor Trip Switch Power Range, Neutron Flux Power Range, Neutron Flux, High Positive Rate Deleted Intermediate Range, Neutron Flux Source Range, Neutron Flux Overtemperature liT Overpower liT Pressurizer Pressure--Low Pressurizer Pressure--High

11. Pressurizer Water Level--High Loss of Flow -Single Loop N.A N.A (7) (2), (3) CHANNEL FUNCTIONAL TEST(15) (9) MODES IN WHICH SURVEILLANCE REQUIRED 1,2,and* 1,2, and 3* 1,2 S/U(1) 1,2 and .. (16)and S/U(l) 2,3,4,5 and .. 1,2 1,2 1,2 1,2 1,2 1 SALEM -UNIT 3/43-11 Amendment No. 299 TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE FUNCTIONAL UNIT Loss of Flow Two Loops Steam Generator Water Level--Low-Low DELETED Undervoltage

-Reactor Coolant Pumps Underfrequency

-Reactor Coolant Pumps Turbine Trip a. Low Autostop Oil Pressure b. Turbine Stop Valve Closure 19. Safety Illjection Input from ESF Reactor Coolant Pump Breaker Position Trip Reactor Trip Breaker 22. Automatic Trip Logic N.A. NA NA NA NA N.A. CHANNEL CALI BRA TION(15) NA N.A. NA NA NA NA MODES IN CHANNEL WHICH FUNCTIONAL SURVEILLANCE TEST(15) REQUIRED NA 1,2 1 1 S/U(1) 1,2 S/U(1) 1,2 (4)(5) 1,2 (5)(11 )(13) 1,2 and * (14) (5) 1,2 and

SALEM -UNIT 3/43-12 Amendment No. 299 TABLE 4.3-1

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

(1) If not performed in previous 31 days. (2) Heat balance only, above 15% of RATED THERMAL POWER. (3) Compare incore to excore axial offset above 15% of RATED THERMAL POWER. Recalibrate if absolute difference 2: 3 percent. (4) Manual SSPS functional input check in accordance with the Surveillance Frequency Control Program. (5) Each train or logic channel shall be tested in accordance with the Surveillance Frequency Control Program. (6) -Neutron detectors may be excluded from CHANNEL CALIBRATION.

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

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

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

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

(12) DELETED (13) Verify operation of Bypass Breakers Shunt Trip function from local pushbutton while breaker is in the test position prior to placing breaker in service. (14) Perform a functional test of the Bypass Breakers UV. Attachment via the SSPS. (15) Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table. (16) At the frequency specified in the Surveillance Frequency Control Program. SALEM -UNIT 1 3/43-13 Amendment No. 299 INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2.1 The Engineered Safety Feature Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4. APPLICABILITY:

As shown in Table 3.3-3. ACTION: a. With an ESFAS instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3-4, declare the channel inoperable and apply the applicable ACTION requirement of Table 3.3-3 until the channel is restored to OPERABLE status with the trip setpoint adjusted consistent with the Trip Setpoint value. b. With an ESFAS instrumentation channel inoperable, take the ACTION shown in Table 3.3-3. SURVEILLANCE REQUIREMENTS 4.3.2.1.1 Each ESFAS instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies specified in the Surveillance Frequency Control Program unless otherwise noted in Table 4.3-2. 4.3.2.1.2 The logic for the interlocks shall be demonstrated OPERABLE during the automatic actuation logic test. The total interlock function shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program during CHANNEL CALIBRATION testing of each channel affected by interlock operation.

4.3.2.1.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be verified to be within the limit in accordance with the Surveillance Frequency Control Program. The provisions of Specification 4.0.4 are not applicable to MSIV closure time testing. The provisions of Specification 4.0.4 are not applicable to the turbine driven auxiliary feedwater pump provided the surveillance is performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the secondary steam generator pressure is greater than 680 psig. SALEM -UNIT 1 3/4 3-14 Amendment No. 299 TABLE 4.3-2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS MODES IN CHANNEL WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK(7) CALIBRATION(7)

TEST(7) REQUIRED SAFETY INJECTION.

TURBINE TRIP AND FEEDWATER ISOLATION

a. Manual Initiation NA NA 1,2,3,4 b. Automatic Actuation Logic NA NA (2) 1,2,3,4 c. Containment Pressure-High (3) 1,2,3 d. Pressurizer Pressure--Low 1,2,3 e. Differential Pressure Between Steam Unes--High 1,2,3 f. Steam Flow in Two Steam Lines--High coincident with Tavg--Low-Low or Steam Line Pressure-Low 1,2,3 CONTAINMENT SPRAY Manual Initiation NA NA 1,2,3,4 Automatic Actuation Logic NA N.A. (2) 1,2,3,4 Containment (3) 1,2,3 Pressu re--H i9 h-H ig h SALEM -UNIT 3/43-31a Amendment No. 299 TABLE 4.3-2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM SURVEILLANCE MODES IN CHANNEL WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK(7) CALIBRATION(?)

TEST(?) REQUIRED CONTAINMENT ISOLATION Phase "A" Isolation Manual NA NA 1,2,3,4 From Safety Injection NA NA (2) 1,2,3,4 Automatic Actuation Logic Phase "8" Isolation Manual NA NA 1,2,3,4 Automatic Actuation NA NA (2) 1,2,3,4 Logic Containment Pressure-(3) 1,2,3 -High-High Containment Manual NA NA 1,2,3,4 Automatic Actuation NA NA (2) 1,2,3,4 Logic Containment Per Surveillance Requirement Atmosphere Radioactivity

-SALEM -UNIT 3/43-32 Amendment No. 299 TABLE 4.3-2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM SURVEILLANCE MODES IN CHANNEL WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK(7) CALI BRATION(7)

TEST(7) REQUIRED STEAM LINE ISOLATION Manual N.A. N.A. 1,2,3** Automatic Actuation Logic N.A. N.A. (2) 1,2,3 Containment (3) 1,2,3 Pressure--High-High Steam Flow in Two Steam 1,2,3 Lines--High Coincident with Tavg--Low-Low or Steam Line Pressure--Low TURBINE TRIP AND FEEDWATER ISOLATION Steam Generator Water 1,2,3 Level--High-High SAFEGUARDS EQUIPMENT CONTROL SYSTEM (SEC) LOGIC a. N.A. N.A. (6) 1,2,3,4 b. Logic, Timing and Outputs

N.A. N.A. (1 ) 1,2,3,4 UNDERVOLTAGE, VITAL BUS Loss of Voltage 1,2,3 Sustained Degraded 1,2,3 Voltage SALEM -UNIT 3/43-32a Amendment No. 299 TABLE 4.3-2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM SURVEILLANCE REQUIREMENTS FUNCTIONAL UNIT CHANNEL CHECK(7) CHANNEL CALIBRATION(7)

CHANNEL FUNCTIONAL TEST(7) MODES IN WHICH SURVEILLANCE REQUIRED 8. AUXILIARY FEEDWATER

a. Automatic Actuation Logic NA NA (2) 1,2,3 b. NOT USED c. Steam Generator Water Level--Low-Low 1,2,3 d. Undervoltage

-RCP 1,2 e, S.1. See 1 above (All S.1. surveillance requirements)

f. Trip of Main Feedwater Pumps NA NA g. Station Blackout See 6b and 7 above (SEC and UN Vital Bus) SALEM -UNIT 1 3/43-33 Amendment No. 299 TABLE 4.3-2 TABLE

Outputs are up to, but not including, the output relays. ** The provisions of Specification 4.0.4 are not applicable.

(1 ) Each logic channel shall be tested in accordance with the Surveillance Frequency Control Program. The CHANNEL FUNCTION TEST of each logic channel shall verify that its associated diesel generator automatic load sequence timer is OPERABLE with the interval between each load block within 1 second of its design interval.

(2) Each train or logic channel shall be tested in accordance with the Surveillance Frequency Control Program. (3) The CHANNEL FUNCTIONAL TEST shall include exercising the transmitter by applying either a vacuum or pressure to the appropriate side of the transmitter.

(4) NOT USED (5) NOT USED (6) Inputs from Undervoltage, Vital Bus, shall be tested in accordance with the Surveillance Frequency Control Program. Inputs from Solid State Protection System shall be tested in accordance with the Surveillance Frequency Control Program. (7) Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table. SALEM -UNIT 1 3/43-34 Amendment No. 299 INSTRUMENTATION 3/4.3.3 MONITORING INSTRUMENTATION RADIATION MONITORING INSTRUMENTATION 3.3.3.1 The radiation monitoring instrumentation channels shown in Table 3.3-6 shall be OPERABLE with their alarm/trip setpoints within the specified limits. APPLICABILITY:

As shown in Table 3.3-6. ACTION: a. With a radiation monitoring channel alarm/trip setpoint exceeding the value shown in Table 3.3-6, adjust the setpoint to within the limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or declare the channel inoperable.

b. With one or more radiation monitoring channels inoperable, take the ACTION shown in Table 3.3-6. c. The provisions of Specification 3.0.3 are not applicable.

4.3.3.1 Each radiation monitoring instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations in accordance with the Surveillance Frequency Control Program. SALEM -UNIT 1 3/43-35 Amendment No. 299 TABLE SALEM -UNIT 1 3/43-38 Amendment No.

REMOTE SHUTDOWN INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.5 The remote shutdown monitoring instrumentation channels shown in Table 3.3-9 shall be OPERABLE with readouts displayed external to the control room. APPLICABILITY:

MODES 1, 2 and 3. ACTION: With the number of OPERABLE remote shutdown monitoring channels less than required by Table 3.3-9, either restore the inoperable channel to OPERABLE status within 30 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . SURVEILLANCE REQUIREMENTS 4.3.3.5 Each remote shutdown monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, and CHANNEL CALIBRATION operations in accordance with the Surveillance Frequency Control Program. SALEM -UNIT 3/43-46 Amendment No. 299 TABLE SALEM -UNIT 1 3/43-48 Amendment No.

INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.7 The accident monitoring instrumentation channels shown in Table 3.3-11 shall be operable.

APPLICABILITY:

MODES 1,2, and 3. ACTION: a. As shown in Table 3.3-11. SURVEILLANCE REQUIREMENTS 4.3.3.7 Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies specified in the Surveillance Frequency Control Program unless otherwise noted in Table 4.3-11. SALEM -UNIT 1 3/43-53 Amendment No. 299 TABLE SURVEILLANCE REQUIREMENTS ACCIDENT MONITORING CHANNEL CHANNEL CHANNEL FUNCTIONAL INSTRUMENT CHECK(1) CALI BRA TION(1) TEST(1) Reactor Coolant Outlet Temperature

-N.A. T HOT (Wide Range) Reactor Coolant Inlet Temperature

-N.A. T COLD (Wide Range) Reactor Coolant Pressure (Wide Range) N.A. Pressurizer Water Level N.A. Steam Line Pressure N.A. Steam Generator Water Level N.A. (Narrow Range) Steam Generator Water Level N.A. (Wide Range) Refueling Water Storage Tank Water Level N.A. deleted 10. Auxiliary Feedwater Flow Rate N.A. 11. Reactor Coolant System N.A* N.A. Margin Monitor # Auxiliary Feedwater System is used on each startup and flow rate indication is verified at that time.

The instruments used to develop RCS subcooling margin are calibrated in accordance with the Surveillance Frequency Control Program; the monitor will be compared with calculated subcooling margin for known input values in accordance with the Surveillance Frequency Control Program. SALEM -UNIT 3/43-57 Amendment No. 299 TABLE 4.3-11 SURVEILLANCE REQUIREMENTS ACCIDENT MONITORING CHANNEL CHANNEL CHANNEL FUNCTIONAL INSTRUMENT CHECK(1) CALIBRATION(1)

TEST(1) 12. PORV Position N.A. 13. PORV Block Valve Position N.A.

14. Pressurizer Safety Valve N.A. Indicator

15. Containment Pressure -Narrow N.A. 16. Containment Pressure -Wide N.A. 17. Containment Water Level -Wide N.A. 18. Core Exit I'J.A. 19. Reactor Vessel Level Instrumentation N.A. System (RVLlS) 20. Containment High Range Accident Radiation Monitor 21. Main Steamline Discharge (Safety Valves and Atmospheric Steam Dumps) Monitor Table Notation (1 ) Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.

Unless the block valve is closed in order to meet the requirements of Action b, or c in specification 3.4.3. SALEM -UNIT 3/43-57a Amendment No. 299 INSTRUMENTATION RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 3.3.3.8 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3-12 shall be operable to ensure that the limits of ODCM Control 3.11.1.1 are not exceeded.

APPLICABILITY:

At all times. ACTION: Not Used With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-12. Exert best efforts to return the instrument to OPERABLE status within 30 days and, if unsuccessful, explain in the next annual radioactive effluent release report why the inoperability was not corrected in a timely manner. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.8 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies specified in the Surveillance Frequency Control Program unless otherwise noted in Table 4.3-12. SALEM -UNIT 3/43-58 Amendment No. 299 TABLE 4.3-12 RADIOACTIVE LIQUID EFFLUENT MONITORING SURVEILLANCE CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONAL INSTRUMENT CHECK(1) CHECK(1)

CALIBRATION(1)

TEST(1) Not Used Not Used Not Used TANK LEVEL INDICATING DEVICES** Temporary Outside Storage D* Tanks as TABLE NOTATION

During liquid additions to the ** If tank level indication is not provided, verification will be done by visual (1 ) Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table. SALEM -UNIT 3/43-60 Amendment No. 299 I NSTRUMENTA POWER DISTRIBUTION MONITORING LIMITING CONDITION FOR OPERATION APPLICABIL TY: MODE 1, above 25% RATED THERMAL POWER With any of the operability criteria listed in 3.3.3.14.a, 3.3.3.14.b, or 3.3.3.14.c not met, correct the deficient operability condition, or declare the PDMS inoperable and use the incore movable detector system to obtain any required core power distribution measurements.

Increase the measured core peaking factors using the values listed in the COLR for the PDMS inoperable condition.

The provisions of Specification 3.0.3 are not applicable.

4.3.3.14.1 The operability criteria listed in 3.3.3.14.a, 3.3.3.14.b, and 3.3.3.14.c shall be verified to be satisfied prior to acceptance of the PDMS core power distribution measurement results. 4.3.3.14.2 Calibration of the PDMS is required: In accordance with the Surveillance Frequency Control Program when the minimum number and core coverage criteria as defined in 3.3.3.14.b.1 and 3.3.3.14.b.2 are satisfied, or In accordance with the Surveillance Frequency Control Program when only the minimum number criterion as defined in 3.3.3.14.b.3 is satisfied.

SALEM -UNIT 3/43-71 Amendment No. 299 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS NORMAL OPERATION 3.4.1.1 All reactor coolant loops shall be in APPLICABILITY:

MODES 1 and With less than the above required reactor coolant loops in operation, be in at least STANDBY within 1 4.4.1.1 The above required reactor coolant loops shall be verified to be in operation and circulating reactor coolant in accordance with the Surveillance Frequency Control Program.

See Special Test Exception 3.10.4. SALEM -UNIT 1 3/44-1 Amendment No. 299 REACTOR COOLANT SYSTEM HOT STANDBY SURVEILLANCE REQUIREMENTS 4.4.1.2.1 At least the above required reactor coolant pumps, if not in operation, shall be determined to be OPERABLE in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignments and indicated power availability.

4.4.1.2.2 At least one cooling loop shall be verified to be in operation and circulating reactor coolant in accordance with the Surveillance Frequency Control Program. 4.4.1.2.3 The required steam generator{s) shall be determined OPERABLE by verifying secondary side water level to be greater than or equal to 5% (narrow range) in accordance with the Surveillance Frequency Control Program. All reactor coolant pumps may be de-energized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided:

(1) no operations are permitted that would cause dilution of the reactor coolant system boron concentration (2) core outlet temperature is maintained at least 10°F below saturation temperature, and (3) the rod control system is de-energized** The rod control system shall be considered de-energized when one or more of the following conditions exist: Both Rod Drive MG set motor breakers are open. Both Rod Drive MG set generator breakers are open. A combination of at least three of the Reactor Trip and/or Reactor Trip Bypass Breakers are open. If none of the above conditions for de-energizing the rod control system are met; the system shall be considered energized.

SALEM -UNIT 3/44-2a Amendment No. 299 REACTOR COOLANT 4.4.1.3.1 The required residual heat removalloop(s) shall be determined OPERABLE per the inservice testing schedule.

4.4.1.3.2 The required reactor coolant pump(s), if not in operation, shall be determined to be OPERABLE in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignments and indicated power availability.

4.4.1.3.3 The required steam generator(s) shall be determined OPERABLE by verifying secondary side level to be greater than or equal to 5% of narrow range in accordance with the Surveillance Frequency Control Program. 4.4.1.3.4 At least one coolant loop shall be verified to be in operation and circulating reactor coolant in accordance with the Surveillance Frequency Control Program. SALEM -UN IT 1 3/44-3a Amendment No. 299 REACTOR COOLANT SYSTEM COLD SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.4 Two# residual heat removal loops shall be OPERABLE*

and at least one RHR loop shall be in operation.**

APPLICABILITY:

MODE 5.## ACTION: a. With less than the above required loops operable, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible.

b. With no RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation.

SURVEILLANCE REQUIREMENTS 4.4.1.4 At least one residual heat removal loop shall be verified to be in operation and circulating reactor coolant in accordance with the Surveillance Frequency Control Program. One RHR loop may be inoperable for up to two hours for surveillance testing, provided the other RHR loop is OPERABLE and in operation.

Additionally, four filled reactor coolant loops, with at least two steam generators with their secondary side water levels greater than or equal to 5% (narrow range), may be substituted for one residual heat removal loop. A reactor coolant pump shall not be started with one or more of the RCS cold leg temperatures less than or equal to 312°F unless 1) the pressurizer water volume is less than 1650 cubic feet (equivalent to approximately 93.2% of level), or 2) the secondary water temperature of each steam generator is less than 50°F above each of the RCS cold leg temperatures. Systems supporting RHR loop operability may be excepted as follows: The normal or emergency power source may be inoperable. The residual heat removal pumps may be de-energized for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months provided 1) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and 2) core outlet temperature is maintained at least 10°F below saturation temperature.

SALEM -UNIT 3/44-3b Amendment No. 299 REACTOR COOLANT SYSTEM 3/4.4.3 RELIEF VALVES SURVEILLANCE REQUIREMENTS 4.4.3.1 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: Operating the PORV through one complete cycle of full travel during MODES 3 or 4, and 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 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 3/44-5a Amendment No. 299 REACTOR COOLANT SYSTEM 3/4.4.4 PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.4 The pressurizer shall be OPERABLE with a water volume of less than or equal to 1650 cubic feet (92% indicated level), and at least two groups of pressurizer heaters each having a capacity of 150 kw and capable of being powered from an emergency power supply. APPLICABILITY:

MODES 1, 2, and 3. ACTION: a. With the pressurizer inoperable due to an inoperable emergency power supply to the pressurizer heaters either restore the inoperable emergency power supply 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 and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . b. With the pressurizer otherwise inoperable, be in at least HOT STANDBY with the reactor trip breakers open within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . 4.4.4.1 The pressurizer water volume shall be determined to be within its limit in accordance with the Surveillance Frequency Control Program. 4.4.4.2 The capacity of each of the above required groups of pressurizer heaters shall be verified by measuring circuit current in accordance with the Surveillance Frequency Control Program. 4.4.4.3 The emergency power supply for the pressurizer heaters shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by manually transferring power from the normal to the emergency power supply and energizing the heaters. SALEM -UNIT 1 3/4 4-6 Amendment No. 299 REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.6.1 The following Reactor Coolant System leakage detection systems shall be OPERABLE:

a. The containment atmosphere particulate radioactivity monitoring system, b. The containment sump level monitoring system, and c. Either the containment fan cooler condensate flow rate or the containment atmosphere gaseous radioactivity monitoring system. APPLICABILITY:

MODES 1, 2, 3 and 4. ACTION: With only two of the above required leakage detection systems OPERABLE, operation may continue for up to 30 days provided grab samples of the containment atmosphere are obtained and analyzed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the required gaseous and/or particulate radioactivity monitoring system is inoperable; 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 . 4.4.6.1 The leakage detection systems shall be demonstrated OPERABLE by: a. Containment atmosphere particulate and gaseous (if being used) monitoring systems-performance of CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST in accordance with the Surveillance Frequency Control Program. b. Containment sump level and containment fan cooler condensate flow rate (if being used) monitoring systems-performance of CHANNEL CALIBRATION in accordance with the Surveillance Frequency Control Program. SALEM -UNIT 1 3/44-14 Amendment No. 299 REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR QPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to: No PRESSURE BOUNDARY LEAKAGE, 1 GPM UNIDENTIFIED LEAKAGE, 150 gallons per day primary-to-secondary leakage through anyone steam generator, and d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System. APPLICABILITY:

MODES 1, 2, 3 and 4 ACTION: With any PRESSURE BOUNDARY LEAKAGE or primary-to-secondary leakage not within limit, be in at least HOT STANDBY within 6 hours6.944444e-5 days 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 . With any Reactor Coolant System leakage greater than anyone of the above limits, excluding PRESSURE BOUNDARY LEAKAGE and primary-to-secondary leakage, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 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 and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . SURVEILLANCE 4.4.6.2 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by; Monitoring the containment atmosphere particulate radioactivity monitor in accordance with the Surveillance Frequency Control Program. Monitoring the containment sump inventory in accordance with the Surveillance Frequency Control Program. SALEM -UNIT 3/44-15 Amendment No. 299 REACTOR COOLANT SURVEILLANCE REQUIREMENTS (Continued)

c.

Verifying primary-to-secondary leakage is :5 150 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. 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 3/44-16 Amendment No. 299 REACTOR COOLANT PRIMARY COOLANT SYSTEM PRESSURE ISOLATION VALVES LIMITING QONDITION 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: 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 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: In accordance with the Surveillance Frequency Control Program. 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. Prior to returning the valve to service following maintenance repair or replacement work on the valve. 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 3/44-16a Amendment No. 299 TABLE PRIMARY COOLANT SPECIFIC ACTIVITY AND ANALYSIS MODES IN WHICH SAMPLE AND TYPE OF MEASUREMENT SAMPLE AND ANALYSIS ANALYSIS AND ANALYSIS FREQUENCY REQUIRED Gross Activity Determination In accordance with the 1,2,3,4 Surveillance Frequency Control Program 2. Isotopic Analysis for In accordance with the 1 EQUIVALENT 1-131 Concentration Surveillance Control Radiochemical for E Determination In accordance with the 1 Surveillance Frequency Control Program* 4. Isotopic Analysis for a) Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , Including 1-131, 1-133, and 1-135 whenever the activity 1.0 IJCi/gram DOSE EQUIVALENT 1-131 or 1 ODIE IJCi/gram, and One sample between 1,2,3 2 &6 hours following a THERMAL POWER change exceeding 15 percent of the RATED THERMAL POWER within a one hour period. Until the specific activity of the primary coolant system is restored within its limits. Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last subcritical for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or longer. SALEM -UNIT 3/44-22 Amendment No. 299 REACTOR COOLANT SYSTEM SURVEILLANCE Rl;QUIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits in accordance with the Surveillance Frequency Control Program during system heatup, cooldown, and inservice leak and hydrostatic testing operations.

4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, at the intervals required by 10 CFR 50, Appendix H. The results of these examinations shall be used to update Figures 3.4-2 and 3.4-3. SALEM -UNIT 1 3/44-25 Amendment No. 299 REACTOR COOLANT SYSTEM PRESSURIZER 3.4.9.2 The pressurizer temperature shall be limited to: a. A maximum heatup of 100°F in anyone hour period, b. A maximum cooldown of 200°F in anyone hour period, and c. A maximum spray water temperature differential of 320°F. APPLICABILITY:

At all times. ACTION: With the pressurizer temperature limits in excess of any of the above limits, restore the temperature to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the fracture toughness properties of the pressurizer; determine that the pressurizer remains acceptable for continued operation 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 reduce the pressurizer pressure to less than 500 psig within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . SURVEILLANCE REQUIREMENTS 4.4.9.2 The pressurizer temperatures shall be determined to be within the limits in accordance with the Surveillance Frequency Control Program during system heatup or cooldown.

The spray water temperature differential shall be determined to be within the limit in accordance with the Surveillance Frequency Control Program during auxiliary spray operation.

SALEM -UNIT 1 3/44-29 Amendment No. 299 REACTOR COOLANT OVERPRESSURE PROTECTION SYSTEMS REQUI REMENTS 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. Performance of a CHANNEL CALIBRATION on the POPS actuation channel in accordance with the Surveillance Frequency Control Program. Verifying the POPS isolation valve is open in accordance with the Surveillance Frequency Control Program when the POPS is being used for overpressure protection. Testing pursuant to the 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 3/44-31 Amendment No. 299 REACTOR COOLANT SYSTEM 3/4.4.12 HEAD VENTS 3.4.12 Four reactor vessel head vent paths shall be operable with the vent paths closed. A vent path consists of at least two head vent valves in series, powered from vital sources, and associated flowpath.

APPLICABILITY:

MODES 1, 2, 3 AND 4. ACTION: With one, two or three reactor vessel head vent path(s) inoperable, STARTUP and/or POWER OPERATION may continue provided the inoperable vent path(s) is maintained closed with the valve actuators key locked in the closed position; restore the inoperable vent path(s) to OPERABLE status within 30 days, or, be in HOT STANDBY within 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 . With four reactor vessel head vent paths inoperable; maintain the inoperable vent paths closed with power removed from the valve actuators of all the vent valves in the inoperable vent paths, and restore at least one of the vent paths to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY within 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 4.4.12 Reactor vessel head vent system vent paths shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by: Verifying all manual isolation valves in each vent path are locked in the open position. Cycling each valve in the vent paths through at least one complete cycle of full travel from the control room during COLD SHUTDOWN or REFUELING. Verifying flow through the reactor vessel head vent system vent path during venting. SALEM -UNIT 3/44-35 Amendment No. 299 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) ACCUMULATORS LIMITING CONDITION FOR OPERATION 3.5.1 Each reactor coolant system accumulator shall be OPERABLE with: The isolation valve open, A contained volume of between 6,223 and 6,500 gallons of borated water, A boron concentration of between 2,200 and 2,500 ppm, and, d. A nitrogen cover-pressure of between 595.5 and 647.5 psig. APPLICABILITY:

MODES 1,2 and 3*. ACTION: With one accumulator inoperable, except as a result of a closed isolation valve or boron concentration outside the required limits, restore the inoperable accumulator to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . With one accumulator inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in HOT STANDBY within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . With the boron concentration of one accumulator outside the required limits, restore the boron concentration to within the required limits 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 and reduce pressurizer pressure to less than or equal to 1000 psig within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . SURVEILLANCE REQUIREMENTS 4.5.1 Each accumulator shall be demonstrated OPERABLE: In accordance with the Surveillance Frequency Control Program by: Verifying the water level and nitrogen cover-pressure in the tanks, and Verifying that each accumulator isolation valve is open.

Pressurizer Pressure above 1000 psig. SALEM -UNIT 3/45-1 Amendment No. 299 EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

b. In accordance with the Surveillance Frequency Control Program and within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after each solution volume increase of 1% of tank volume by verifying the boron concentration of the accumulator solution.

c. In accordance with the Surveillance Frequency Control Program when the RCS pressure is greater than 1000 psig by verifying that the power lockout switch is in lockout. d. In accordance with the Surveillance Frequency Control Program by verifying that each accumulator isolation valve opens automatically upon receipt of a safety injection test signal. SALEM -UNIT 1 3/4 5-2 Amendment No. 299 EIVIERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE: In accordance with the Surveillance Frequency Control Program by: Verifying that the following valves are in the indicated positions with power to the valve operators removed: Valve Number Valve Valve Position a. 1 SJ 69 a. RHR pump suction a. open b. 1 SJ 30 b. SI pump suction b. open c. 11 SJ 40 c. SI discharge to hot legs c. closed d. 12 SJ 40 d. SI discharge to hot legs d. closed e. 1 RH 26 e. RHR discharge to hot legs e. closed f. 11 SJ 49 f. RHR discharge to cold legs f. open g. 12 SJ 49 g. RHR discharge to cold legs g. open h. 1 CS 14# h. Spray additive tank discharge

h. open i. 1 SJ 135 i. SI discharge to cold legs i. open j. 1 SJ 67 j. SI recirc. line isolation

j. open k. 1 SJ 68 k. SI recirc. line isolation

k. open I. 11 SJ 44 I. Containment sump isolation valve I. closed m. 12 SJ 44 m. Containment sump isolation valve m. closed Verifying that the following valves are in the indicated positions:

Valve Number Valve Valve Position a. RH 19 a. RHR crosstie valve a. open b. 12 RH 19 b. RHR crosstie b. open 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. Verifying that the ECCS piping is full of water by venting the ECCS pump casings and accessible discharge piping high points. # If inoperable, the applicable Technical Specification is SALEM -UNIT 1 Amendment No.

EMERGENCY CORE COOLING SURVEI LLANCEREQUI REMENTS (Continued) By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the containment sump and cause restriction of the pump suctions during LOCA conditions.

This visual inspection shall be performed: For all accessible areas of the containment prior to establishing CONTAINMENT INTEGRITY, and In accordance with the Surveillance Frequency Control Program the areas affected within containment by containment entry and during the final entry when CONTAINMENT INTEGRITY is established. In accordance with the Surveillance Frequency Control Program by: A visual inspection of the containment sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or corrosion. In accordance with the Surveillance Frequency Control Program, during shutdown, by: Verifying that each automatic valve in the flow path actuates to its correct position on a safety injection test signal. Verifying that each of the following pumps start automatically upon receipt of a safety injection test signal: a) Centrifugal charging pump b) Safety injection pump c) Residual heat removal pump SALEM -UNIT 3/45-5 Amendment No. 299 EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 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: 1. Centrifugal charging pump 2338 psi TDH 2. Safety Injection Pump 1369 psi TDH 3. Residual heat removal pump 165 psi TDH By verifying the correct position of each of the following ECCS throttle valves: 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. 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 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: For Safety Injection pumps, with a single pump running: The sum of the injection line flow rates, excluding the highest flow rate, is 453 gpm; and The total flow rate through all four injection lines is s 647 gpm, and The difference between any pair of injection line flow rates is s 12.0 gpm, and The total pump flow rate is ::; 664 gpm in the cold leg alignment, and The total pump flow rate is s 654 gpm in the hot leg alignment.

SALEM -UNIT 3/45-5a Amendment No. 299 EMERGENCY CORE COOLING ECCS SUBSYSTEMS

-Tavg < 350°F SURVEILLANCE REQUIREMENTS 4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per applicable Surveillance Requirements of 4.5.2. 4.5.3.2 All safety injection pumps and centrifugal charging pumps, except the above required OPERABLE pump, shall be demonstrated to be inoperable in accordance with the Surveillance Frequency Control Program while in MODE 4 and 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 by either of the following methods: a. By verifying that the motor circuit breakers have been removed from their electrical power supply circuits or, b. For verifying that the pump is in a recirculation ftow path and that two independent means of preventing RCS injection are utilized.

SALEM -UNIT 1 3/45-6a Amendment No. 299 EMERGENCY CORE COOLING SYSTEMS SEAL INJECTION FLOW LIMITING CONDITION FOR OPERATION 3.5.4 Reactor coolant pump seal injection flow shall be :s; 40 gpm with centrifugal charging pump discharge header pressure 2!: 2430 psig and the charging flow control valve full open. APPLICABILITY:

MODES 1. 2. and 3 ACTION: With seal injection flow not within the limit. adjust manual seal injection throttle valves to give a flow within the limit with the charging pump discharge pressure 2!: 2430 psig and the charging flow control valve full open within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 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 and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . 4.5.4 In accordance with the Surveillance Frequency Control Program. verify manual seal injection throttle valves are adjusted to give a flow within the limit with centrifugal charging pump discharge header pressure 2!: 2430 psig, and the charging flow control valve full open. The provisions of Specification 4.0.4 are not applicable for entry into Mode 3. This exemption is allowed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after the Reactor Coolant System pressure stabilizes at 2235 +/- 20 psig. SALEM -UNIT 1 3/45-6b Amendment No. 299 EMERGENCY CORE COOLING SYSTEMS REFUELING WATER STORAGE TANK CONDIIION FOR OPERATION 3.5.5 The refueling water storage tank (RWST) shall be OPERABLE with: a. A contained volume of between 364,500 and 400,000 gallons of borated water. b. A boron concentration of between 2300 and 2500 ppm, and c. A minimum water temperature of 35°F. APPLICABILITY:

MODES 1.2.3 and 4. ACTION: With the refueling water storage tank inoperable.

restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within 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 . 4.5.5 The RWST shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by: 1. Verifying the water level in the tank. and 2. Verifying the boron concentration of the water. b. In accordance with the Surveillance Frequency Control Program by verifying the RWST temperature when the outside air temperature is < 35°F. SALEM -UNIT 1 3/45-7 Amendment No. 299 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATIOf'J 3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY:

MODES 1, 2, 3 and 4. ACTION: Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within one hour 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 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated: In accordance with the Surveillance Frequency Control Program by verifying that each containment manual valve or blind flange that is located outside containment and required to be closed during accident conditions is closed, except for containment isolation valves that are open under administrative controls.

Valves and blind flanges in high radiation areas may be verified by use of administrative controls. Prior to entering Mode 4 from Mode 5 if not performed within the last 92 days by verifying that each containment manual valve or blind flange that is located inside containment and required to be closed during accident conditions is closed, except for containment isolation valves that are open under administrative controls.

Valves and blind flanges in high radiation areas may be verified by use of administrative controls. By verifying that each containment air lock is OPERABLE per Specification 3.6.1.3. In accordance with the Surveillance Frequency Control Program by verifying that the surveillance requirements of 4.6.2.3.a are met for penetrations associated with the containment fan coil units. In accordance with the Surveillance Frequency Control Program by verifying that the surveillance requirements of 4.6.2.3.d are met for penetrations associated with the containment fan coil units. SALEM UNIT 3/46-1 Amendment No. 299 CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATIONS (Continued) One or more containment air locks inoperable for reasons other than condition

a. or b. Immediately initiate action to evaluate overall containment leakage per LCO 3.6.1, and: Verify that at least one door is closed in the affected air lock within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and: Restore the air lock to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . If the ACTIONS and associated completion times of a., b., or c. cannot be met, be in Hot Standby within 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 . 4.6.1.3 Each containment air lock shall be demonstrated OPERABLE:

a. By verifying seal leakage rate in accordance with the Containment Leakage Rate Testing program. b. By conducting an overall air lock leakage test in accordance with the Containment Leakage Rate Testing Program. c. In accordance with the Surveillance Frequency Control Program by verifying that only one door in each air lock can be opened at a time. SALEM -UNIT 3/46-5a Amendment No. 299 CONTAI NMENT SYSTEMS INTERNAL PRESSURE LIMITING CONDITION FOR OPERATION 3.6.1.4 Primary containment internal pressure shall be maintained between -1.5 and +0.3 APPLICABILITY:

MODES 1, 2, 3 and With the containment internal pressure outside of the limits above, restore the internal to within the limits within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 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 and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . 4.6.1.4 The primary containment internal pressure shall be determined to be within the limits in accordance with the Surveillance Frequency Control Program. SALEM -UNIT 1 3/4 6-6 Amendment No. 299 CONTAINMENT SYSTEMS AIR TEMPERATURE LIMITING CONDITION FOR OPERATION 3.6.1.5 Primary containment average air temperature shall not exceed APPLICABILITY:

MODES 1,2,3 and With the containment average air temperature>

120°F, reduce the average air temperature within the limit within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 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 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 4.6.1.5 Verify the Containment Average Air Temperature is within limit in accordance with the Surveillance Frequency Control Program. SALEM -UNIT 1 3/4 6-7 Amendment 299 CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM LIMITING CONDITION 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, 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 SPRAY ADDITIVE SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.2 The spray additive system shall be OPERABLE with: a. A spray additive tank containing a volume of between 2568 and 4000 gallons of between 30 and 32 percent by weight NaOH solution, and b. Two spray additive eductors each capable of adding NaOH solution from the chemical additive tank to a containment spray system pump flow. APPLICABILITY:

MODES 1, 2, 3 and 4. ACTION: With the spray additive system inoperable, restore the 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 spray additive 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 4.6.2.2 The spray additive system shall be demonstrated OPERABLE: 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. In accordance with the Surveillance Frequency Control Program by: Verifying the solution level in the tank, and Verifying the concentration of the NaOH solution by chemical analYSis. In accordance with the Surveillance Frequency Control Program during shutdown, by verifying that each automatic valve in the flow path actuates to its correct position on a Containment High-High pressure test signal. In accordance with the Surveillance Frequency Control Program by: Verifying a NaOH solution flow rate of 12 +/- 3 gpm from the spray additive tank through sample valve 1 CS61 with the spray additive tank at 2.5 +/- 0.5 psig and SALEM -UNIT 3/46-10 Amendment No. 299 CONTAINMENT SURVEILLANCE REQUIREMENTS (Continued) In accordance with the Surveillance Frequency Control Program by: Verifying the water level in each service water accumulator vessel is greater than or equal to 226 inches and less than or equal to 252 inches. Verifying the temperature in each service water accumulator vessel is greater than or equal to 55"F and less than or equal to 95°F. Verifying the nitrogen cover pressure in each service water accumulator vessel is greater than or equal to 135 pSig and less than or equal to 160 pSig. In accordance with the Surveillance Frequency Control Program by: Starting (unless already operating) each fan from the control room in low speed. Verifying that each fan operates for at least 15 minutes in low speed. Verifying a cooling water flow rate of greater than or equal to 1300 gpm to each cooler. In accordance with the Surveillance Frequency Control Program by verifying that on a safety injection test signal: Each fan starts automatically in low speed. The automatic valves and dampers actuate to their correct positions and that the cooling water flow rate to each cooler is greater than or equal to 1300 gpm. In accordance with the Surveillance Frequency Control Program by verifying that on a loss of offsite power test Signal, each service water accumulator vessel discharge valve response time is within limits. SALEM -UNIT 3/46-11a Amendment No. 299 CONTAINMENT 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. 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. 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. 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: Required Containment Purge Supply and Exhaust Isolation Valves in accordance with the Surveillance Frequency Control Program. 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 3/46-13 Amendment No. 299 PLANT AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least three independent steam generator auxiliary feedwater pumps and associated manual activation switches in the control room and flow paths shall be OPERABLE with: a. Two feedwater pumps, each capable of being powered from separate vital busses, and b. One feedwater pump capable of being powered from an OPERABLE steam supply system. APPLICABILITY:

MODES 1, 2 and 3. ACTION: With one auxiliary feedwater pump inoperable, restore the required auxiliary feedwater 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 within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . With two auxiliary feedwater pumps inoperable be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . With three auxiliary feedwater pumps inoperable, immediately initiate corrective ACTION to restore at least one auxiliary feedwater pump to OPERABLE status as soon as possible. LCO 3.0A.b is not applicable.

4.7.1.2 Each auxiliary feedwater pump shall be demonstrated OPERABLE: In accordance with the Surveillance Frequency Control Program by: Verifying that each non-automatic valve in the flow path that is not locked, sealed or otherwise secured in position, is in its correct position. Verify the manual maintenance valves in the flow path to each steam generator are locked open. SALEM -UNIT 3/47-5 Amendment No. 299 PLANT SURVEILLANCE REQUIREMENTS (continued) In accordance with the Surveillance Frequency Control Program by: Verify that the developed head of each motor driven pump at the flow test point is greater than or equal to the required developed head. Verify that the developed head of the steam driven pump at the flow test point is greater than or equal to the required developed head when the steam generator pressure is > 680 psig. The provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after secondary side pressure is greater than 680 psig. In accordance with the Surveillance Frequency Control Program by: Verifying that each auxiliary feedwater automatic valve that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal. Verifying that each auxiliary feedwater pump starts automatically on an actual or simulated actuation signal. The provisions of Specification 4.0.4 are not applicable to the turbine driven auxiliary feedwater pump provided the surveillance is performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the secondary steam generator pressure is greater than 680 psig. SALEM -UNIT 3/4 7-6 Amendment No. 299 PLANT AUXILIARY FEED STORAGE TANK LIMITING CONDITION FOR OPERATION 3.7.1.3 The auxiliary feed storage tank (AFST) shall be OPERABLE with a minimum volume of 200,000 gallons of APPLICABILITY:

MODES 1, 2 and With the auxiliary feed storage tank inoperable, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months a. Restore the AFST to OPERABLE status or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , or b. Demonstrate the OPERABILITY of a demineralized water or a fire protection/domestic water storage tank as a backup supply to the auxiliary feedwater pumps and restore the auxiliary feed storage tank to OPERABLE status within 7 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . 4.7.1.3.1 The auxiliary feed storage tank shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by verifying the water level is within its limits when the tank is the supply source for the auxiliary feedwater pumps. 4.7.1.3.2 A demineralized water storage tank shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by verifying the tank contains;::

200,000 gallons of water and by verifying proper alignment of valves for taking suction from this tank when it is the supply source for the auxiliary feedwater pumps. 4.7.1.3.3 A fire protection/domestic water storage tank shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by verifying the tank contains ;:: 200,000 gallons of water and by verifying proper alignment of valves for taking suction from this tank when it is the supply source for the auxiliary feedwater pumps. SALEM -UNIT 1 3/4 7-7 Amendment No. 299 TABLE SECONDARY COOLANT SYSTEM SPECIFIC SAMPLE AND ANALYSIS TYPE OF MEASUREMENT AND ANALYSIS SAMPLE AND ANALYSIS FREQUENCY Gross Activity Determination In accordance with the Surveillance Frequency Control Program 2. Isotopic Analysis for DOSE a) In accordance with the Surveillance EQUIVALENT 1-131 Concentration Frequency Control Program, whenever the gross activity determination indicates iodine concentrations greater than 10% of the allowable limit. b) In accordance with the Surveillance Frequency Control Program, whenever the gross activity determination indicates iodine concentrations below 10% of the allowable limit. SALEM -UNIT 3/4 7-9 Amendment No. 299 PLANT SYSTEMS 3/4.7.2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION LIMITING CONDITION FOR OPERATION 3.7.2.1 The temperatures of both the primary and secondary coolants in the steam generators shall be> 70°F when the pressure of either coolant in the steam generator is> 200 psig. APPLICABILITY:

At all times. With the requirements of the above specification not Reduce the steam generator pressure of the applicable side to 200 psig within 30 minutes, and Perform an engineering evaluation to determine the effect of the overpressurization on the structural integrity of the steam generator.

Determine that the steam generator remains acceptable for continued operation prior to increasing its temperatures above 200°F. 4.7.2.1 The pressure in each side of the steam generator shall be determined to be < 200 psig in accordance with the Surveillance Frequency Control Program when the temperature of either the primary or secondary coolant is < 70°F. SALEM -UNIT 3/47-14 Amendment No. 299 PLANT SYSTEMS 3/4.7.3 COMPONENT COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.3.1 At least two independent component cooling water loops shall be APPLICABILITY:

MODES 1, 2, 3 and With only one component cooling water loop OPERABLE, restore at least two loops 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 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 4.7.3.1 At least two component cooling water loops shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by verifying that each valve (manual, power operated or automatic) servicing safety related equipment that is not locked, sealed, or otherwise secured in position, is in its correct position.

SALEM -UNIT 1 3/47-15 Amendment No. 299 PLANT SYSTEMS 3/4.7.4 SERVICE WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.4.1 At least two independent service water loops shall be APPLICABILITY:

MODES 1, 2, 3 and With only one service water loop OPERABLE, restore at least two loops to OPERABLE 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 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 4.7.4.1 At least two service water loops shall be demonstrated OPERABLE:

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

b. In accordance with the Surveillance Frequency Control Program during shutdown, by verifying that each automatic valve servicing safety related equipment actuates to its correct position on Safeguards Initiation signal.

Operation with only the 11 service water loop OPERABLE may continue for up to 10 days. This note is applicable for one time use during Salem Unit No.1 Cycle 15. SALEM -UNIT 1 3/47-16 Amendment No. 299 PLANT 3/4.7.5 FLOOD PROTECTION LIMITING CONDITION FOR OPERATION 3.7.5.1 Flood protection shall be provided for all safety related systems, components and structures when the water level of the Delaware River exceeds 10.5' Mean Sea Level USGS datum, at the service water intake structure.

APPLICABILITY:

At all times. ACTION: a. With the water level at the service water intake structure above elevation 10.5' Mean Sea Level USGS datum, close all watertight doors within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . b. With the water level at the service water intake structure above elevation 11.5' Mean Sea Level USGS datum, be in at least HOT STANDBY within 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 4.7.5.1 The water level at the service water intake structure shall be determined to be within the limits by: Measurement in accordance with the Surveillance Frequency Control Program when the water level is below elevation 10.5' Mean Sea Level USGS datum, and Measurement in accordance with the Surveillance Frequency Control Program when the water level is equal to or above elevation 10.5' Mean Sea Level USGS datum. SALEM UNIT 3/47-17 Amendment No. 299 PLANT 4.7.6.1 Each control room emergency air conditioning system filtration train shall be demonstrated OPERABLE: In accordance with the Surveillance Frequency Control Program by initiating flow through the HEPA filter and charcoal adsorber train(s) and verifying that the train(s) operates with each fan operating for at least 15 minutes. In accordance with the Surveillance Frequency Control Program or prior to return to service (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the system, by: Verifying that the charcoal adsorbers remove 2! 99% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place while operating the ventilation system at a flow rate of 8000 cfm +/- 10%. Verifying that the HEPA filter banks remove 2! 99% of the OOP when they are tested in-place while operating the ventilation system at a flow rate of 8000 cfm +/- 10%. Verifying within 31 days after removal from the CREACS unit, that a laboratory test of a sample of the charcoal adsorber, when obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 2.5% when tested in accordance with ASTM 03803-1989 at a temperature of 30"C and a relative humidity of 95%. After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation by verifying within 31 days after removal from the CREACS unit, that a laboratory analysis of a representative carbon sample, when obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows a methyl iodide penetration less than 2.5% when tested in accordance with ASTIVI 03803-1989 at a temperature of 30"C and a relative humidity of 95%. In accordance with the Surveillance Frequency Control Program by: Verifying that the pressure drop across the combined HEPA filter and charcoal adsorber bank is ::; 3.5 inches water gauge while operating the ventilation system at a flow rate of 8000 cfm +/- 10%. SALEM -UNIT 3/4 7-20 Amendment No. 299 PLANT SURVEILLANCE REQUIREMENTS 4.7.7.1 The above required Auxiliary Building Ventilation System shall be demonstrated OPERABLE by: a. In accordance with the Surveillance Frequency Control Program by verifying negative pressure in the Auxiliary Building.

b. In accordance with the Surveillance Frequency Control Program by starting each fan, from the control room, and verifying that each fan operates for at least 15 minutes. c. In accordance with the Surveillance Frequency Control Program by verifying that the System starts following a Safety Injection Test Signal. SALEM -UNIT 1 3/47-23 Amendment No. 299 PLANT 3/4.7.8 SEALED SOURCE CONTAMINATION LIMITING CONDITlQN FOR OPERATION 3.7.8.1 Each sealed source containing radioactive material either in excess of 100 microcuries of beta and/or gamma emitting material or 5 microcuries of alpha emitting material shall be free of;?; 0.005 microcuries of removable contamination.

APPLICABILITY:

At all times. ACTION: a. Each sealed source with removable contamination in excess of the above limits shall be immediately withdrawn from use and: 1. Either decontaminated and repaired, or 2. Disposed of in accordance with Commission Regulations.

b. The provisions of Specification 3.0.3 are not applicable.

4.7.8.1.1 Test Requirements

-Each sealed source shall be tested for leakage and/or contamination by: The licensee, or b. Other persons specifically authorized by the Commission or an Agreement State. The test method shall have a detection sensitivity of at least 0.005 microcuries per test sample. 4.7.8.1.2 Test Frequencies

-Each category of sealed sources shall be tested at the frequency described below. Sources in use (excluding startup sources and fission detectors previously subjected to core flux) -In accordance with the Surveillance Frequency Control Program for all sealed sources containing radioactive materials.

SALEM UNIT 3/47-26 Amendment No. 299 PLANT SYSTEMS LIMITING CONPITION FOR OPERATION (Continued)

ACTION: MODES 5 and 6 or during movement of irradiated fuel assemblies.

With one chiller inoperable: Remove the appropriate non-essential heat loads from the chilled water system within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and; Restore the chiller to OPERABLE status within 14 days or; Suspend CORE ALTERATIONS and movement of irradiated fuel assemblies. With two chillers inoperable: Remove the appropriate non-essential heat loads from the chilled water system within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and; Align the control room emergency air conditioning system (CREACs) for single filtration operation using the Salem Unit 2 train within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and; Restore at least one chiller to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or; Suspend CORE ALTERATIONS and movement of irradiated fuel assemblies. With one chilled water pump inoperable, restore the chilled water pump to OPERABLE status within 7 days or suspend CORE ALTERATIONS and movement of irradiated fuel assemblies.

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

a. In accordance with the Surveillance Frequency Control Program by verifying that each manual valve in the chilled water system flow path servicing safety related components that is not locked, sealed, or otherwise secured in position, is in its correct position.

b. In accordance with the Surveillance Frequency Control Program, by verifying that each automatic valve actuates to its correct position on a Safeguards Initiation signal. c. In accordance with the Surveillance Frequency Control Program by verifying that each chiller starts and runs. During Modes 5 and 6 and during movement of irradiated fuel assemblies, chilled water components are not considered to be inoperable solely on the basis that the backup emergency power source, diesel generator, is inoperable.

SALEM -UNIT 3/47-34 Amendment No. 299 PLANT SYSTEMS 3/4.7.11 FUEL STORAGE POOL BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.7.11 The fuel storage pool boron concentration shall be 800 ppm. APPLICABILITY:

When fuel assemblies are stored in the fuel storage pool and a fuel storage pool verification has not been performed since the last movement of fuel assemblies in the fuel storage pool. ACTION: With fuel storage pool boron concentration not within limit: a. Immediately suspend movement of fuel assemblies in the fuel storage pool and b. I nitiate action to: 1. immediately restore fuel storage pool boron concentration to within limit or 2. immediately perform a fuel storage pool verification.

c. LCO 3.0.3 is not applicable.

SURVEILLANCE REQUIREMENTS 4.7.11 Verify the fuel storage pool boron concentration is within limit in accordance with the Surveillance Frequency Control Program. SALEM -UNIT 1 3/47-35 Amendment No. 299 ELECTRICAL POWER 4.8.1.1.1 Two physically independent AC. circuits between the offsite transmission network and the onsite Class 1 E distribution system (vital bus system) shall be: Determined OPERABLE in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignments, power availability, and Demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program during shutdown by transferring (manually and automatically) vital bus supply from one 13/4 kv transformer to the other 13/4 kv transformer.

4.8.1.1.2 Each diesel generator shall be demonstrated OPERABLE: In accordance with the Surveillance Frequency Control Program by: Verifying the fuel level in its day tank. Verifying the diesel generator starts from standby conditions*

and achieves 3910 volts and 58.8 Hz in s 13 seconds, and subsequently achieves steady state voltage of 3910 and :S 4400 volts and frequency of 60 +/- 1.2 Hz. Subsequently, verifying the generator is synchronized with voltage maintained 3910 and s 4580 volts, gradually loaded to 2340-2600 kw**, and operates at a load of 2340-2600 kw for greater than or equal to 60 minutes. Verifying the diesel generator is aligned to provide standby power to the associated vital bus. In accordance with the Surveillance Frequency Control Program and after each operation of the diesel where the period of operation was greater than or equal to one hour by checking for and removing accumulated water from the day tanks. In accordance with the Surveillance Frequency Control Program by verifying the diesel generator starts from standby conditions*

and achieves;:::

3910 volts and 58.8 Hz in S 13 seconds, and subsequently achieves steady state voltage of 3910 and S 4400 volts and frequency of 60 +/- 1.2 Hz. The generator shall be synchronized to its emergency bus with voltage maintained 3910 and S 4580 volts, loaded to 2340-2600**

kw in less than or equal to 60 seconds, and operate at a load of 2340-2600 kw for at least 60 minutes. This test, if it is performed so it coincides with the testing required by Surveillance Requirement 4.8.1.1.2.a.2, may also serve to concurrently meet those requirements.

SALEM -UNIT 3/4 8-3 Amendment No. 299 ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) In accordance with the Surveillance Frequency Control Program during shutdown by: DELETED Verifying that, on rejection of a load greater than or equal to 820 kw, the voltage and frequency are restored to 2: 3910 and :s; 4400 volts and 60 +/- 1.2 Hz within 4 seconds, and subsequently achieves a steady state frequency of 2: 58.8 and :s; 60.5 Hz. Simulating a loss of offsite power by itself, and: Verifying de-energization of the vital bus and load shedding from the vital bus. Verifying the diesel starts on the auto-start signal*, energizes the vital bus with permanently connected loads within 13 seconds, energizes the auto-connected shutdown loads through the load sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with the shutdown loads. After energization of these loads, the steady state voltage and frequency of the vital bus shall be maintained at 2: 3910 and :s; 4400 volts and 2: 58.8 and :s; 60.5 Hz during this test. Verifying that on an ESF actuation test signal without loss of offsite power the diesel generator starts on the auto-start signal and operates on standby for greater than or equal to 5 minutes*.

The diesel generator shall achieve 2: 3910 volts and 2: 58.8 Hz in :s; 13 seconds, and subsequently achieves steady state voltage of 2: 3910 and :s; 4400 volts and frequency of 2: 58.8 and :s; 60.5 Hz. Not Used. Simulating a loss of offsite power in conjunction with an ESF actuation test signal, and: Verifying de-energization of the vital bus and load shedding from the vital bus. Verifying the diesel starts on the auto-start signal*, energizes the vital bus with permanently connected loads within 13 seconds, energizes the auto-connected emergency (accident) loads through the load sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with the emergency loads. After energization of these loads, the steady state voltage and frequency of the vital bus shall be maintained at 2: 3910 and :s; 4400 volts and 2: 58.8 and :s; 60.5 Hz during this test. SALEM -UNIT 3/4 8-4 Amendment No. 299 ELECTRICAL POWER Verifying that all nonessential automatic diesel generator trips (Le., other than engine overspeed, lube oil pressure low, 4 KV bus differential and generator differential), are automatically bypassed upon loss of voltage on the vital bus concurrent with a safety injection actuation signal. Deleted Verifying that the auto-connected loads to each diesel generator do not exceed the two hour rating of 2860 kw. Verifying that with the diesel generator operating in a test mode (connected to its bus), a simulated safety injection signal overrides the test mode by (1) returning the diesel generator to standby operation and (2) automatically energizing the emergency loads with offsite power. In accordance with the Surveillance Frequency Control Program or after any modifications which could affect diesel generator interdependence by starting all diesel generators simultaneously*, during shutdown, and verifying that all diesel generators accelerate to at least 58.8 Hz in less than or equal to 13 seconds. In accordance with the Surveillance Frequency Control Program, the following test shall be performed within 5 minutes of diesel shutdown after the diesel has operated for at least two hours at 2340-2600 kw**: Verifying the diesel generator starts and achieves 3910 volts and 58.8 Hz in s 13 seconds, and subsequently achieves steady state voltage of 3910 and s 4400 volts and frequency of 60 +/- 1.2 Hz. In accordance with the Surveillance Frequency Control Program verifying the diesel generator operates for at least 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s*. During the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of this test, the diesel generators shall be loaded to 2760-2860 Kw**. During the remaining 22 hours2.546296e-4 days 0.00611 hours 3.637566e-5 weeks 8.371e-6 months of this test, the diesel generator shall be loaded to 2500-2600 Kw**. The steady state voltage and frequency shall be maintained at 3910 and s 4580 volts and 60 +/- 1.2 Hz during this test. 4.8.1.1.3 The diesel fuel oil storage and transfer system shall be demonstrated OPERABLE: In accordance with the Surveillance Frequency Control Program by: Verifying the level in each of the above required fuel storage tanks. Verifying that both fuel transfer pumps can be started and transfer fuel from the fuel storage tanks to the day tanks. SALEM -UNIT 3/48-5 Amendment No. 299 ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) In accordance with the Surveillance Frequency Control Program by verifying that a sample of diesel fuel from each of the above required fuel storage tanks is within the acceptable limits specified in Table 1 of ASTM 0975-77 when checked for viscosity, water and sediment.

4.8.1.1.4 Reports -NOT USED Surveillance testing may be conducted in accordance with the manufacturer's recommendations regarding engine prelube, warm-up and loading (unless loading times are specified in the individual Surveillance Requirements). This band is meant as guidance to preclude routine exceedances of the diesel generator manufacturer's design ratings. Loads in excess of this band for special testing or momentary variations due to changing bus loads shall not invalidate the test. SALEM -UNIT 3/48-5a Amendment No. 299 ELECTRICAL POWER SYSTEMS 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS AC. DISTRIBUTION

-OPERATING 3.8.2.1 The following AC. electrical busses shall be OPERABLE and energized from sources of power other than the diesel generators:

4 kvolt Vital Bus # 1 A 4 kvolt Vital Bus # 1 B 4 kvolt Vital Bus # 1 C 460 volt Vital Bus # 1 A and associated control centers 460 volt Vital Bus # 1 B and associated control centers 460 volt Vital Bus # 1 C and associated control centers 230 volt Vital Bus # 1 A and associated control centers 230 volt Vital Bus # 1 B and associated control centers 230 volt Vital Bus # 1 C and associated control centers 115 volt Vital Instrument Bus # 1A and Inverter

115 volt Vital Instrument Bus # 1 B and Inverter

115 volt Vital Instrument Bus # 1 C and Inverter

115 volt Vital Instrument Bus # 1 D and Inverter

APPLICABILITY:

MODES 1, 2, 3 and 4. ACTION: With less than the above complement of AC. busses OPERABLE or energized, restore the inoperable bus to OPERABLE and energized status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 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 and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . With one inverter inoperable, energize the associated AC. Vital Bus within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; restore the inoperable 1A, 1B, or 1C inverter to OPERABLE and energized status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 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 and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ; restore the inoperable 1 D inverter to OPERABLE and energized 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 and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . 4.8.2.1 The specified AC. busses shall be determined OPERABLE and energized from AC. sources other than the diesel generators in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignment and indicated power availability. An inverter may be disconnected from its DC source for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for the purpose of performing an equalizing charge on its associated battery bank provided (1) its vital bus is OPERABLE and energized, and (2) the vital busses associated with the other battery banks are OPERABLE and energized.

SALEM -UNIT 3/4 8-6 Amendment No. 299 ELECTRICAL POWER SYSTEMS AC. DISTRIBUTION

-SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.2.2 As a minimum, two AC. electrical bus trains shall be OPERABLE and energized from sources of power other than a diesel generator but aligned to an OPERABLE diesel generator with each train consisting of: 1 4 kvolt Vital Bus 1 460 volt Vital Bus and associated control centers 1 230 volt Vital Bus and associated control centers 1 115 volt Instrument Bus energized from its respective inverter connected to its respective D.C. bus train. APPLICABILITY:

MODES 5 and 6. During movement of irradiated fuel assemblies.

ACTION: With less than the above complement of AC. busses and inverters OPERABLE and energized, immediately declare the affected required features inoperable, or suspend all operations involving CORE ALTERATIONS, positive reactivity changes, and movement of irradiated fuel assemblies until the minimum required AC. electrical power sources are restored to OPERABLE status. 4.8.2.2 The specified AC. busses and inverters shall be determined OPERABLE and energized from AC. sources other than the diesel generators in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignment and indicated voltage on the busses. SALEM -UNIT 1 3/4 8-7 Amendment No. 299 ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS 4.8.2.3.1 Each D.C. bus train shall be determined OPERABLE and energized in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignment and power availability.

4.8.2.3.2 Each 125-volt battery and above required charger shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by verifying that: 1. The parameters in Table 4.8.2.3-1 meet Category A limits. 2. The overall battery voltage is greater than or equal to 125 volts on float charge. b. In accordance with the Surveillance Frequency Control Program and once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a battery discharge

< 110 V and once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a battery overcharge>

150 V by verifying that the parameters in Table 4.8.2.3-1 meet the Category B limits. c. In accordance with the Surveillance Frequency Control Program by verifying that: 1. There is no visible corrosion at terminals or connectors or the connection resistance is: s; 150 micro ohms for inter-cell connections, s; 350 micro ohms for inter-rack connections, s; 350 micro ohms for inter-tier connections, s; 70 micro ohms for field cable terminal connections, and s; 2500 micro ohms for the total battery connection reSistance which includes all inter-cell connections (including bus bars), all inter-rack connections (including cable resistance) all inter-tier connections (including cable resistance) and all field terminal connections at the battery. The average electrolyte temperature of the representative cells is above 65°F. In accordance with the Surveillance Frequency Control Program by verifying that: The cells, cell plates and battery racks show no visual indication of physical damage or abnormal deterioration. Remove visible terminal corrosion and verify cell-to-cell and terminal connections are coated with anti-corrosion material.

SALEM -UNIT 3/4 8-9 Amendment No. 299 ELECTRICAL POWER 3. The connection resistance is: s 150 micro ohms for inter-cell connections, s 350 micro ohms for inter-rack connections, s 350 micro ohms for inter-tier connections, s 70 micro ohms for field cable terminal connections, and s 2500 micro ohms for the total battery connection resistance which includes all inter-cell connections (including bus bars), all inter-rack connections (including cable resistance), all inter-tier connections (including cable resistance), and all field terminal connections at the battery. e. In accordance with the Surveillance Frequency Control Program by verifying that the battery charger will supply at least 170 amperes at 125 volts for at least 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . f. In accordance with the Surveillance Frequency Control Program, during shutdown, by verifying that the battery capacity is adequate to supply and maintain in OPERABLE status all of the actual or simulated emergency loads for the design duty cycie when the battery is subjected to a battery service test. g. In accordance with the Surveillance Frequency Control Program, during shutdown, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test. Satisfactory completion of this performance discharge test shall also satisfy the requirements of Specification 4.8.2.3.2.f if the performance discharge test is conducted during a shutdown where that test and the battery service test would both be required.

h. At least once per 12 months, during shutdown, if the battery shows signs of degradation OR has reached 85% of the service life with a capacity less than 100% of manufacturers rating, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test. Degradation is indicated when the battery capacity drops more than 10% of rated capacity from its capacity on the previous performance test, or is below 90% of the manufacturer's rating. i. At least once per 24 months, during shutdown, if the battery has reached 85% of the service life with capacity greater than or equal to 100% of manufacturers rating, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test. SALEM -UNIT 1 3/48-9a Amendment No. 299 ELECTRICAL POWER SYSTEMS 125-VOLT D.C. DISTRIBUTION

-SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.2.4 As a minimum, the following D.C. electrical equipment and bus shall be energized and OPERABLE: 2 125-volt D.C. busses, and 2 125-volt batteries, each with at least one full capacity charger, associated with each of the above D.C. busses.

MODES 5 and 6. During movement of irradiated fuel assemblies.

ACTION: With less than the above complement of D.C. equipment and busses OPERABLE, immediately declare the affected required features inoperable, or suspend all operations involving CORE ALTERATIONS, positive reactivity changes, and movement of irradiated fuel assemblies until the minimum required 125 Volt D.C. electrical power sources are restored to OPERABLE status. SURVEILLANCE REQUIREMENTS 4.8.2.4.1 The above required 125-volt D.C. busses shall be determined OPERABLE and energized in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignment and indicated power availability.

4.8.2.4.2 The above required 125-volt batteries and chargers shall be demonstrated OPERABLE per Surveillance Requirement 4.8.2.3.2.

SALEM -UNIT 3/48-10 Amendment No. 299 ELECTRICAL POWER 4.8.2.5.1 Each D.C. bus train shall be determined OPERABLE and energized in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignment and power availability.

4.8.2.5.2 Each 28-volt battery and above required charger shall be demonstrated OPERABLE: In accordance with the Surveillance Frequency Control Program by verifying that: The parameters in Table 4.8.2.5-1 meet Category A limits. The overall battery voltage is greater than or equal to 27 volts on float charge. In accordance with the Surveillance Frequency Control Program and once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a battery discharge

< 25.7 V and once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a battery overcharge>

35 V by verifying that the parameters in Table 4.8.2.5-1 meet the Category B limits. In accordance with the Surveillance Frequency Control Program by verifying that: There is no visible corrosion at terminals or connectors or the connection resistance is: s; 50 micro ohms for inter-cell connections, s; 200 micro ohms for inter-tier connections, s; 70 micro ohms for field cable terminal connections, and s; 500 micro ohms for the total battery connection resistance which includes all inter-cell connections (including bus bars), all inter-tier connections (including cable resistance) and all field terminal connections at the battery. The average electrolyte temperature of the representative cells is ;;:= 65°F. In accordance with the Surveillance Frequency Control Program by verifying that: The cells, cell plates and battery racks show no visual indication of physical damage or abnormal deterioration. Remove visible terminal corrosion and verify cell-to-cell and terminal connections are coated with anti-corrosion material.

SALEM -UNIT 3/48-12 Amendment No. 299 ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) The connection resistance is: s 50 micro ohms for inter-cell connections, s 200 micro ohms for inter-tier connections, s 70 micro ohms for field cable terminal connections, and s 500 micro ohms for the total battery connection resistance which includes all inter-cell connections (including bus bars), all inter-tier connections (including cable resistance) and all field terminal connections at the battery. In accordance with the Surveillance Frequency Control Program by verifying that the battery charger will supply;:::

150 amperes at ;::: 28 volts for;::: 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . In accordance with the Surveillance Frequency Control Program, during shutdown, by verifying that the battery capacity is adequate to supply and maintain in OPERABLE status all of the actual or simulated emergency loads for the design duty cycle when the battery is subjected to a battery service test. In accordance with the Surveillance Frequency Control Program, during shutdown, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test. Satisfactory completion of this performance discharge test shall also satisfy the requirements of Specification 4.8.2.5.2.f if the performance discharge test is conducted during a shutdown where that test and the battery service test would both be required. At least once per 12 months, during shutdown, if the battery shows signs of degradation OR has reached 85% of the service life with a capacity less than 100% of manufacturers rating, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test. Degradation is indicated when the battery capacity drops more than 10% of rated capacity from its capacity on the previous performance test, or is below 90% of the manufacturer's rating. At least once per 24 months, during shutdown, if the battery has reached 85% of the service life with capacity greater than or equal to 100% of manufacturers rating, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test. SALEM -UNIT 3/48-12a Amendment No. 299 ELECTRICAL POWER SYSTEIVIS 28*VOL T D.C. DISTRIBUTION*

SHUTDOWN 3.8.2.6 As a minimum, the following D.C. electrical equipment and bus shall be energized and OPERABLE: 1 28-volt D.C. bus, and 1 28-volt battery and at least one full capacity charger associated with the above D.C. bus.

MODES 5 and 6. During movement of irradiated fuel assemblies.

ACTION: With less than the above complement of D.C. equipment and busses OPERABLE, immediately declare the affected required features inoperable, or suspend all operations involving CORE ALTERATIONS, positive reactivity changes, and movement or irradiated fuel assemblies until the minimum required 28 Volt D.C. electrical power sources are restored to OPERABLE status. 4.8.2.6.1 The above required 28-volt D.C. bus shall be determined OPERABLE and energized in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignment and voltage on the bus. 4.8.2.6.2 The above required 28-volt batteries and charger shall be demonstrated OPERABLE per Surveillance Requirement 4.8.2.5.2.

SALEM -UNIT 3/48-13 Amendment No. 299 ELECTRICAL POWER SYSTEMS 3/48.3 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES LIMITING CONDITION FOR OPERATION 3.8.3.1 All containment penetration conductor overcurrent protective devices required provide thermal protection of penetrations shall be APPLICABILITY:

MODES 1, 2, 3 and With one or more of the required containment penetration conductor overcurrent device(s)

a. Restore the protective device(s) to OPERABLE status or de-energize the circuit(s) by tripping either the primary or backup protective device, or racking out or removing the primary or backup device within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , declare the affected system or component inoperable, and verify the primary or backup protective device to be tripped, or the primary or backup device racked out or removed at least once per 7 days thereafter; or b. Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . SURVEILLANCE REQUIREMENTS 4.8.3.1 All required containment penetration conductor overcurrent protective devices shall be demonstrated OPERABLE: In accordance with the Surveillance Frequency Control Program: For 4.16 KV reactor coolant pump circuits, by performance of: A CHANNEL CALIBRATION of the associated protective relays, and An integrated system functional test which includes simulated automatic actuation of the system and verifying that each relay and associated circuit breakers and control circuits function as designed.

SALEM -UNIT 3/48-14 Amendment No. 299 ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) By verifying the OPERABILITY of the required molded case and lower voltage circuit breakers, by selecting and functionally testing a representative sample of at least 10% of all the circuit breakers of that type. Circuit breakers selected for functional testing shall be selected on a rotating basis. The functional test shall consist of injecting a current input at the specified setpoint to each selected circuit breaker and verifying that each circuit breaker functions as designed.

Circuit breakers found inoperable during functional testing shall be restored to OPERABLE status prior to resuming operation.

For each circuit breaker found inoperable during the functional tests, an additional representative sample of at least 10% of all the circuit breakers of the inoperable type shall also be functionally tested until no more failures are found or all circuit breakers of that type have been functionally tested. In accordance with the Surveillance Frequency Control Program by subjecting each circuit breaker to an inspection and preventive maintenance in accordance with procedures prepared in conjunction with its manufacturer's recommendations.

SALEM -UNIT 3/48-15 Amendment No. 299 3/4.9 REFUELING OPERATIONS BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.1 The boron concentration of the Reactor Coolant System, the fuel storage pool, the refueling canal, and the refueling cavity shall be maintained within the limit specified in the CORE OPERATING LIMITS REPORT (COLR).

MODE 6 (Only applicable to the refueling canal, the fuel storage pool and refueling cavity when connected to the Reactor Coolant System) ACTION: With the requirements of the above specification not satisfied, immediately

a. Suspend CORE ALTERATIONS and b. Suspend positive reactivity additions and c. Initiate action to restore boron concentration to within limit specified in the COLR. d. The provisions of Specification 3.0.3 are not applicable.

4.9.1. Verify the boron concentration is within the limit of the COLR in accordance with the Surveillance Frequency Control Program. SALEM -UNIT 3/49-1 Amendment No. 299 REFUELING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2 As a minimum, two source range neutron flux monitors shall be operating, each with continuous visual indication in the control room and one with audible indication in the containment and control room. APPLICABILITY:

MODE 6. ACTION: With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.2 Each source range neutron flux monitor shall be demonstrated OPERABLE by performance of: a. A CHANNEL FUNCTIONAL TEST in accordance with the Surveillance Frequency Control Program, and b. A CHANNEL CHECK in accordance with the Surveillance Frequency Control Program during CORE ALTERATIONS.

SALEM -UNIT 1 3/4 9-2 Amendment No. 299 REFUELING CONTAINMENT BUILDING PENETRATIONS LIMITING CONDITION FOR OPERATION 3.9.4 The containment building penetrations shall be in the following status: a. The equipment hatch inside door is capable of being closed and held in place by a minimum of four bolts, or an equivalent closure device installed and capable of being closed, b. A minimum of one door in each airlock is capable of being closed c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either: 1. closed by a manual or automatic isolation valve, blind flange, or equivalent, or 2. capable of being closed by the Containment Purge and Pressure-Vacuum Relief Isolation System. Note: Penetration flow path(s) providing direct access from the containment atmosphere to the outside atmosphere may be unisolated under administrative controls.

APPLICABILITY:

During movement of irradiated fuel within the containment.

ACTION: With the requirements of the above specification not satisfied, immediately suspend all operations involving movement of irradiated fuel in the containment building.

The provisions of Specification 3.0.3 are not applicable.

4.9.4.1 Each of the above required containment building penetrations shall be determined to be either in its required condition or capable of being closed by a manual or automatic containment isolation valve in accordance with the Surveillance Frequency Control Program. 4.9.4.2 Once per refueling prior to the start of movement of irradiated fuel assemblies within the containment building, verify the capability to close, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , the equipment hatch inside door or an equivalent closure device. Applicable only when the equipment hatch is open during movement of irradiated fuel in the containment building.

4.9.4.3 Verify, in accordance with the Surveillance Frequency Control Program, each required containment purge isolation valve actuates to the isolation position on a manual actuation signal. SALEM -UNIT 1 3/4 9-4 Amendment No. 299 REFUELING 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION ALL WATER LEVELS LIMITING CONDITION FOR OPERATION 3.9.8.1 At least one residual heat removal loop shall be in APPLICABILITY:

MODE a. With less than one residual heat removal loop in operation, except as provided in b. below. suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System, Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . b, The residual heat removal loop may be removed from operation for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period during the performance of CORE ALTERATIONS in the vicinity of the reactor pressure vessel hot legs, c, The provisions of Specification 3,0,3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.8.1 In accordance with the Surveillance Frequency Control Program one RHR loop shall be verified in operation and circulating coolant at a flow rate of: a. greater than or equal to 1000 gpm, and b. sufficient to maintain the RCS temperature at less than or equal to 140°F, SALEM -UNIT 1 3/4 9-8 Amendment No. 299 REFUELING WATER LEVEL -REACTOR VESSEL LIMITING CONDITION FOR OPERATION 3.9.10 At least 23 feet of water shall be maintained over the top of the reactor pressure vessel flange.

During movement of fuel assemblies or control rods within the reactor pressure vessel while in MODE 6. ACTION: With the requirements of the above specification not satisfied, suspend all operations involving movement of fuel assemblies or control rods within the pressure vessel. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.10 The water level shall be determined to be at least its minimum required depth within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to the start of and in accordance with the Surveillance Frequency Control Program thereafter during movements of fuel assemblies or control rods. SALEM -UNIT 3/49-10 Amendment No. 299 REFUELING STORAGE POOL WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.11 At least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks. APPLICABILITY:

Whenever irradiated fuel assemblies are in the storage pool. ACTION: With the requirements of the specification not satisfied, suspend all movement of fuel assemblies and crane operations with loads in the fuel storage areas and restore water level to within its limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.11 The water level in the storage pool shall be determined to be at least its minimum required depth in accordance with the Surveillance Frequency Control Program when irradiated fuel assemblies are in the fuel storage pool. SALEM -UNIT 1 3/49-11 Amendment No. 299 REFUELING OPERATIONS FUEL HANDLING AREA VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.9.12 The Fuel Handling Area Ventilation System shall be OPERABLE with: Two exhaust fans and one supply fan OPERABLE and operating, and b. Capable of maintaining slightly negative pressure in the Fuel Handling Building.

APPLICABILITY:

During movement of irradiated fuel within the Fuel Handling Building ACTION: a. With no Fuel Handling Area Ventilation System OPERABLE, suspend all operations involving movement of fuel within the storage pool until the Fuel Handling Area Ventilation System is restored to OPERABLE status. b. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.12 The above required ventilation system shall be demonstrated OPERABLE: In accordance with the Surveillance Frequency Control Program by verifying that, the Fuel Handling Building is maintained at a slightly negative pressure with respect to atmospheric pressure. In accordance with the Surveillance Frequency Control Program by verifying both exhaust fans and one supply fan start and operate for at least 15 minutes, if not operating already. In accordance with the Surveillance Frequency Control Program by verifying a system flowrate of 19,490 cfm +/- 10% during system operation.

SALEM -UNIT 3/49-12 Amendment No. 299 3/4.10 SPECIAL TEST EXCEPTIONS SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of control rod worth and shutdown margin provided the reactivity equivalent to at least the highest estimated control rod worth is available for trip insertion from OPERABLE control rod(s). APPLICABILITY:

MODE 2. ACTION: a. With any full length control rod not fully inserted and with less than the above reactivity equivalent available for trip insertion, immediately initiate and continue boration at 33 gpm of a solution containing 6,560 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.

b. With all full length control rods inserted and the reactor subcritical by less than the above reactivity equivalent, immediately initiate and continue boration at 33 gpm of a solution containing 6,560 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.

4.10.1.1 The position of each full length and part length rod either partially or FULLY WITHDRAWN shall be determined in accordance with the Surveillance Frequency Control Program. 4.10.1.2 Each full length rod not fully inserted shall be demonstrated capable of full insertion when tripped from at least the 50% withdrawn position within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to reducing the SHUTDOWN MARGIN to less than the limits of Specification 3.1.1.1. SALEM -UNIT 1 3/410-1 Amendment No. 299 SPECIAL TEST EXCEPTIONS GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION 3.10.2 The group height, insertion and power distribution limits of Specifications 3.1.3.1, 3.1.3.4, 3.1.3.5, 3.2.1, and 3.2.4 may be suspended during the performance of PHYSICS TESTS provided: The THERMAL POWER is maintained

5 85% of RATED THERMAL POWER, and The limits of Specifications 3.2.2 and 3.2.3 are maintained and determined at the frequencies specified in Specification 4.10.2.2 below. APPLICABILITY:

MODE 1 With any of the limits of Specifications 3.2.2 or 3.2.3 being exceeded while the requirements Specifications 3.1.3.1, 3.1.3.4, 3.1.3.5, 3.2.1 and 3.2.4 are suspended, Reduce THERMAL POWER sufficient to satisfy the ACTION requirements of Specifications 3.2.2 and 3.2.3, or Be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . 4.10.2.1 The THERMAL POWER shall be determined to be :5 85% of RATED THERMAL POWER in accordance with the Surveillance Frequency Control Program during PHYSICS TESTS. 4.10.2.2 The below listed surveillance requirements shall be performed in accordance with the Surveillance Frequency Control Program during PHYSICS TESTS: Surveillances 4.2.2.2 and 4.2.2.3. Surveillances 4.2.3.1 and 4.2.3.2. SALEM -UNIT 3/4 10-2 Amendment No. 299 SPECIAL TEST EXCEPTIONS PHYSICS TESTS LIMITING CONDITION FOR OPERATION 3.10.3 The limitations of Specifications 3.1.1.4, 3.1.3.1, 3.1.3.4, and 3.1.3.5 may be suspended during the performance of PHYSICS TESTS provided: The THERMAL POWER does not exceed 5% of RATED THERMAL POWER, and The reactor trip setpoints on the OPERABLE Intermediate and Power Range Nuclear Channels are set at:s; 25% of RATED THERMAL POWER. APPLICABILITY:

MODE 2. With the THERMAL POWER> 5% of RATED THERMAL POWER, immediately open reactor trip SURVEILLANCE REQUIREMENTS 4.10.3.1 The THERMAL POWER shall be determined to be:s; 5% of RATED THERMAL POWER in accordance with the Surveillance Frequency Control Program during PHYSICS TESTS. 4.10.3.2 Each Intermediate and Power Range Channel shall be subjected to a CHANNEL FUNCTIONAL TEST prior to initiating PHYSICS TESTS. SALEM -UNIT 3/410-3 Amendment No. 299 SPECIAL TEST EXCEPTION NO FLOW TESTS LIMITING CONDITION FOR OPERATION 3.10.4 The limitations of Specification 3.4.1.1 may be suspended during the performance of startup and PHYSICS TESTS, provided:

a. The THERMAL POWER does not exceed the P-7 Interlock Setpoint, and b. The Reactor Trip Setpoints on the OPERABLE Intermediate and Power Range Channels are set:s; 25% of RATED THERMAL POWER APPLICABILITY:

During operation below the P-7 Interlock Setpoint.

ACTION: With the THERMAL POWER greater than the P-7 Interlock Setpoint, immediately open the reactor trip breakers.

SURVEILLANCE REQUIREMENTS 4.10.4.1 The THERMAL POWER shall be determined to be less than P-7 Interlock Setpoint in accordance with the Surveillance Frequency Control Program during startup and PHYSICS TESTS. 4.10.4.2 Each Intermediate, Power Range Channel and P-7 Interlock shall be subjected to a CHANNEL FUNCTIONAL TEST prior to initiating startup or PHYSICS TESTS. SALEM -UNIT 1 3/4 10-4 Amendment No. 299 RADIOACTIVE EFFLUENTS LIQUID HOLDUP TANKS* JJMITING CONDITION FOR OPERATION 3.11.1.4 The quantity of radioactive material contained in each outdoor temporary tank shall be limited to less than or equal to 10 curies, excluding tritium and dissolved or entrained noble gases. APPLICABILITY:

At all times. ACTION: With the quantity of radioactive material in any of the above listed tanks exceeding the above limit. immediately suspend all additions of radioactive material to the tank and within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limit. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE 4.11.1.4 The quantity of radioactive material contained in each outdoor temporary tank shall be determined to be within the above limit by analyzing a representative sample of the tank's contents in accordance with the Surveillance Frequency Control Program when radioactive materials are being added to the tank. Tanks included in this Specification are those outdoor temporary tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system. SALEM -UNIT 3/4 11-7 Amendment No. 299 RADIOACTIVE EFFLUENTS EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR OPERAIIQN 3.11.2.5 The concentration of oxygen in the waste gas holdup system shall be limited to less than or equal to 2% by volume. APPLICABILITY:

At all times.

ACTION: With the concentration of oxygen in the waste gas holdup system greater than 2% by volume but less than or equal to 4% by volume, reduce the oxygen concentration to the above limits within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . With the concentration of oxygen in the waste gas holdup system greater than 4% by volume immediately suspend all additions of waste gases to the system and reduce the concentration of oxygen to less than or equal to 2% by volume without delay. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.5 The concentration of oxygen in the waste gas holdup system shall be determined to be within the above limits by continuously**

monitoring the waste gases in the waste gas holdup system with the oxygen monitor. If hydrogen is not measured, the concentration of hydrogen shall be assumed to exceed 4% by volume. Not applicable to portions of the Waste Gas System removed from service for maintenance provided that, the portions removed for maintenance are isolated, and purged of hydrogen to less than 4% by volume. If the oxygen monitoring instrumentation is inoperable, operation of the waste gas holdup system may continue provided grab samples are collected in accordance with the Surveillance Frequency Control Program and analyzed within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . SALEM -UNIT 3/4 11-15 Amendment No. 299 ADMINISTRATIVE 6.8.4.1 Surveillance Frequency Control Program This program provides controls for Surveillance Frequencies.

The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met. The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1. The provisions of Surveillance Requirements 4.0.2 and 4.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program. SALEM -UNIT 6-19f Amendment No. 299 UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 PSEG NUCLEAR, EXELON GENERATION COMPANY, DOCKET NO. SALEM NUCLEAR GENERATING STATION, UNIT AMENDMENT TO FACILITY OPERATING Amendment No. 282 License No. DPR-75 The Nuclear Regulatory Commission (the Commission) has found that: The application for amendment filed by PSEG Nuclear LLC, acting on behalf of itself and Exelon Generation Company, LLC (the licensees) dated March 23, 2010, as supplemented by letters dated November 19, 2010, January 31, 2011, and February 23, 2011, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in Title 10 of the Code of Federal Regulations (10 CFR), Chapter I; The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-75 is hereby amended to read as follows:

-2 Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 282, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical SpeCifications. This license amendment is effective as of its date of issuance and shall be implemented within 120 days. ISSION Harold K. Chernoff, Chief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Facility Operating License and the Technical Specifications Date of Issuance:

March 21, 2011 FOR THE NUCLEAR REGULATORY CO ATTACHMENT TO LICENSE AMENDMENT NO. 282 FACILITY OPERATING LICENSE NO. DPR-75 DOCKET NO. 50-311 Replace the following page of Facility Operating License No. DPR-75 with the attached revised page as indicated.

The revise page is identified by amendment number and contains a marginal line indicating the area of change. Remove Page 4 Replace the following pages of the Appendix A, Technical Specifications, with the attached revised pages as indicated.

The revised pages are identified by amendment number and contain marginal lines indicating the areas of change. Remove Insert Remove Insert Remove Insert Remove Insert 3/4 1-1 3/4 1-1 3/43-43 3/43-43 3/4 6-7 3/46-7 3/4 9-2 3/4 9-2 3/4 1-2 3/4 1-2 3/43-45 3/43-45 3/46-10 3/46-10 3/4 9-4 3/4 9-4 3/4 1-3 3/4 1-3 3/43-50 3/43-50 3/46-11 3/46-11 3/49-8 3/4 9-8 3/4 1-7 3/4 1-7 3/43-52 3/43-52 3/46-13 3/46-13 3/4 9-11 3/49-11 3/4 1-8 3/4 1-8 3/43-52a 3/43-52a 3/46-15 3/46-15 3/49-12 3/49-12 3/41-11 3/4 1-11 3/43-53 3/43-53 3/47-5 3/47-5 3/49-13 3/49-13 3/4 1-12 3/4 1-12 3/43-55 3/43-55 3/47-6 3/47-6 3/4 10-1 3/4 10-1 3/4 1-14 3/4 1-14 3/43-66 3/43-66 3/47-7 3/47-7 3/4 10-2 3/4 10-2 3/41-16a 3/4 1-16a 3/44-1 3/44-1 3/47-9 3/47-9 3/4 10-4 3/4 10-4 3/41-18 3/4 1-18 3/44-2a 3/44-2a 3/47-11 3/47-11 3/4 10-5 3/4 10-5 3/41-19 3/41-19 3/4 4-4 3/4 4-4 3/47-12 3/47-12 3/4 11-7 3/411-7 3/4 1-20 3/4 1-20 3/44-4a 3/44-4a 3/47-13 3/47-13 3/411-15 3/411-15 3/4 2-2 3/42-2 3/44-7 3/44-7 3/47-14 3/47-14 6-19g 3/42-3 3/4 2-3 3/44-8a 3/44-8a 3/47-17 3/47-17 3/42-6 3/42-6 3/44-16 3/44-16 3/47-19 3/47-19 3/42-7 3/42-7 3/44-17 3/44-17 3/47-21 3/47-21 3/42-10 3/42-10 3/44-18 3/44-18 3/47-29 3/47-29 3/42-15 3/42-15 3/44-25 3/44-25 3/47-30 3/47-30 3/42-16 3/42-16 3/44-27 3/44-27 3/4 8-3 3/4 8-3 3/4 3-1 3/4 3-1 3/44-30 3/44-30 3/4 8-4 3/4 8-4 3/43-11 3/4 3-11 3/44-32 3/44-32 3/4 8-5 3/4 8-5 3/43-12 3/43-12 3/44-34 3/44-34 3/4 8-6 3/4 8-6 3/43-13 3/43-13 3/4 5-1 3/45-1 3/48-8 3/4 8-8 3/43-13a 3/4 5-2 3/45-2 3/48-9 3/4 8-9 3/43-14 3/43-14 3/4 5-4 3/45-4 3/4 8-11 3/4 8-11 3/43-33 3/43-33 3/4 5-5 3/45-5 3/48-11a 3/48-11a 3/43-34 3/43-34 3/4 5-6 3/45-6 3/48-12 3/48-12 3/43-35 3/43-35 3/4 5-8 3/4 5-8 3/48-14 3/48-14 3/43-36 3/43-36 3/45-8a 3/45-8a 3/48-14a 3/48-14a 3/43-37 3/43-37 3/45-9 3/45-9 3/48-15 3/48-15 3/43-38 3/43-38 3/46-1 3/46-1 3/48-16 3/48-16 3/43-41 3/43-41 3/4 6-5 3/46-5 3/48-17 3/48-17 3/43-42 3/46-6 3/4 6-6 3/49-1 3/49-1 I -(2) Technical Specifications and Environmental Plan The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 282, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

Amendment No. 282 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN -T > 200°F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.3% delta APPLICABILITY:

MODES 1,2*,3, and With the SHUTDOWN MARGIN less than 1.3% delta k/k, immediately initiate and boration at 33 gpm of a solution containing 6,560 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.3% delta k/k: a. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after detection of an inoperable control rod(s) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the rod(s) is inoperable.

If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the immovable or untrippable control rod(s). b. When in MODE 1 or MODE 2 with Kelt greater than or equal to 1.0, in accordance with the Surveillance Frequency Control Program by verifying that control banks are within the limits in the COLR per Specification 3.1.3.5. c. When in MODE 2 with Kelt less than 1.0, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to achieving reactor criticality by verifying that the predicted critical control rod position is within the limits in the COLR per Specification 3.1.3.5. *See Special Test Exception 3.10.1 SALEM -UNIT 2 3/4 1-1 Amendment No. 282 REACTIVITY CONTROL Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of e below, with the control banks at the maximum insertion limit in the COLR per Specification 3.1.3.5. When in MODES 3 or 4, in accordance with the Surveillance Frequency Control Program by consideration of the following factors: Reactor coolant system boron concentration, Control rod position, Reactor coolant system average temperature, Fuel burnup based on gross thermal energy generation, Xenon concentration, and Samarium concentration.

4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within +/- 1 % delta klk in accordance with the Surveillance Frequency Control Program. This comparison shall consider at least those factors stated in Specification 4.1.1.1.1.e, above. The predicted reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 Effective Full Power Days after each fuel loading. SALEM* UNIT 3/4 1*2 Amendment No. 282 REACTIVITY CONTROL SHUTDOWN MARGIN -T£YlL 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to 1.0% delta APPLICABILITY:

MODE With the SHUTDOWN MARGIN less than 1.Q% delta k/k, immediately initiate and boration at 33 gpm of a solution containing 6,560 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

4.1.1.2 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.0% delta k/k: Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after detection of an inoperable control rod(s) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the rod(s) is inoperable.

If the inoperable control rod is immovable or untrippable, the SHUTDOWN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the immovable or untrippable control rod(s). In accordance with the Surveillance Frequency Control Program by consideration of the following factors: Reactor coolant system boron concentration, Control rod position, Reactor coolant system average temperature, Fuel burnup based on gross thermal energy generation, Xenon concentration, and Samarium concentration.

SALEM -UNIT 3/4 1-3 Amendment No. 282 REACTIVITY CONTROL SYSTEMS 3/4.1.2 BORA TION SYSTEMS FLOW PATHS -SHUTDOWN LIMITING CQNDITIQNFOR QPERATIQN As a minimum, one of the following boron injection flow paths shall be OPERABLE: A flow path from the boric acid tanks via a boric acid transfer pump and a charging pump to the Reactor Coolant System if the boric acid storage system is OPERABLE, per Specification 3.1.2.6a while in MODE 4, or per Specification 3.1.2.5a while in MODE 5 or 6, or A flow path from the refueling water storage tank via a charging pump to the Reactor Coolant System if the refueling water storage tank is OPERABLE per Specification 3.1.2.6b while in MODE 4, or per Specification 3.1.2.5b while in MODE 5 or 6. APPLICABILITY:

MODES 4, 5 and 6. ACTION: With none of the above flow paths OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until at least one injection path is restored to OPERABLE status. SURj,[EI LLANCE REQUIREMENTS 4.1.2.1 At least one of the above required flow paths shall be demonstrated OPERABLE: . When the boric acid tank is a required water source, by verifying in accordance with the Surveillance Frequency Control Program that: The flow path from the boric acid tank to the boric acid transfer pump, the boric acid transfer pump, and the recirculation path from the boric acid transfer pump to the boric acid tank is 63°F, and The flow path between the boric acid transfer pump recirculation line to the charging pump suction line is 50"F, 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.

SALEM -UNIT 3/41-7 Amendment No. 282 REACTIVITY CONTROL SYSTEMS FLOW PATHS -OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.2 At least two of the following three boron injection flow paths shall be OPERABLE:

a. A flow path from the boric acid tanks via a boric acid transfer pump and a charging pump to the Reactor Coolant System. b. Two flow paths from the refueling water storage tank via charging pumps to the Reactor Coolant System. APPLICABILITY:

MODES 1, 2 and 3. ACTION: With only one of the above required boron injection flow paths to the Reactor Coolant System OPERABLE, restore at least two boron injection flow paths to the Reactor Coolant 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 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 flow paths 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 . 4.1.2.2 Each of the above required flow paths shall be demonstrated OPERABLE: By verifying in accordance with the Surveillance Frequency Control Program that: The flow path from the boric acid tank to the boric acid transfer pump and from the recirculation line back to the boric acid tank is 63°F, and The flow path between the boric acid tank recirculation line to the charging pump suction line is 50°F, 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. In accordance with the Surveillance Frequency Control Program during shutdown by verifying that each automatic valve in the flow path actuates to its correct position on a safety injection test signal. In accordance with the Surveillance Frequency Control Program by verifying that the flow path required by Specification 3.1.2.2.a delivers at least 33 gpm to the Reactor Coolant System. SALEM -UNIT 3/4 1-8 Amendment No. 282 REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES -SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.5 As a minimum, one of the following borated water sources shall be OPERABLE: A boric acid storage system with: A minimum contained volume of 2,600 gallons, Between 6,560 and 6,990 ppm of boron, and A minimum solution temperature of 63°F. The refueling water storage tank with: A minimum contained volume of 37,000 gallons, A minimum boron concentration of 2,300 ppm, and 3. A minimum solution temperature of 35°F. APPLICABILITY:

MODES 5 and 6. ACTION: With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until at least one borated water source is restored to OPERABLE status. SURVEILLANCE REQUIREMENTS 4.1.2.5 The above required borated water source shall be demonstrated OPERABLE: For the boric acid storage system, when it is the source of borated water in accordance with the Surveillance Frequency Control Program by: Verifying the boron concentration of the water, Verifying the water level of the tank, and Verifying the boric acid storage tank solution temperature when it is the source of borated water. For the refueling water storage tank by: Verifying the boron concentration in accordance with the Surveillance Frequency Control Program, Verifying the borated water volume in accordance with the Surveillance Frequency Control Program, and Verifying the solution temperature in accordance with the Surveillance Frequency Control Program, when it is the source of borated water and the outside air temperature is less than 35°F. SALEM -UNIT 3/4 1-11 Amendment No. 282 REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES -OPERATING LIMITING CQNPITION FOR OPERATION 3.1.2.6 As a minimum, the following borated water source(s) shall be OPERABLE as required by Specifications 3.1.2.1 and 3.1.2.2: A boric acid storage system with: A contained volume of borated water in accordance with figure 3.1-2, A Boron concentration in accordance with Figure 3.1-2, and A minimum solution temperature of 63°F. The refueling water storage tank with: A contained volume of between 364,500 and 400,000 gallons of water, A boron concentration of between 2,300 and 2,500 ppm, and 3. A minimum solution temperature of 35°F. APPLICABILITY:

MODES 1, 2, 3 and 4. ACTION: With the boric acid storage system inoperable and being used as one of the above required borated water sources, restore the storage 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 and borated to a SHUTDOWN MARGIN equivalent to at least 1% delta klk at 200°F; restore the boric acid storage system 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 . With the refueling water storage tank inoperable, restore the tank to OPERABLE status within one hour 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 . 4.1.2.6 Each borated water source shall be demonstrated OPERABLE: For the boric acid storage system, when it is the source of borated water in accordance with the Surveillance Frequency Control Program by: Verifying the boron concentration in each water source. Verifying the water level of each water source, and Verifying the boric acid storage system solution temperature. For the refueling water storage tank by: Verifying the boron concentration in accordance with the Surveillance Frequency Control Program, Verifying the borated water volume in accordance with the Surveillance Frequency Control Program, and Verifying the solution temperature in accordance with the Surveillance Frequency Control Program when the outside air temperature is less than 35°F. SALEM -UNIT 3/41-12 Amendment No. 282 REACTIVITY CONTROL LIMITING CONDITION FOR OPERATION A reevaluation of each accident analysis of Table 3.1-1 is performed within 5 days; this reevaluation shall confirm that the previously analyzed results of these accidents remain valid for the duration of operation under these conditions. The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . A core power distribution measurement is obtained and Fo(Z) and FN t.H are verified to be within their limits within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . The THERMAL POWER level is reduced to less than or equal to 75% of RATED THERMAL POWER within one hour and within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the high neutron flux trip setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER. THERMAL POWER shall be maintained less than or equal to 75% of RATED THERMAL POWER until compliance with ACTIONS 3.1.3.1.c.3.a and 3.1.3.1.c.3.c above are demonstrated.

SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The position of each full length rod shall be determined to be within the limits established in the limiting condition for operation in accordance with the Surveillance Frequency Control Program (allowing for one hour thermal soak after rod motion) except during time intervals when the Rod Position Deviation Monitor is inoperable, then verify the group positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . 4.1.3.1.2 Each full length rod not fully inserted in the core shall be determined to be OPERABLE by movement of at least 10 steps in anyone direction in accordance with the Surveillance Frequency Control Program. SALEM -UNIT 3/4 1-14 Amendment No. 282 REACTIVITY CONTROL LIMITING CONDITION FOR OPERATION With a maximum of one group demand position indicator per bank inoperable either: Verify that all analog rod position indicators for the affected bank are OPERABLE and that the most withdrawn rod and the least withdrawn rod of the bank are within a maximum of 18 steps when reactor power is s; 85% RATED THERMAL POWER or if reactor power is> 85% RATED THERMAL POWER, 12 steps of each other at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , or Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . SURVEILLANCE REQUIREMENTS 4,1.3.2.1.1 Each analog rod position indicator shall be determined to be OPERABLE by verifying that the demand position indication system and the rod position indication system agree within 18 steps when reactor power is s; 85% RATED THERMAL POWER or if reactor power is > 85% RATED THERMAL POWER, 12 steps (allowing for one hour thermal soak after rod motion) in accordance with the Surveillance Frequency Control Program except during time intervals when the Rod Position Deviation Monitor is inoperable, then compare the demand position indication system and the rod position indication system at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . 4.1.3.2.1.2 Each of the above required rod position indicator(s) shall be determined to be OPERABLE by performance of a CHANNEL calibration in accordance with the Surveillance Frequency Control Program. SALEM -UNIT 3/41-16a Amendment No. 282 REACTIVITY CONTROL ROD DROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.3 The individual full length (shutdown and control) rod drop time from 230 steps withdrawn shall be less than or equal to 2.7 seconds from beginning of decay of stationary gripper coil voltage to dashpot entry with: Tav9 greater than or equal to 541°F, and b. All reactor coolant pumps operating.

APPLICABILITY:

MODES 1 & 2. ACTION: With the drop time of any full length rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2. With the rod drop times within limits but determined with 3 reactor coolant pumps operating, operation may proceed provided THERMAL POWER is restricted to less than or equal to 76% of RATED THERMAL POWER. SURVEILLANCE REQUIREMENTS 4.1.3.3 The rod drop time of full length rods shall be demonstrated through measurement prior to reactor criticality: For all rods following each removal of the reactor vessel head, For specifically affected individual rods following any maintenance on or modification to the control rod drive system which could affect the drop time of those specific rods, and In accordance with the Surveillance Frequency Control Program. SALEM -UNIT 3/41-18 Amendment No. 282 REACTIVITY CONTROL SYSTEMS SHUTDOWN ROD INSERTION LIMIT LIMITING CONDITION FOR OPERATION 3.1.3.4 All shutdown rods shall be FULLY APPLICABILITY:

MODES 1*, and With a maximum of one shutdown rod not FULLY WITHDRAWN, except for surveillance pursuant to Specification 4.1.3.1.2, within one hour a. FULLY WITHDRAW the rod, or, b. Declare the rod to be inoperable and apply Specification 3.1.3.1. SURVEILLANCE REQUIREMENTS 4.1.3.4 Each shutdown rod shall be determined to be FULLY WITHDRAWN by use of the group demand counters, and verified by the analog rod position indicators**:

a. Within 15 minutes prior to withdrawal of any rods in control banks A, B, C, and 0 during an approach to reactor critically, and b. In accordance with the Surveillance Frequency Control Program thereafter.

See Special Test Exceptions 3.10.2 and 3.10.3. ** For power levels below 50% one hour thermal "soak time" is permitted.

During this soak time, the absolute value of rod motion is limited to six steps. @ Surveillance 4.1.3.4.a is applicable prior to withdrawing any control banks in preparation for startup (Mode 2). # With Keff greater than or equal to 1.0. Note: This page effective prior to startup from fifth refueling outage scheduled to begin March 1990. Letter dated Jan. 11, 1990. SALEM -UNIT 2 3/4 1-19 Amendment No. 282 REACTIVITY CONTROL CONTROL ROD INSERTION 3.1.3.5 The control banks shall be limited in physical insertion as specified in the OPERATING LIMITS REPORT APPLICABILITY:

MODES 1*, and With the control banks inserted beyond the above insertion limits, except for surveillance pursuant to Specification 4.1.3.1.2, Restore the control banks to within the limits within two hours, or Reduce THERMAL POWER within two hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the bank position using the insertion limits specified in the CLOR, or Be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . SURVEILLANCE REQUIREMENTS 4.1.3.5 The position of each control bank shall be determined to be within the insertion limits in accordance with the Surveillance Frequency Control Program by use of the group demand counters and verified by the analog rod position indicators**

except during time intervals when the Rod Insertion Limit Monitor is inoperable, then verify the individual rod positions at least once per 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s**.

See Special Test Exceptions 3.10.2 and 3.10.3 ** For power levels below 50% one hour thermal "soak time" is permitted.

During this soak time, the absolute value of rod motion is limited to six steps. # With Keff greater than or equal to 1.0 SALEM -UNIT 3/4 1-20 Amendment No. 282 POWER DISTRIBUTION LIMITING CONDITION FOR OPERATION

b. THERMAL POWER shall not be increased above 90% of RATED THERMAL POWER unless the indicated AFD is within the target band as specified in the COLR and ACTION a.2.a) 1), above has been satisfied.

c. THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER unless the indicated AFD has not been outside of the target band as specified in the COLR for more than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months penalty deviation cumulative during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Power increases above 50% of RATED THERMAL POWER do not require being within the target band provided the accumulative penalty deviation is not violated.

4.2.1.1 The indicated AXIAL FLUX DIFFERENCE shall be determined to be within its limits during POWER OPERATION above 15% of RATED THERMAL POWER by: Monitoring the indicated AFD for each OPERABLE excore channel: In accordance with the Surveillance Frequency Control Program when the AFD Monitor Alarm is OPERABLE, and In accordance with the Surveillance Frequency Control Program for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after restoring the AFD Monitor Alarm to OPERABLE status. Monitoring and logging the indicated AXIAL FLUX DIFFERENCE for each OPERABLE excore channel in accordance with the Surveillance Frequency Control Program for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and in accordance with the Surveillance Frequency Control Program thereafter, when the AXIAL FLUX DIFFERENCE Monitor Alarm is inoperable.

The logged values of the indicated AXIAL FLUX DIFFERENCE shall be assumed to exist during the interval preceding each logging. 4.2.1.2 The indicated AFD shall be considered outside of its target band when at least 2 or more OPERABLE excore channels are indicating the AFD to be outside the target band. Penalty deviation outside of the target band shall be accumulated on a time basis of: One minute penalty deviation for each one minute of POWER OPERATION outside of the target band at THERMAL POWER levels equal to or above 50% of RATED THERMAL POWER, and One-half minute penalty deviation for each one minute of POWER OPERATION outside of the target band at THERMAL POWER levels below 50% of RATED THERMAL POWER. SALEM -UNIT 3/4 2-2 Amendment No. 282 POWER DISTRIBUTION LIMITS £URVEILLANCE REQUIREMENTS (Continued) 4.2.1.3 The target flux difference of each OPERABLE excore channel shall be determined by measurement in accordance with the Surveillance Frequency Control Program. The provisions of Specification 4.0.4 are not applicable.

4.2.1.4 The target flux difference shall be updated in accordance with the Surveillance Frequency Control Program by either determining the target flux difference pursuant to 4.2.1.3 above or by linear interpolation between the most recently measured value and 0% at the end of the cycle life. The provisions of Specification 4.0.4 are not applicable.

SALEM -UNIT 2 3/4 2-3 Amendment No. 282 POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 Fxy shall be evaluated to determine if Fa(Z) is within its limit by: Using the movable incore detectors to obtain a power distribution map: When THERMAL POWER is ::; 25%, but> 5% of RATED THERMAL POWER, or When the Power Distribution Monitoring System (PDMS) is inoperable; and increasing the Measured Fa(Z) by the applicable manufacturing and measurement uncertainties as specified in the COLR. Using the PDMS or the moveable incore detectors when THERMAL POWER is> 25% of RATED THERMAL POWER, and increasing the measured Fa(Z) by the applicable manufacturing and measurement uncertainties as specified in the COLR. Comparing the Fxy computed (Fxy c) obtained in b, above to: The Fxy limits for RATED THERMAL POWER (Fxy RTP) for the appropriate measured core planes given in e. and f., below, and The Fxy L = Fxy RTP [1 + PF xy where Fxy L is the limit for fractional THERMAL POWER operation expressed as a function of Fxt TP , PF xy is the power factor multiplier for Fxy in the COLR, and P is the fraction of RATED THERMAL POWER at which Fxy was measured. Remeasuring Fxy according to the following schedule: When Fxyc is greater than the F x/ TP limit for the appropriate measured core plane but less than the Fx/ relationship, additional core power distribution measurements shall be taken and Fxyc compared to Fxy RTP and F L. xy . Either within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding by 20% of RATED THERMAL POWER or greater, the THERMAL POWER at which Fxyc was last determined, or In accordance with the Surveillance Frequency Control Program, whichever occurs first. SALEM -UNIT 3/4 2-6 Amendment No. 282 POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) When the Fxy C is less than or equal to the Fxy RTP limit for the appropriate measured core plane, additional core power distribution measurements shall be taken and Fxyc compared to FXyRTP and FxyL in accordance with the Surveillance Frequency Control Program. The Fxy limit for Rated Thermal Power (Fxy RTP) shall be provided for all core planes containing bank "D" control rods and all unrodded core planes in the COLR per specification 6.9.1.9. The Fxy limits of e., above, are not applicable in the following core plane regions as measured in percent of core height from the bottom of the fuel: Lower core region from 0% to 15%, inclusive. Upper core region from 85% to 100%, inclusive. Grid plane regions at 17.8% +/- 2%,32.1% +/- 2%,46.4% +/- 2%, 60.6% +/- 2% and 74.9% +/- 2%, inclusive. Core plane regions within +/- 2% of core height (+/- 2.88 inches) about the bank demand position of the bank "D" control rods. Evaluating the effects of F:y on Fa(Z) to determine if Fa(Z) is within its limit whenever Fxyc exceeds Fxy . 4.2.2.3 When Fa(Z) is measured pursuant to specification 4.10.2.2, an overall measured Fa(Z) shall be obtained from a core power distribution measurement and increased by the applicable manufacturing and measurement uncertainties as specified in the COLR. SALEM -UNIT 3/42-7 Amendment No. 282 POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.3.1 FN LlH shall be determined to be within its limit by obtaining a core power distribution measurement:

a. Prior to operation above 75% of RATED THERMAL POWER after each fuel loading, and b. In accordance with the Surveillance Frequency Control Program. c. The provisions of Specification 4.0.4 are not applicable.

4.2.3.2 The measured FNLlH of 4.2.3.1 above, shall be increased by the applicable FNLlH uncertainties*

specified in the COLR. For Cycle 11, when the number of available movable detector thimbles is greater than or equal to 50% and less than 75% of the total, the 4% measurement uncertainty shall be increased to [4% + (3-T/14.5)(1%)1 where T is the number of available thimbles.

SALEM -UNIT 3/42-10 Amendment No. 282 POWER DISTRIBUTION LIMITING CONDITION FOR OPERATION Calculate the QUADRANT POWER TILT RATIO at least once per hour until: Either the QUADRANT POWER TILT RATIO is reduced to within its limit, or THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified at 95% or greater RATED THERMAL POWER. SURVEILLANCE REQUIREMENTS 4.2.4.1 The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50% of RATED THERMAL POWER by: Calculating the ratio in accordance with the Surveillance Frequency Control Program when the alarm is OPERABLE. Calculating the ratio in accordance with the Surveillance Frequency Control Program during steady-state operation when the alarm is inoperable.

4.2.4.2 The QUADRANT POWER TILT RATIO shall be determined to be within the limit when above 75% of RATED THERMAL POWER with one Power Range Channel inoperable by obtaining a core power distribution measurement*

to confirm that the normalized symmetric power distribution is consistent with the indicated QUADRANT POWER TILT RATIO in accordance with the Surveillance Frequency Control Program. Using either the movable incore detectors in the four pairs of symmetric thimble locations or the power distribution monitoring system. SALEM -UNIT 3/42-15 Amendment No. 282 POWER DISTRIBUTION LIMITS 3/4.2.5 DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB related parameters shall be maintained within the limits shown on Table 3.2-1: a. Reactor Coolant System Tavg. b. Pressurizer Pressure.

c. Reactor Coolant System Total Flow Rate. APPLICABILITY:

MODE 1 ACTION: With any of the above parameters exceeding its limit. restore the parameter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . SURVEILLANCE REQUIREMENTS 4.2.5.1 Each of the parameters of Table 3.2-1 shall be verified to be within their limits in accordance with the Surveillance Frequency Control Program. 4.2.5.2 The Reactor Coolant System Total Flow Rate shall be determined to be within the limits of Table 3.2-1 by performing a precision heat balance within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after achieving steady state conditions

<:! 90% RATED THERMAL POWER in accordance with the Surveillance Frequency Control Program. The provisions of Specification 4.0.4 are not applicable.

SALEM -UNIT 2 3/42-16 Amendment No. 282 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION 3.3.1.1 As a minimum, the reactor trip system instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE.

APPLICABILITY:

As shown in Table 3.3-1 As shown in Table 4.3.1.1.1 Each reactor trip system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies specified in the Surveillance Frequency Control Program unless otherwise noted in Table 4.3-1. 4.3.1.1.2 The logic for the interlocks shall be demonstrated OPERABLE prior to each reactor startup unless performed during the preceding 92 days. The total interlock function shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program during CHANNEL CALIBRATION testing of each channel affected by interlock operation.

4.3.1.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip function shall be verified to be within its limit in accordance with the Surveillance Frequency Control Program. Neutron detectors are exempt from response time testing. SALEM -UNIT 2 3/43-1 Amendment No. 282 TABLE REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE FUNCTIONAL UNIT Manual Reactor Trip Switch Power Range, Neutron Flux Power Range, Neutron Flux, High Positive Rate Deleted Intermediate Range, Neutron Flux Source Range, Neutron Flux Overtemperature I1T Overpower I1T Pressurizer Pressure--Low

10. Pressurizer Pressure--High

11. Pressurizer Water High Loss of Flow -Single Loop (7) (2), (3) (6) (6) MODES IN CHANNEL WHICH FUNCTIONAL SURVEILLANCE TEST(15) REQUIRED (9) 1,2, and" 1,2, and 3" 1,2 1,2 and" (16) and S/U(1) 2, 3,4,5 and" 1,2 1,2 1,2 1.2 1,2 1 SALEM -UNIT 3/43-11 Amendment No. 282 TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE FUNCTIONAL UNIT Loss of Flow Two Loops Steam Generator Water Level--Low-Low DELETED Undervoltage

-Reactor Coolant Pumps Underfrequency

-Reactor Coolant Pumps 18. Turbine Trip a. Low Autostop Oil Pressure b. Turbine Stop Valve Closure Safety Injection Input from ESF Reactor Coolant Pump Breaker Position Trip Reactor Trip Breaker Automatic Trip Logic N.A. N.A. N.A. N.A. N.A. N.A. N.A. N.A. CHANNEL CALIBRATION(15)

N.A. N.A. N.A. N.A. N.A. N.A. MODES IN CHANNEL WHICH FUNCTIONAL SURVEILLANCE TEST(15) REQUIRED 1 1,2 1 1 S/U(1) 1,2 S/U(1) 1,2 (4)(5) 1,2 (5)(11 )(13)(14) 1, 2 and * (5) 1, 2 and

SALEM -UNIT 3/43-12 Amendment No. 282 TABLE 4.3-1

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

(1) -If not performed in previous 31 days. (2) Heat balance only, above 15% of RATED THERMAL POWER. (3) Compare incore to excore axial offset above 15% of RATED THERMAL POWER. Recalibrate jf absolute difference;::

3 percent. (4) Manual SSPS functional input check in accordance with the Surveillance Frequency Control Program. (5) Each train or logic channel shall be tested in accordance with the Surveillance Frequency Control Program. (6) Neutron detectors may be excluded from CHANNEL CALIBRATION.

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

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

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

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

(12) DELETED (13) Verify operation of Bypass Breakers Shunt Trip function from local pushbutton while breaker is in the test position prior to placing breaker in service. (14) Perform a functional test of the Bypass Breakers U.V. Attachment via the SSPS. (15) Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table. (16) At the frequency specified in the Surveillance Frequency Control Program. SALEM -UNIT 2 3/43-13 Amendment No. 282

3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2.1 The Engineered Safety Feature Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4. APPLICABILITY:

As shown in Table 3.3-3. ACTION: a. With an ESFAS instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3-4, declare the channel inoperable and apply the applicable ACTION requirement of Table 3.3-3 until the channel is restored to OPERABLE status with the trip setpoint adjusted consistent with the Trip Setpoint value. b. With an ESFAS instrumentation channel inoperable, take the ACTION shown in Table 3.3-3. SURVEILLANCE REQUIREMENTS 4.3.2.1.1 Each ESFAS instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at frequencies specified in the Surveillance Frequency Control Program unless otherwise noted in Table 4.3-2. 4.3.2.1.2 The logic for the interlocks shall be demonstrated OPERABLE during the automatic actuation logic test. The total interlock function shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program during CHANNEL CALIBRATION testing of each channel affected by interlock operation.

4.3.2.1.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be verified to be within the limit in accordance with the Surveillance Frequency Control Program. The provisions of Specification 4.0.4 are not applicable to MSIV closure time testing. The provisions of Specification 4.0.4 are not applicable to the turbine driven auxiliary feedwater pump provided the surveillance is performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the secondary steam generator pressure is greater than 680 psig. SALEM -UNIT 2 3/43-14 Amendment No. 282 TABLE ENGINEERED SAFETY FEATURE ACTUATION SYSTEM SURVEILLANCE MODES IN CHANNEL WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK(7) CALI BRATION(7)

TEST(7) REQUIRED SAFETY INJECTION, TURBINE TRIP AND FEEDWATER ISOLATION Manual Initiation NA NA 1.2.3,4 Automatic Actuation Logic NA NA (2) 1,2,3,4 Containment Pressure--

(3) 1.2,3 High Pressurizer Pressure--Low 1,2,3 Differential Pressure 1,2.3 Between Steam High Steam Flow in Two Steam 1.2,3 Lines--High Coincident with Tavg--Low-Low or Steam Line Low CONTAINMENT SPRAY Manual Initiation NA NA 1,2,3,4 Automatic Actuation Logic NA NA (2) 1,2,3,4 Containment Pressure--

(3) 1,2,3 High-High SALEM -UNIT 3/43-33 Amendment No. 282 TABLE 4.3-2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM SURVEILLANCE FUNCTIONAL UNIT CHANNEL CHECK(7) CHANNEL CALISRATION(7)

CHANNEL FUNCTIONAL TEST(7) MODES IN WHICH SURVEILLANCE REQUIRED 3. CONTAINMENT ISOLATION

a. Phase "A" Isolation

1) Manual N.A. N.A. 1,2,3,4 2) From Safety Injection Automatic Actuation Logic N.A. N.A. (2) 1,2,3,4 b. Phase "S" Isolation

1) Manual N.A. N.A. 1,2,3,4 2) Automatic Actuation Logic N.A. N.A. (2) 1,2,3,4 3) Containment -High-High (3) 1,2,3 c. Containment Ventilation Isolation

1) Manual N.A. N.A. 1,2,3,4 2) Automatic Actuation Logic N.A. N.A. (2) 1,2,3,4 3) Containment Atmosphere Gaseous Radioactivity-High Per Surveillance Requirement 4.3.3.1 SALEM -UNIT 2 3/43-34 Amendment No. 282 TABLE 4.3-2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM SURVEILLANCE CHANNEL FUNCTIONAL UNIT CHECK(7) CHANNEL CALI BRA TION(7) CHANNEL FUNCTIONAL TEST(7) MODES IN WHICH SURVEILLANCE REQUIRED 4. STEAM LINE ISOLATION Manual NA NA 1,2,3** Automatic Actuation Logic NA NA (2) 1,2,3 Containment Pressure--

(3) 1,2,3 High-High Steam Flow in Two Steam 1,2,3 Lines--High Coincident with Tavg--Low-Low or Steam Line Low TURBINE TRIP AND FEEDWATER ISOLATION Steam Generator Water 1,2,3 Level--High-High SAFEGUARDS EQUIPMENT CONTROL SYSTEM (SEC) LOGIC Inputs NA NA (6) 1,2,3,4 Logic, Timing and Outputs NA NA (1 ) 1,2,3,4

UNDERVOLTAGE, VITAL BUS Loss of Voltage 1,2,3 Sustained Degraded 1,2,3 Voltage SALEM -UNIT 3/43-35 Amendment No. 282 TABLE 4.3-2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM SURVEILLANCE REQUIREMENTS FUNCTIONAL UNIT CHANNEL CHECK(7) CHANNEL CALIBRATION(7)

CHANNEL FUNCTIONAL TEST(7) MODES IN WHICH SURVEILLANCE REQUIRED 8. AUXILIARY FEEDWATER

a. Automatic Actuation Logic NA NA (2) 1,2,3 b. NOT USED c. Steam Generator Water Level--Low-Low 1,2,3 d. Undervoltage

-RCP .1,2 e. S.1. See 1 above (All S.1. surveillance requirements)

f. Trip of Main Feedwater Pumps NA NA S/U(4) 1,2 g. Station Blackout See 6 and 7 above (SEC and UN Vital Bus) 9. SEMIAUTOMATIC TRANSFER TO RECIRCULATION

a. RWST Low Level 1,2,3 b. Automatic Initiation Logic NA NA (2) 1,2,3,4 SALEM -UNIT 2 3/43-36 Amendment No. 282 TABLE 4.3-2 TABLE

Outputs are up to, but not including, the Output Relays. ** The provisions of Specification of 4.0.4 are not applicable.

(1 ) Each logic channel shall be tested in accordance with the Surveillance Frequency Control Program. The CHANNEL FUNCTION TEST of each logic channel shall verify that its associated diesel generator automatic load sequence timer is OPERABLE with the interval between each load block within 1 second of its design interval.

(2) Each train or logic channel shall be tested in accordance with the Surveillance Frequency Control Program. (3) The CHANNEL FUNCTIONAL TEST shall include exercising the transmitter by applying either a vacuum or pressure to the appropriate side of the transmitter.

(4) If not performed in the previous 92 days. (5) NOT USED (6) Inputs from undervoltage, Vital Bus, shall be tested in accordance with the Surveillance Frequency Control Program. Inputs from Solid State Protection System, shall be tested in accordance with the Surveillance Frequency Control Program. (7) Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table. SALEM -UNIT 2 3/43-37 Amendment No. 282 INSTRUMENTATION 3/4.3.3 MONITORING INSTRUMENTATION RADIATION MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.1 The radiation monitoring instrumentation channels shown in Table 3.3-6 shall be OPERABLE with their alarm/trip setpoints within the specified limits. APPLICABILITY:

As shown in Table 3.3-6. ACTION: a. With a radiation monitoring channel alarm/trip setpoint exceeding the value shown in Table 3.3-6, adjust the setpoint to within the limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or declare the channel inoperable.

b. With one or more radiation monitoring channels inoperable, take the ACTION shown in Table 3.3-6. c. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.1 Each radiation monitoring instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations in accordance with the Surveillance Frequency Control Program. SALEM -UNIT 2 3/43-38 Amendment No. 282 TABLE Page 3/4 3-42 Deleted SALEM -UNIT 2 3/43-41 Amendment No. 282

REMOTE SHUTDOWN INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.5 The remote shutdown monitoring instrumentation channels shown in Table 3.3-9 shall be OPERABLE with readouts displayed external to the control room. APPLICABILITY:

MODES 1, 2 and 3. ACTION: With the number of OPERABLE remote shutdown monitoring channels less than required by Table 3.3-9, restore the inoperable channel to OPERABLE status within 7 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . SURVEILLANCE REQUIREMENTS 4.3.3.5 Each remote shutdown monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, and CHANNEL CALIBRATION operations in accordance with the Surveillance Frequency Control Program. SALEM -UNIT 3/43-43 Amendment No. 282 TABLE SALEM -UNIT 2 3/43-45 Amendment No.

INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.7 The accident monitoring instrumentation channels shown in Table 3.3-11 shall be operable.

APPLICABILITY:

MODES 1,2, and 3. ACTION: a. As shown in Table 3.3-11. 4.3.3.7 Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies specified in the Surveillance Frequency Control Program unless otherwise noted in Table 4.3-11. SALEM -UNIT 2 3/43-50 Amendment No. 282 TABLE SURVEILLANCE REQUIREMENTS ACCIDENT MONITORING CHANNEL CHANNEL CHANNEL FUNCTIONAL INSTRUMENT CHECKS(1)

CALIBRATION(1)

TEST(1) Reactor Coolant Outlet Temperature

-N.A. T HOT (Wide Range) Reactor Coolant Inlet Temperature

-N.A. T COLD (Wide Range) Reactor Coolant Pressure (Wide Range) N.A. Pressurizer Water Level N.A. Steam Line Pressure N.A. Steam Generator Water Level N.A. (Narrow Range) Steam Generator Water Level N.A. (Wide Range) Refueling Water Storage Tank Water Level N.A. deleted 10. Auxiliary Feedwater Flow Rate N.A. 11. Reactor Coolant System N.A.* N.A. Margin Monitor # Auxiliary Feedwater System is used on each startup and flow rate indication is verified at that time.

The instruments used to develop RCS subcooling margin are calibrated in accordance with the Surveillance Frequency Control Program; the monitor will be compared with calculated subcooling margin for known input values in accordance with the Surveillance Frequency Control Program. SALEM -UNIT 3/43-52 Amendment 1\10. 282 TABLE 4.3-11 SURVEILLANCE REQUIREMENTS ACCIDENT MONITORING CHANNEL CHANNEL CHANNEL FUNCTIONAL INSTRUMENT CHECKS(1)

CALIBRATION(1)

TEST(1) 12. PORV Position N.A. 13. PORV Block Valve Position N.A.

14. Pressurizer Safety Valve Position N.A. 15. Containment Pressure -Narrow N.A. 16. Containment Pressure -Wide N.A. 17. Containment Water Level -Wide N.A. 18. Core Exit N.A. 19. Reactor Vessel Level N.A. System (RVLlS) 20. Containment High Range Accident Radiation monitor 21. Main Steam line Discharge (Safety Valves and Atmospheric Steam Dumps) Monitor Table Notation (1 ) Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table. Unless the block valve is closed in order to meet the requirements of Action b, or c in specification 3.4.5. SALEM -UNIT 3/43-52a Amendment No. 282 INSTRUMENTATION RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.8 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3-12 shall be OPERABLE to ensure that the limits of ODCM Control 3.11.1.1 are not exceeded.

APPLICABILITY:

At all times. ACTION: Not Used With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-12. Exert best efforts to return the instrument to OPERABLE status within 30 days and, if unsuccessful, explain in the next annual radioactive effluent release report why the inoperability was not corrected in a timely manner. The provisions of Specification 3.0.3 are not applicable.

SURVEILLAtiCE REQUIREMENTS 4.3.3.8 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies specified in the Surveillance Frequency Control Program unless otherwise noted in Table 4.3-12. SALEM -UNIT 3/43-53 Amendment No. 282 TABLE RADIOACTIVE LIQUID EFFLUENT MONITORING SURVEILLANCE CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONAL INSTRUMENT CHECK(1) CHECK(1) CALIBRATION(1)

TEST(1) Not Used Not Used Not Used TANK LEVEL INDICATING DEVICES** Temporary Outside Storage D* Tanks as TABLE NOTATION

During liquid additions to the ** If tank level indication is not provided, verification will be done by visual Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table. SALEM -UNIT 3/43-55 Amendment No. 282 I NSTRUMENTA POWER DISTRIBUTION MONITORING LIMITING CONDITION FOR OPERATION APPLICABILTY:

MODE 1, above 25% RATED THERMAL POWER With any of the operability criteria listed in 3.3.3.14.a, 3.3.3.14.b, or 3.3.3.14.c not met, correct the deficient operability condition, or declare the PDMS inoperable and use the incore movable detector system to obtain any required core power distribution measurements.

Increase the measured core peaking factors using the values listed in the COLR for the PDMS inoperable condition.

The provisions of Specification 3.0.3 are not applicable.

SURVEI LLANCE REQUI REMENTS 4.3.3.14.1 The operability criteria listed in 3.3.3.14.a, 3.3.3.14.b, and 3.3.3.14.c shall be verified to be satisfied prior to acceptance of the PDIVIS core power distribution measurement results. 4.3.3.14.2 Calibration of the PDMS is required: In accordance with the Surveillance Frequency Control Program when the minimum number and core coverage criteria as defined in 3.3.3.14.b.1 and 3.3.3.14.b.2 are satisfied, or In accordance with the Surveillance Frequency Control Program when only the minimum number criterion as defined in 3.3.3.14.b.3 is satisfied.

SALEM* UNIT Amendment No. 282 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER OPERATION 3.4.1.1 All reactor coolant loops shall be in APPLICABILITY:

MODES 1 and 2. With less than the above required reactor coolant loops in operation, be in at least STANDBY within 1 SURVEILLANCE REQUIREMENT 4.4.1.1 The above required reactor coolant loops shall be verified to be in operation and circulating reactor coolant in accordance with the Surveillance Frequency Control Program. *See Special Test Exception 3.10.4. SALEM -UNIT 2 3/44-1 Amendment No. 282 REACTOR COOLANT SYSTEM HOT STANDBY SURVEILLANCE REQUIREMENTS 4.4.1.2.1 At least the above required reactor coolant pumps, if not in operation, shall be determined to be OPERABLE in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignments and indicated power availability.

4.4.1.2.2 At least one cooling loop shall be verified to be in operation and circulating reactor coolant in accordance with the Surveillance Frequency Control Program. 4.4.1.2.3 The required steam generator(s) shall be determined OPERABLE by verifying secondary side water level to be greater than or equal to 5% (narrow range) in accordance with the Surveillance Frequency Control Program. All reactor coolant pumps may be de-energized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided:

(1) no operations are permitted that would cause dilution of the reactor coolant system boron concentration (2) core outlet temperature is maintained at least 10°F below saturation temperature, and (3) the rod control system is de-energized** The rod control system shall be considered de-energized when one or more of the following conditions exist: Both Rod Drive MG set motor breakers are open. Both Rod Drive MG set generator breakers are open. A combination of at least three of the Reactor Trip and/or Reactor Trip Bypass Breakers are open. If none of the above conditions for de-energizing the rod control system are met; the system shall be considered energized.

SALEM -UNIT 3/44-2a Amendment No. 282 REACTOR COOLANT 4.4.1.3.2 The required reactor coolant pump(s), if not in operation, shall be determined to be OPERABLE in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignments and indicated power availability.

4.4.1.3.3 The required steam generator(s) shall be determined OPERABLE by verifying secondary side water level to be greater than or equal to 5% (narrow range) in accordance with the Surveillance Frequency Control Program. 4.4.1.3.4 At least one coolant loop shall be verified to be in operation and circulating reactor coolant in accordance with the Surveillance Frequency Control Program. SALEM -UNIT 2 3/44-4 Amendment No. 282 REACTOR COOLANT COLD SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.4 Two# residual heat removal loops shall be OPERABLE*

and at least one RHR loop shall be in operation.**

APPLICABILITY:

MODE 5.## ACTION: a. With less than the above required loops operable, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible.

b. With no RHR loop in operation; suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation.

4.4.1.4 At least one residual heat removal loop shall be verified to be in operation and circulating reactor coolant in accordance with the Surveillance Frequency Control Program. # One RHR loop may be inoperable for up to two hours for surveillance testing, provided the other RHR loop is OPERABLE and in operation.

Additionally, four filled reactor coolant loops, with at least two steam generators with their secondary side water levels greater than or equal to 5% (narrow range), may be substituted for one residual heat removal loop. ## A reactor coolant pump shall not be started with one or more of the RCS cold leg temperatures less than or equal to 312°F unless 1) the pressurizer water volume is less than 1650 cubic feet (equivalent to approximately 92% of level), or 2) the secondary water temperature of each steam generator is less than 50°F above each of the RCS cold leg temperatures.

Systems supporting RHR loop operability may be excepted as follows: a. The normal or emergency power source may be inoperable.

- The residual heat removal pumps may be de-energized for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months provided 1) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and 2) core outlet temperature is maintained at least 10°F below saturation temperature.

SALEM -UNIT 2 3/44-4a Amendment No. 282 REACTOR COOLANT 3/4.4.4 PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.4 The pressurizer shall be OPERABLE with a water volume of less than or equal to 1650 cubic feet (92% indicated level), and at least two groups of pressurizer heaters each having a capacity of 150 kw and capable of being powered from an emergency power supply. APPLICABILITY:

MODES 1, 2, and 3. ACTION: a. With the pressurizer inoperable due to an inoperable emergency power supply to the pressurizer heaters either restore the inoperable emergency power supply 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 and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . b. With the pressurizer otherwise inoperable, be in at least HOT STANDBY with the reactor trip breakers open within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . 4.4.4.1 The pressurizer water volume shall be determined to be within its limit in accordance with the Surveillance Frequency Control Program. 4.4.4.2 The capacity of each of the above required groups of pressurizer heaters shall be verified by measuring circuit current in accordance with the Surveillance Frequency Control Program. 4.4.4.3 The emergency power supply for the pressurizer heaters shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by manually transferring power from the normal to the emergency power supply and energizing the heaters. SALEM UNIT 2 3/44-7 Amendment No. 282 REACTOR COOLANT SYSTEM 3/4.4.5 RELIEF VALVES 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 REACTOR COOLANT 3/4.4.7 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.7.1 The following Reactor Coolant System leakage detection systems shall be OPERABLE:

a. The containment atmosphere particulate radioactivity monitoring system, b. The containment pocket sump level monitoring system, and c. Either the containment fan cooler condensate flow rate or the containment atmosphere gaseous radioactivity monitoring system. APPLICABILITY:

MODES 1, 2, 3 and 4. ACTION: With only two of the above required leakage detection systems OPERABLE, operation may continue for up to 30 days provided grab samples of the containment atmosphere are obtained and analyzed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the required gaseous and/or particulate radioactivity monitoring system is inoperable; 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 4.4.7.1 The leakage detection systems shall be demonstrated OPERABLE by: Containment atmosphere particulate and gaseous (if being used) monitoring systems-performance of CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST in accordance with the Surveillance Frequency Control Program. Containment pocket sump level and containment fan cooler condensate flow rate (if being used) monitoring systems-performance of CHANNEL CALIBRATION in accordance with the Surveillance Frequency Control Program. SALEM -UNIT 3/44-16 Amendment No. 282 OPERATIONAL LEAKAGE LIMITING CONPITION FOR OPERATION 3.4.7.2 Reactor Coolant System leakage shall be limited to: No PRESSURE BOUNDARY LEAKAGE, 1 GPM UNIDENTIFIED LEAKAGE, 150 gallons per day primary-to-secondary leakage through anyone steam generator, and 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and NOT USED 1 GPM leakage at a Reactor Coolant System pressure of 2230 +/- 20 psig from any Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1. APPLICABILITY:

MODES 1, 2, 3 and 4 ACTION: With any PRESSURE BOUNDARY LEAKAGE, or primary-to-secondary leakage not within limit, be in at least HOT STANDBY within 6 hours6.944444e-5 days 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 . With any Reactor Coolant System leakage greater than anyone of the above limits, excluding PRESSURE BOUNDARY LEAKAGE and leakage from Reactor Coolant System Pressure Isolation Valves, and primary-to-secondary leakage, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 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 and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days 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 4.4.7.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by: Monitoring the containment atmosphere particulate radioactivity monitor in accordance with the Surveillance Frequency Control Program. Monitoring the containment sump inventory in accordance with the Surveillance Frequency Control Program. SALEM -UNIT 3/44-17 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: In accordance with the Surveillance Frequency Control Program. 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. Prior to returning the valve to service following maintenance repair or replacement work on the valve. 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 3/44-18 Amendment No. 282 TABLE PRIMARY COOLANT SPECIFIC ACTIVITY AND ANALYSIS TYPE OF MEASUREMENT AND ANALYSIS Gross Activity Determination Isotopic Analysis for DOSE EQU IVALENT 1-131 Concentration Radiochemical for E Determination Isotopic Analysis for Iodine Including 1-131,1-133, and 1-135 SAMPLE AND ANALYSIS FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program* Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , whenever the specific activity exceeds 1.0 DOSE EQUIVALENT 1-131 or 100/E jJCi/gram, and One sample between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months following a THERMAL POWER change exceeding 15 percent of the RATED THERMAL POWER within a one hour period. MODES IN SAMPLE 1,2,3,4 1 1 Until the specific activity of the primary coolant system is restored within its limits. Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last subcritical for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or longer. SALEM -UNIT 3/44-25 Amendment No. 282 REACTOR COOLANT 3/4.4.10 PRESSURE/TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.10.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2 and 3.4-3 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with: a. A maximum heatup of 100°F in anyone hour period, b. A maximum cooldown of 100°F in anyone hour period, and c. A maximum temperature change of less than or equal to 5°F in anyone hour period, during hydrostatic testing operations above system design pressure.

APPLICABILITY:

At all times. ACTION: With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation 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 reduce the RCS T avg and pressure to less than 200"F and 500 psig, respectively, within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . 4.4.10.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits in accordance with the Surveillance Frequency Control Program during system heatup, cooldown, and inservice leak and hydrostatic testing operations.

4.4.10.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, at the intervals required by 10 CFR 50, Appendix H. The results of these examinations shall be used to update Figures 3.4-2 and 3.4-3. SALEM -UNIT 2 3/44-27 Amendment No. 282 REACTOR COOLANT PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.10.2 The pressurizer temperature shall be limited to: a. A maximum heatup of 100°F in anyone hour period, b. A maximum cooldown of 200°F in anyone hour period, and c. A maximum spray water temperature differential of 320°F. APPLICABILITY:

At all times. ACTION: With the pressurizer temperature limits in excess of any of the above limits, restore the temperature to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the pressurizer; determine that the pressurizer remains acceptable for continued operation 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 reduce the pressurizer pressure to less than 500 psig within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . 4.4.10.2 The pressurizer temperatures shall be determined to be within the limits in accordance with the Surveillance Frequency Control Program during system heatup or cooldown.

The spray water temperature differential shall be determined to be within the limit in accordance with the Surveillance Frequency Control Program during auxiliary spray operation.

SALEM -UNIT 2 3/44-30 Amendment No. 282 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 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. Performance of a CHANNEL CALIBRATION on the POPS actuation channel in accordance with the Surveillance Frequency Control Program. Verifying the POPS isolation valve is open in accordance with the Surveillance Frequency Control Program when the POPS is being used for overpressure protection. Testing pursuant to the 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 3/44-32 Amendment No. 282 REACTOR COOLANT 3/4.4.12 HEAD VENTS LIMITING CONDITION FOR OPERATION 3.4.12 Four reactor vessel head vent paths shall be operable with the vent paths closed. A vent path consists of at least two head vent valves in series, powered from vital sources, and associated flowpath.

APPLICABILITY:

MODES 1, 2, 3 AND 4. ACTION: With one, two or three reactor vessel head vent path(s) inoperable, STARTUP and/or POWER OPERATION may continue provided the inoperable vent path(s) is maintained closed with the valve actuators key locked in the closed position; restore the inoperable vent path(s) to OPERABLE status within 30 days, or, be in HOT STANDBY within 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 . With four reactor vessel head vent paths inoperable; maintain the inoperable vent paths closed with power removed from the valve actuators of all the vent valves in the inoperable vent paths, and restore at least one of the vent paths to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY within 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 4.4.12 Reactor vessel head vent system vent paths shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by: Verifying all manual isolation valves in each vent path are locked in the open position. Cycling each valve in the vent paths through at least one complete cycle of full travel from the control room during COLD SHUTDOWN or REFUELING. Verifying flow through the reactor vessel head vent system vent path during venting. SALEM -UNIT 3/44-34 Amendment No. 282 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) ACCUMULATORS LIMITING CONDITION FOR OPERATION 3.5.1 Each reactor coolant system accumulator shall be OPERABLE with: The isolation valve open, A contained volume of between 6223 and 6500 gallons of borated water, A boron concentration of between 2200 and 2500 ppm, and d. A nitrogen cover-pressure of between 595.5 and 647.5 psig. APPLICABILITY:

MODES 1,2 and 3*. ACTION: With one accumulator inoperable, except as a result of a closed isolation valve or boron concentration outside the required limits, restore the inoperable accumUlator to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . With one accumulator inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in HOT STANDBY within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . With the boron concentration of one accumulator outside the required limits, restore the boron concentration to within the required limits 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 and reduce pressurizer pressure to less than or equal to 1000 psig within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . SURVEI LLANCE REQUIREMENTS 4.5.1 Each accumulator shall be demonstrated OPERABLE: In accordance with the Surveillance Frequency Control Program by: Verifying the water level and nitrogen cover-pressure in the tanks, and Verifying that each accumulator isolation valve is open.

Pressurizer Pressure above 1000 psig. SALEM -UNIT 3/45-1 Amendment No. 282 EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

b. In accordance with the Surveillance Frequency Control Program and within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after each solution volume increase of greater than or equal to 1 % of tank volume by verifying the boron concentration of the accumulator solution.

c. In accordance with the Surveillance Frequency Control Program when the RCS pressure is greater than 1000 psig by verifying that the power lockout switch is in lockout. In accordance with the Surveillance Frequency Control Program by verifying that each accumulator isolation valve opens automatically upon receipt of a safety injection test signal. SALEM -UNIT 3/4 5-2 Amendment No. 282 EMERGENCY CORE COOLING SURVEILLANCE REQUIREMENTS Each ECCS subsystem shall be demonstrated OPERABLE: In accordance with the Surveillance Frequency Control Program by: Verifying that the following valves are in the indicated positions with power to the valve operators removed: Valve Number Valve Valve Position a. 2 SJ 69 a. RHR pump suction a. open b. 2 SJ 30 b. SI pump suction b. open c. 21 SJ 40 c. SI discharge to hot legs c. closed d. 22 SJ 40 d. SI discharge to hot legs d. closed e. 2 RH 26 e. RHR discharge to hot legs e. closed f. 21 SJ 49 f. RHR discharge to cold legs f. open g. 22 SJ 49 g. RHR discharge to cold legs g. open h. 2 CS 14# h. Spray additive tank discharge

h. open i. 2 SJ 135 i. SI discharge to cold legs i. open j. 2 SJ 67 j. SI recirc. line isolation

j. open k. 2 SJ 68 k. SI recirc. line isolation

k. open I. 21 SJ 44 I. Containment sump isolation valve I. closed m. 22 SJ 44 m. Containment sump isolation valve m. closed Verifying that the following valves are in the indicated positions:

Valve Number Valve Function Valve Position a. RH 19 a. RHR crosstie valve a. Open b. 22 RH 19 b. RHR crosstie valve b. Open 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. Verifying that the ECCS piping is full of water by venting the ECCS pump casings and accessible discharge piping high points. # If inoperable, the applicable Technical Specification is SALEM -UNIT 2 3/4 5-4 Amendment No.

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the containment sump and cause restriction of the pump suctions during LOCA conditions.

This visual inspection shall be performed: For all accessible areas of the containment prior to establishing CONTAINMENT INTEGRITY, and In accordance with the Surveillance Frequency Control Program the areas affected within containment by containment entry and during the final entry when CONTAINMENT INTEGRITY is established. In accordance with the Surveillance Frequency Control Program by: A visual inspection of the containment sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or corrosion. In accordance with the Surveillance Frequency Control Program, during shutdown, by: Verifying that each automatic valve in the flow path actuates to its correct position on a safety injection test signal. Verifying that each of the following pumps start automatically upon receipt of a safety injection test signal: a) Centrifugal charging pump b) Safety injection pump c) Residual heat removal pump SALEM UNIT 3/4 5-5 Amendment No. 282 w EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 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: Centrifugal Charging pump 2338 psi TDH Safety Injection pump 1369 psi TDH Residual Heat Removal pump 165 psi TDH g. By verifying the correct position of each of the following ECCS throttle valves: 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. In accordance with the Surveillance Frequency Control Program. HPSI System LPSI Valve Number Valve 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 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: 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 3/45-6 Amendment No. 282 EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS

-Tavg < 350°F 4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per applicable Surveillance Requirements of 4.5.2. 4.5.3.2 All safety injection pumps and centrifugal charging pumps, except the above required OPERABLE pump, shall be demonstrated to be inoperable in accordance with the Surveillance Frequency Control Program while in MODE 4 and 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 by either of the following methods: a. By verifying that the motor circuit breakers have been removed from their electrical power supply circuits or, b. By verifying that the pump is in a recirculation flow path and that two independent means of preventing RCS injection are utilized.

SALEM -UNIT 2 3/4 5-8 Amendment No. 282 EIVIERGENCY CORE COOLING SEAL INJECTION FLOW LIMITING CONDITION FOR OPERATION 3.5.4 Reactor coolant pump seal injection flow shall be :540 gpm with centrifugal charging pump discharge header pressure 2430 psig and the charging flow control valve full open. APPLICABILITY:

MODES 1,2, and 3 ACTION: With seal injection flow not within the limit, adjust manual seal injection throttle valves to give a flow within the limit with the charging pump discharge pressure 2430 psig and the charging flow control valve full open within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 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 and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . SURVEILLANCE REQUIREMENTS 4.5.4 In accordance with the Surveillance Frequency Control Program, verify manual seal injection throttle valves are adjusted to give a flow within the limit with centrifugal charging pump discharge header pressure 2430 psig, and the charging flow control valve full open. The provisions of Specification 4.0.4 are not applicable for entry into Mode 3. This exemption is allowed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after the Reactor Coolant System pressure stabilizes at 2235 +/- 20 psig. SALEM -UNIT 2 3/45-8a Amendment No. 282 EMERGENCY CORE COOLING SYSTEMS REFUELING WATER STORAGE TANK 3.5.5 The refueling water storage tank (RWST) shall be OPERABLE with: a. A contained volume of between 364,500 and 400,000 gallons of borated water. b. A boron concentration of between 2,300 and 2,500 ppm, and c. A minimum water temperature of 35°F. APPLICABILITY:

MODES 1, 2, 3 and 4. ACTION: With the refueling water storage tank inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within 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 . 4.5.5 The RWST shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by: 1. Verifying the water level in the tank, and 2. Verifying the boron concentration of the water. b. In accordance with the Surveillance Frequency Control Program by verifying the RWST temperature when the outside air temperature is < 35°F. SALEM -UNIT 2 3/4 5-9 Amendment No. 282 3/4.6 CONTAINMENT 3/4.6.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 Primary CONTAINMENT INTEGRITY shall be APPLICABILITY:

MODES 1, 2, 3 and Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY one hour 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 . 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated: In accordance with the Surveillance Frequency Control Program by verifying that each containment manual valve or blind flange that is located outside containment and required to be closed during accident conditions is closed, except for containment isolation valves that are open under administrative controls.

Valves and blind flanges in high radiation areas may be verified by use of administrative controls. Prior to entering Mode 4 from Mode 5 if not performed within the last 92 days by verifying that each containment manual valve or blind flange that is located inside containment and required to be closed during accident conditions is closed, except for containment isolation valves that are open under administrative controls.

Valves and blind flanges in high radiation areas may be verified by use of administrative controls. By verifying that each containment air lock is OPERABLE per Specification 3.6.1.3. After each closing of a penetration subject to Type B testing, except containment air locks, if opened following a Type A or B test, by leak rate testing in accordance with the Containment Leakage Rate Testing Program. In accordance with the Surveillance Frequency Control Program by verifying that the surveillance requirements of 4.6.2.3.a are met for penetrations associated with the containment fan coil units. In accordance with the Surveillance Frequency Control Program by verifying that the surveillance requirements of 4.6.2.3.d are met for penetrations associated with the containment fan coil units. SALEM -UNIT 3/46-1 Amendment No. 282 CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATIONS (Continued) One or more containment air locks inoperable for reasons other than condition

a. or b. Immediately initiate action to evaluate overall containment leakage per LCO 3.6.1, and: Verify that at least one door is closed in the affected air lock within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and: Restore the air lock to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . If the ACTIONS and associated completion times of a., b., or c. cannot be met, be in Hot Standby within 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 4.6.1.3 Each containment air lock shall be demonstrated OPERABLE: By verifying seal leakage rate in accordance with the Containment Leakage Rate Testing program. By conducting an overall air lock leakage test in accordance with the Containment Leakage Rate Testing Program. In accordance with the Surveillance Frequency Control Program by verifying that only one door in each air lock can be opened at a time. SALEM -UNIT 3/46-5 Amendment No. 282 CONTAINMENT INTERNAL 3.6.1.4 Primary containment internal pressure shall be maintained between -1.5 and +0.3 APPLICABILITY:

MODES 1, 2, 3 and With the containment internal pressure outside of the limits above, restore the internal to within the limits within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 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 and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . 4.6.1.4 The primary containment internal pressure shall be determined to be within the limits in accordance with the Surveillance Frequency Control Program.

SALEM -UNIT 2 3/4 6-6 Amendment No. 282 CONTAINMENT AIR TEMPERATURE LIMITING CONDITION FOR OPERATION 3.6.1.5 Primary containment average air temperature shall not exceed APPLICABILITY:

MODES 1, 2, 3 and With the containment average air temperature greater than 120°F, reduce the average temperature to within the limit within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 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 and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . 4.6.1.5 Verify containment average air temperature is within limit in accordance with the Surveillance Frequency Control Program. SALEM -UNIT 2 3/4 6-7 Amendment No. 282 3/4.6.2 DEPRESSURIZATION AND COOLING CONTAINMENT SPRAY 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. 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 SPRAY ADDITIVE SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.2 The spray additive system shall be OPERABLE with: a. A spray additive tank containing a volume of between 2568 and 4000 gallons of between 30 and 32 percent by weight NaOH solution, and b. Two spray additive eductors each capable of adding NaOH solution from the chemical additive tank to a containment spray system pump flow. APPLICABILITY:

MODES 1, 2, 3 and 4, ACTION: With the spray additive system inoperable, restore the 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 spray additive 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 . 4.6.2.2 The spray additive 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. In accordance with the Surveillance Frequency Control Program by: 1. Verifying the solution level in the tank, and 2. Verifying the concentration of the NaOH solution by chemical analysis.

c. In accordance with the Surveillance Frequency Control Program during shutdown, by verifying that each automatic valve in the flow path actuates to its correct position on a Containment High-High pressure test signal. d. In accordance with the Surveillance Frequency Control Program by: 1. Verifying a NaOH solution flow rate of 12.0 + 3.0 gpm from the spray additive tank through sample valve 2CS61 with the spray additive tank at 2.5 +/- 0.5 psig and SALEM -UNIT 2 3/46-11 Amendment No. 282 CONTAINMENT SURVEILLANCE REQUIREMENTS (Continued) In accordance with the Surveillance Frequency Control Program by: Verifying the water level in each service water accumulator vessel is greater than or equal to 226 inches and less than or equal to 252 inches. Verifying the temperature in each service water accumulator vessel is greater than or equal to 55°F and less than or equal to 95°F. Verifying the nitrogen cover pressure in each service water accumulator vessel is greater than or equal to 135 psig and less than or equal to 160 psig. In accordance with the Surveillance Frequency Control Program by: Starting (unless already operating) each fan from the control room in low speed. Verifying that each fan operates for at least 15 minutes in low speed. Verifying a cooling water flow rate of greater than or equal to 1300 gpm to each cooler. In accordance with the Surveillance Frequency Control Program by verifying that on a safety injection test Signal: Each fan starts automatically in low speed. The automatic valves and dampers actuate to their correct positions and that the cooling water flow rate to each cooler is greater than or equal to 1300 gpm. In accordance with the Surveillance Frequency Control Program by verifying that on a loss of offsite power test signal, each service water accumulator vessel discharge valve response time is within limits. SALEM -UNIT 3/46-13 Amendment No. 282 CONTAINMENT SURVEILLANCE REQUIREMENTS 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 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. 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. 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: Required Containment Purge Supply and Exhaust Isolation Valves in accordance with the Surveillance Frequency Control Program. 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 3/46-15 Amendment No. 282 PLANT AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least three independent steam generator auxiliary feedwater pumps and associated manual activation switches in the control room and flow paths shall be OPERABLE with: Two feedwater pumps, each capable of being powered from separate vital busses, and One feedwater pump capable of being powered from an OPERABLE steam supply system. APPLICABILITY:

MODES 1, 2 and 3. ACTION: With one auxiliary feedwater pump inoperable, restore the required auxiliary feedwater 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 within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . With two auxiliary feedwater pumps inoperable be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . With three auxiliary feedwater pumps inoperable, immediately initiate corrective action to restore at least one auxiliary feedwater pump to OPERABLE status as soon as possible. LCO 3.0A.b is not applicable.

SURVEILLANCE REQUIREMENTS 4.7.1.2 Each auxiliary feedwater pump shall be demonstrated OPERABLE: In accordance with the Surveillance Frequency Control Program by: Verifying that each non-automatic valve in the flow path that is not locked, sealed or otherwise secured in position, is in its correct position. Verify the manual maintenance valves in the flow path to each steam generator are locked open. SALEM -UNIT 3/4 7-5 Amendment No. 282 PLANT SURVEILLANCE REQUIREMENTS (Continued) In accordance with the Surveillance Frequency Control Program by: Verify that the developed head of each motor driven pump at the flow test point is greater than or equal to the required developed head. Verify that the developed head of the steam driven pump at the flow test point is greater than or equal to the required developed head when the steam generator pressure is > 680 psig. The provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after secondary side pressure is greater than 680 psig. In accordance with the Surveillance Frequency Control Program by: Verifying that each auxiliary feedwater automatic valve that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal. Verifying that each auxiliary feedwater pump starts automatically on an actual or simulated actuation signal. The provisions of Specification 4.0.4 are not applicable to the turbine driven auxiliary feedwater pump, provided the surveillance is performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the secondary steam generator pressure is greater than 680 psig. SALEM -UNIT 3/4 7-6 Amendment No. 282 PLANT AUXILIARY FEED STORAGE TANK LIMITING CONDITION FOR OPERATION 3.7.1.3 The auxiliary feed storage tank (AFST) shall be OPERABLE with a contained volume at least 200,000 gallons of APPLICABILITY:

MODES 1, 2 and With the auxiliary feed storage tank inoperable, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Restore the AFST to OPERABLE status or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , or Demonstrate the OPERABILITY of a demineralized water or a fire protection/domestic water storage tank as a backup supply to the auxiliary feedwater pumps and restore the auxiliary feed storage tank to OPERABLE status within 7 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . 4.7.1.3.1 The auxiliary feed storage tank shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by verifying the water level is within its limits when the tank is the supply source for the auxiliary feedwater pumps. 4.7.1.3.2 A demineralized water storage tank shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by verifying the tank contains greater than or equal to 200,000 gallons of water and by verifying proper alignment of valves for taking suction from this tank when it is the supply source for the auxiliary feedwater pumps. 4.7.1.3.3 A fire protection/domestic water storage tank shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by verifying the tank contains greater than or equal to 200,000 gallons of water and by verifying proper alignment of valves for taking suction from this tank when it is the supply source for the auxiliary feedwater pumps. 4.7.1.3.4 The Service Water System shall be demonstrated capable of providing a water supply to the Auxiliary Feedwater System in accordance with the Surveillance Frequency Control Program by verifying that the required spool-piece is on-site. SALEM -UNIT 3/4 7-7 Amendment No. 282 TABLE SECONDARY COOLANT SYSTEM SPECIFIC SAMPLE AND ANALYSIS TYPE OF MEASUREMENT AND ANALYSIS SAIVIPLE AND ANALYSIS FREQUENCY Gross Activity Determination In accordance with the Surveillance Frequency Control Program 2. Isotopic Analysis for a. In accordance with the Surveillance Frequency EQUIVALENT 1-131 Concentration Control Program, whenever the gross activity determination indicates iodine concentrations greater than 10% of the allowable limit. In accordance with the Surveillance Frequency Control Program, whenever the gross activity determination indicates iodine concentrations below 10% of the allowable limit. SALEM -UNIT 3/4 7-9 Amendment No. 282 PLANT 3/4.7.2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION LIMITING CONDITION FOR OPERATION 3.7.2 The temperatures of both the primary and secondary coolants in the steam generators shall be greater than 70°F when the pressure of either coolant in the steam generator is greater than 200 psig. APPLICABILITY:

At all With the requirements of the above specification not

a. Reduce the steam generator pressure of the applicable side to less than or equal to 200 psig within 30 minutes, and b. Perform an engineering evaluation to determine the effect of the overpressurization on the structural integrity of the steam generator.

Determine that the steam generator remains acceptable for continued operation prior to increasing its temperatures above 200°F. 4,7.2 The pressure in each side of the steam generator shall be determined to be less than 200 psig in accordance with the Surveillance Frequency Control Program when the temperature of either the primary or secondary coolant is less than 70°F. SALEM-UNIT 2 3/4 7-11 Amendment No. 282 PLANT 3/4.7.3 COMPONENT COOLING WATER 3.7.3 At least two independent component cooling water loops shall be APPLICABILITY:

MODES 1, 2, 3 and With only one component cooling water loop OPERABLE, restore at least two loops 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 and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

4.7.3 At least two component cooling water loops shall be demonstrated OPERABLE: In accordance with the Surveillance Frequency Control Program by verifying that each valve (manual, power operated or automatic) servicing safety related equipment that is not locked, sealed, or otherwise secured in position, is in its correct position. The time for completion of repairs to number 22 component cooling water heat exchanger shall be extended from 0700 hours0.0081 days 0.194 hours 0.00116 weeks 2.6635e-4 months on November 23, 1982 to 1900 hours0.022 days 0.528 hours 0.00314 weeks 7.2295e-4 months on November 23, 1982. If repairs are not completed by that time, the unit shall be placed in 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 . SALEM -UNIT 3/47-12 Amendment No. 282 PLANT 3/4.7.4 SERVICE WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.4 At least two independent service water loops shall be APPLICABILITY:

MODES 1, 2, 3 and With only one service water loop OPERABLE, restore at least two loops to OPERABLE 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 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 4.7.4 At least two service water loops shall be demonstrated OPERABLE:

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

b. In accordance with the Surveillance Frequency Control Program during shutdown, by verifying that each automatic valve servicing safety related equipment actuates to its correct position on Safeguards Initiation signal. SALEM -UNIT 2 3/47-13 Amendment No. 282 PLANT 3/4.7.5 FLOOD PROTECTION LIMITING CONDITION FOR OPERATION 3.7.5 Flood protection shall be provided for all safety related systems, components and structures when the water level of the Delaware River exceeds 10.5' Mean Sea Level USGS datum, at the service water intake structure.

APPLICABILITY:

At all times. ACTION: a. With the water level at the service water intake structure above elevation 10.5' Mean Sea Level USGS datum, close all watertight doors within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . b. With the water level at the service water intake structure above elevation 11.5' Mean Sea Level USGS datum, be in at least HOT STANDBY within 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 4.7.5 The water level at the service water intake structure shall be determined to be within the limits by: Measurement in accordance with the Surveillance Frequency Control Program when the water level is below elevation 10.5' Mean Sea Level USGS datum, and Measurement in accordance with the Surveillance Frequency Control Program when the water level is equal to or above elevation 10.5' Mean Sea Level USGS datum. SALEM -UNIT 3/47-14 Amendment No. 282 PLANT 4.7.6.1 The control room emergency air conditioning system shall be demonstrated OPERABLE: In accordance with the Surveillance Frequency Control Program by initiating flow through the HEPA filter and charcoal adsorber train(s) and verifying that the train(s) operates with each fan operating for at least 15 minutes. In accordance with the Surveillance Frequency Control Program or prior to return to service (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the system, by: Verifying that the charcoal adsorbers remove;;::

99% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place while operating the ventilation system at a flow rate of 8000 cfm +/- 10%. Verifying that the HEPA filter banks remove;;::

99% of the OOP when they are tested in-place while operating the ventilation system at a flow rate of 8000 cfm +/- 10%. Verifying within 31 days after removal from the CREACS unit, that a laboratory test of a sample of the charcoal adsorber, when obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 2.5% when tested in accordance with ASTM 03803-1989 at a temperature of 30°C and a relative humidity of 95%. After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation by verifying within 31 days after removal from the CREACS unit, that a laboratory analysis of a representative carbon sample, when obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows a methyl iodide penetration less than 2.5% when tested in accordance with ASTM 03803-1989 at a temperature of 30°C and a relative humidity of 95%. In accordance with the Surveillance Frequency Control Program by: Verifying that the pressure drop across the combined HEPA filter and charcoal adsorber bank is S 3.5 inches water gauge while operating the ventilation system at a flow rate of 8000 cfm +/- 10%. Verifying that on a safety injection test signal or control room intake high radiation test signal, the system automatically actuates in the pressurization mode by opening the outside air supply and diverting air flow through the HEPA filter and charcoal adsorber bank. Deleted. SALEM -UNIT 3/47-17 Amendment No. 282 PLANT 4.7.7 The above required Auxiliary Building Ventilation System shall be demonstrated OPERABLE:

a) In accordance with the Surveillance Frequency Control Program by verifying negative pressure in the Auxiliary Building.

b) In accordance with the Surveillance Frequency Control Program by starting each fan, from the control room, each fan operates for at least 15 minutes. c) In accordance with the Surveillance Frequency Control Program by verifying that the system starts following a Safety Injection Test Signal. SALEM -UNIT 2 3/47-19 Amendment No. 282 PLANT 3/4.7.8 SEALED SOURCE 3.7.8 Each sealed source containing radioactive material either in excess of 100 microcuries of beta and/or gamma emitting material or 5 microcuries of alpha emitting material shall be free of greater than or equal to 0.005 microcuries of removable contamination.

APPLICABILITY:

At all times. ACTION: a. With a sealed source having removable contamination in excess of the above limits, immediately withdraw the sealed source from use and: 1. Either decontaminate and repair the sealed source, or 2. Dispose of the sealed source in accordance with Commission Regulations.

b. The provisions of Specification 3.0.3 are not applicable.

4.7.8.1 Test Requirements

-Each sealed source shall be tested for leakage and/or contamination by: The licensee, or b. Other persons specifically authorized by the Commission or an Agreement State. The test method shall have a detection sensitivity of at least 0.005 microcuries per test sample. 4.7.8.2 Test Frequencies

-Each category of sealed sources (excluding startup sources and fission detectors previously subjected to core flux) shall be tested at the frequency described below. Sources in use -In accordance with the Surveillance Frequency Control Program for all sealed sources containing radioactive materials:

SALEM -UNIT 3/4 7-21 Amendment No. 282 PLANT LIMITING CONDITION FOR ACTION: MODES 5 and 6 or during movement of irradiated fuel With one chiller inoperable: Remove the appropriate non-essential heat loads from the Chilled Water System within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and; Restore the chiller to OPERABLE status within 14 days or; Suspend CORE ALTERATIONS and movement of irradiated fuel assemblies. With two chillers inoperable: Remove the appropriate non-essential heat loads from the chilled water system within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and; Align the control room emergency air conditioning system (CREACs) for single filtration operation using the Salem Unit 1 train within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and; Restore at least one chiller to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or; Suspend CORE ALTERATIONS and movement of irradiated fuel assemblies. With one chilled water pump inoperable, restore the chilled water pump to OPERABLE status within 7 days or suspend CORE ALTERATIONS and movement of irradiated fuel assemblies.

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

a. In accordance with the Surveillance Frequency Control Program by verifying that each manual valve in the chilled water system flow path servicing safety related components that is not locked, sealed, or otherwise secured in position, is in its correct position.

b. In accordance with the Surveillance Frequency Control Program, by verifying that each automatic valve actuates to its correct position on a Safeguards Initiation signal. c. In accordance with the Surveillance Frequency Control Program by verifying that each chillers starts and runs. During Modes 5 and 6 and during movement of irradiated fuel assemblies, chilled water components are not considered to be inoperable solely on the basis that the backup emergency power source, diesel generator, is inoperable.

SALEM -UNIT 3/47-29 Amendment No. 282 PLANT SYSTEMS 3/4.7.11 FUEL STORAGE POOL BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.7.11 The fuel storage pool boron concentration shall be;::: 800 ppm APPLICABILITY:

When fuel assemblies are stored in the fuel storage pool and a fuel storage pool verification has not been performed since the last movement of fuel assemblies in the fuel storage pool. ACTION: With fuel storage pool boron concentration not within limit: a. Immediately suspend movement of fuel assemblies in the fuel storage pool and b. Initiate action to: 1. immediately restore fuel storage pool boron concentration to within limit or 2. immediately perform a fuel storage pool verification.

c. The provisions of Specification 3.0.3 is not applicable.

4.7.11 Verify the fuel storage pool boron concentration is within limit in accordance with the Surveillance Frequency Control Program. SALEM -UN IT 2 3/47-30 Amendment No. 282 ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS 4.8.1.1.1 Two physically independent AC. circuits between the offsite transmission network and the onsite Class 1 E distribution system (vital bus system) shall be: Determined OPERABLE in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignments, power availability, and Demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program during shutdown by transferring (manually and automatically) vital bus supply from one 13/4 kv transformer to the other 13/4 kv transformer.

4.8.1.1.2 Each diesel generator shall be demonstrated OPERABLE: In accordance with the Surveillance Frequency Control Program by: Verifying the fuel level in its day tank. Verifying the diesel generator starts from standby conditions*

and achieves;:::

3910 volts and;::: 58.8 Hz in 13 seconds, and subsequently achieves steady state voltage of;::: 3910 and 4400 volts and frequency of 60 +/- 1.2 Hz. Subsequently, verifying the generator is synchronized with voltage maintained;:::

3910 and 4580 volts, gradually loaded to 2340-2600 kw**, and operates at a load of 2340-2600 kw for greater than or equal to 60 minutes. Verifying the diesel generator is aligned to provide standby power to the associated vital bus. In accordance with the Surveillance Frequency Control Program and after each operation of the diesel where the period of operation was greater than or equal to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months by checking for and removing accumulated water from the day tanks. In accordance with the Surveillance Frequency Control Program by verifying the diesel generator starts from standby conditions*

and achieves;:::

3910 volts and ;::: 58.8 Hz in 13 seconds, and subsequently achieves steady state voltage of ;::: 3910 and 4400 volts and frequency of 60 +/- 1.2 Hz. The generator shall be synchronized to its emergency bus with voltage maintained;:::

3910 and 4580 volts, loaded to 2340-2600**

kw in less than or equal to 60 seconds, and operate at a load of 2340-2600 kw for at least 60 minutes. This test, if it is performed so it coincides with the testing required by Surveillance Requirement 4.8.1.1.2.a.2, may also serve to concurrently meet those requirements.

SALEM -UNIT 3/4 8-3 Amendment No. 282 ELECTRICAL POWER In accordance with the Surveillance Frequency Control Program during shutdown by: DELETED Verifying that, on rejection of a load greater than or equal to 820 kw, the voltage and frequency are restored to 2: 3910 and s 4400 volts and 60 +/- 1.2 Hz within 4 seconds, and subsequently achieves a steady state frequency of 2: 58.8 and s 60.5 Hz. Simulating a loss of offsite power by itself, and: Verifying de-energization of the vital bus and load shedding from the vital bus. Verifying the diesel starts on the auto-start signal*, energizes the vital bus with permanently connected loads within 13 seconds, energizes the auto-connected shutdown loads through the load sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with the shutdown loads. After energization of these loads, the steady state voltage and frequency of the vital bus shall be maintained at 2: 3910 and s 4400 volts and 2: 58.8 and s 60.5 Hz during this test. Verifying that on an ESF actuation test signal without loss of offsite power the diesel generator starts on the auto-start signal and operates on standby for greater than or equal to 5 minutes*.

The diesel generator shall achieve 2: 3910 volts and 2: 58.8 Hz in s 13 seconds, and subsequently achieves steady state voltage of 2: 3910 and s 4400 volts and frequency of 2: 58.8 and s 60.5 Hz. Deleted Simulating a loss of offsite power in conjunction with an ESF actuation test signal, and Verifying de-energization of the vital bus and load shedding from the vital bus. Verifying the diesel starts on the auto-start signal*, energizes the vital bus with permanently connected loads within 13 seconds, energizes the auto-connected emergency (accident) loads through the load sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with the emergency loads. After energization of these loads, the steady state voltage and frequency of the vital bus shall be maintained at 2: 3910 and s 4400 volts and 2: 58.8 and s 60.5 Hz during this test. SALEM -UNIT 3/4 8-4 Amendment No. 282 ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTSm(Qqotinued) Verifying that all nonessential automatic diesel generator trips (i.e., other than engine overspeed, lube oil pressure low, 4 KV Bus differential and generator differential) are automatically bypassed upon loss of voltage on the vital bus concurrent with a safety injection actuation signal. Deleted Verifying that the auto-connected loads to each diesel generator do not exceed the two hour rating of 2860 kw. Verifying that with the diesel generator operating in a test mode (connected to its bus), a simulated safety injection signal overrides the test mode by (1) returning the diesel generator to standby operation and (2) automatically energizing the emergency loads with offsite power. In accordance with the Surveillance Frequency Control Program or after any modifications which could affect diesel generator interdependence by starting all diesel generators simultaneously*, during shutdown, and verifying that all diesel generators accelerate to at least 58.8 Hz in less than or equal to 13 seconds. In accordance with the Surveillance Frequency Control Program, the following test shall be performed within 5 minutes of diesel shutdown after the diesel has operated for at least two hours at 2340-2600 kw**: Verifying the diesel generator starts and achieves 3910 volts and 58.8 Hz in $; 13 seconds, and subsequently achieves steady state voltage of 3910 and $; 4400 volts and frequency of 60 +/- 1.2 Hz. In accordance with the Surveillance Frequency Control Program verifying the diesel generator operates for at least 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s*. During the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of this test, the diesel generators shall be loaded to 2760-2860 Kw**. During the remaining 22 hours2.546296e-4 days 0.00611 hours 3.637566e-5 weeks 8.371e-6 months of this test, the diesel generator shall be loaded to 2600 Kw**. The steady state voltage and frequency shall be maintained at 3910 and $; 4580 volts and 60 +/- 1.2 Hz during this test. 4.8.1.1.3 The diesel fuel oil storage and transfer system shall be demonstrated OPERABLE: In accordance with the Surveillance Frequency Control Program by: Verifying the level in each of the above required fuel storage tanks. Verifying that both fuel transfer pumps can be started and transfer fuel from the fuel storage tanks to the day tanks. SALEM -UNIT 3/4 8-5 Amendment No, 282 ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) In accordance with the Surveillance Frequency Control Program by verifying that a sample of diesel fuel from each of the above required fuel storage tanks is within the acceptable limits specified in Table 1 of ASTM 0975-77 when checked for viscosity, water and sediment.

4.8.1.1.4 Reports -NOT USED Surveillance testing may be conducted in accordance with the manufacturer's recommendations regarding engine prelube, warm-up and loading (unless loading times are specified in the individual Surveillance Requirements). This band is meant as guidance to preclude routine exceedances of the diesel generator manufacturer's design ratings. Loads in excess of this band for special testing or momentary variations due to changing bus loads shall not invalidate the test. SALEM -UNIT 3/4 8-6 Amendment No. 282 ELECTRICAL POWER SYSTEMS 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS AC. DISTRIBUTION

-OPERATING LIMITING CONDITION FOR OPERATION 3.8.2.1 The following AC. electrical busses shall be OPERABLE and energized from sources of power other than the diesel generators:

4 kvolt Vital Bus # 2A 4 kvolt Vital Bus # 2B 4 kvolt Vital Bus # 2C 460 volt Vital Bus # 2A and associated control centers 460 volt Vital Bus # 2B and associated control centers 460 volt Vital Bus # 2C and associated control centers 230 volt Vital Bus # 2A and associated control centers 230 volt Vital Bus # 2B and associated control centers 230 volt Vital Bus # 2C and associated control centers 115 volt Vital Instrument Bus # 2A and Inverter

115 volt Vital Instrument Bus # 2B and Inverter

115 volt Vital Instrument Bus # 2C and Inverter

115 volt Vital Instrument Bus # 2D and Inverter

APPLICABILITY:

MODES 1,2, 3 and 4. ACTION: With less than the above complement of AC. busses OPERABLE or energized, restore the inoperable busses to OPERABLE and energized status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 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 and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . With one inverter inoperable, energize the associated AC. Vital Bus within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; restore the inoperable 2A, 2B, or 2C inverter to OPERABLE and energized status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 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 and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ; restore the inoperable 2D inverter to OPERABLE and energized 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 and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . 4.8.2.1 The specified AC. busses and inverters shall be determined OPERABLE and energized from AC. sources other than the diesel generators in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignment and indicated voltage on the busses. An inverter may be disconnected from its D.C. source for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for the purpose of performing an equalizing charge on its associated battery bank provided (1) its vital bus is OPERABLE and energized.

and (2) the vital busses associated with the other battery banks are OPERABLE and energized.

SALEM -UNIT 3/4 8-8 Amendment No. 282 ELECTRICAL POWER SYSTEMS AC. DISTRIBUTION

-SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.2.2 As a minimum, two AC. electrical bus trains shall be OPERABLE and energized from sources of power other than a diesel generator but aligned to an OPERABLE diesel generator with each train consisting of: 1 4 kvolt Vital Bus 1 460 volt Vital Bus and associated control centers 1 230 volt Vital Bus and associated control centers 1 115 volt Instrument Bus energized from its respective inverter connected to its respective D.C. bus train.

MODES 5 and 6. During movement of irradiated fuel assemblies.

ACTION: With less than the above complement of AC. busses and inverters OPERABLE and energized, immediately declare the affected required features inoperable, or suspend all operations involving CORE ALTERATIONS, positive reactivity changes, and movement of irradiated fuel assemblies until the minimum required AC. electrical power sources are restored to OPERABLE status. 4.8.2.2 The specified AC. busses and inverters shall be determined OPERABLE and energized from AC. sources other than the diesel generators in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignment and indicated voltage on the busses. SALEM -UNIT 3/48-9 Amendment No. 282 ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS 4.8.2.3.1 Each D.C. bus train shall be determined OPERABLE and energized in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignment and voltage on the bus. 4.8.2.3.2 Each required 125-volt battery and charger shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by verifying that: 1. The parameters in Table 4.8.2.3-1 meet Category A limits. 2. The overall battery voltage is greater than or equal to 125 volts on float charge. b. In accordance with the Surveillance Frequency Control Program and once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a battery discharge

< 110 V and once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a battery overcharge>

150 V by verifying that the parameters in Table 4.8.2.3-1 meet the Category B limits. c. In accordance with the Surveillance Frequency Control Program by verifying that: 1. There is no visible corrosion at terminals or connectors or the connection resistance is: :5 150 micro ohms for inter-cell connections, :5 350 micro ohms for inter-rack connections, :5 350 micro ohms for inter-tier connections, :5 70 micro ohms for field cable terminal connections, and :5 2500 micro ohms for the total battery connection resistance which includes all inter-cell connections (including bus bars), all inter-rack connections (including cable resistance), all inter-tier connections (including cable resistance), and all field terminal connections at the battery. The average electrolyte temperature of the representative cells is above 65°F. In accordance with the Surveillance Frequency Control Program by verifying that: The cells, cell plates and battery racks show no visual indication of physical damage or abnormal deterioration. Remove visible terminal corrosion and verify cell-to-cell and terminal connections are coated with anti-corrosion material.

SALEM -UNIT 3/48-11 Amendment No. 282 ELECTRICAL POWER SYSTEMS SURVEILLANCE REQU I REMENISm(Qontinued) The connection resistance is: ;5;150 micro ohms for inter-cell connections, ;5;350 micro ohms for inter-rack connections, ;5;350 micro ohms for inter-tier connections, 5'070 micro ohms for field cable terminal connections,and

5;2500 micro ohms for the total battery connection resistance which includes all inter-cell connections (including bus bars), all inter-rack connections (including cable resistance) all inter-tier connections (including cable resistance) and all field terminal connections at the battery. In accordance with the Surveillance Frequency Control Program by verifying that the battery charger will supply at least 170 amperes at 125 volts for at least 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . In accordance with the Surveillance Frequency Control Program, during shutdown*, by verifying that the battery capacity is adequate to supply and maintain in OPERABLE status all of the actual or simulated emergency loads for the design duty cycle when the battery is subjected to a battery service test. In accordance with the Surveillance Frequency Control Program, during shutdown*, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test. Satisfactory completion of this performance discharge test shall also satisfy the reqUirements of Specification 4.8.2.3.2.f if the performance discharge test is conducted during a shutdown*

where that test and the battery service test would both be required. At least once per 12 months, during shutdown, if the battery shows signs of degradation OR has reached 85% of the service life with a capacity less than 100% of manufacturers rating, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test. Degradation is indicated when the battery capacity drops more than 10% of rated capacity from its capacity on the previous performance test, or is below 90% of the manufacturers rating. At least once per 24 months, during shutdown, if the battery has reached 85% of the service life with capacity greater than or equal to 100% of manufacturers rating, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test. This battery surveillance may be performed, as required, associated with a one-time replacement of station battery 2C when the unit is not shutdown.

This testing shall be done when the battery is disconnected from the 2C DC bus. This testing must be completed by February 1, 2011. SALEM -UNIT 3/48-11a Amendment No. 282 ELECTRICAL POWER SYSTEMS 125-VOLT D.C. DISTRIBUTION

-SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.2.4 As a minimum, the following D.C. electrical equipment and bus shall be energized and OPERABLE: 2 125-volt D.C. busses, and 2 125-volt batteries, each with at least one full capacity charger, associated with each of the above D.C. busses.

MODES 5 and 6. During movement of irradiated fuel assemblies.

ACTION: With less than the above complement of D.C. equipment and busses OPERABLE, immediately declare the affected required features inoperable, or suspend all operations involving CORE ALTERATIONS, positive reactivity changes, and movement of irradiated fuel assemblies until the minimum required 125 Volt D.C. electrical power sources are restored to OPERABLE status. SURVEILLANCE REQUIREMENTS 4.8.2.4.1 The above required 125-volt D.C. busses shall be determined OPERABLE and energized in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignment and indicated power availability.

4.8.2.4.2 The above required 125-volt batteries and chargers shall be demonstrated OPERABLE per Surveillance Requirement 4.8.2.3.2.

SALEM -UNIT 3/48-12 Amendment No. 282 ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS 4.8.2.5.1 Each D.C. bus train shall be determined OPERABLE and energized in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignment and power availability.

4.8.2.5.2 Each 28-volt battery and above required charger shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by verifying that: 1. The parameters in Table 4.8.2.5-1 meet Category A limits. 2. The overall battery voltage is greater than or equal to 27 volts on float charge. b. In accordance with the Surveillance Frequency Control Program and once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a battery discharge

< 25.7 V and once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a battery overcharge>

35 V by verifying that the parameters in Table 4.8.2.5-1 meet the Category B limits. c. In accordance with the Surveillance Frequency Control Program by verifying that: 1. There is no visible corrosion at terminals or connectors or the connection resistance is: :5 50 micro ohms for inter-cell connections, :5 200 micro ohms for inter-tier connections, :5 70 micro ohms for field cable terminal connections, and :5 500 micro ohms for the total battery connection resistance which includes all inter-cell connections (including bus bars), all inter-tier connections (including cable resistance) and all field terminal connections at the battery. 2. The average electrolyte temperature of the representative cells is 65°F. SALEM -UNIT 2 3/48-14 Amendment No. 282 ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) In accordance with the Surveillance Frequency Control Program by verifying that: The cells, cell plates and battery racks show no visual indication of physical damage or abnormal deterioration. Remove visible terminal corrosion and verify cell-to-cell and terminal connections are coated with anti-corrosion material. The connection resistance is: s 50 micro ohms for inter-cell connections, s 200 micro ohms for inter-tier connections, s 70 micro ohms for field cable terminal connections, and s 500 micro ohms for the total battery connection resistance which includes all inter-cell connections (including bus bars), all inter-tier connections (including cable resistance) and all field terminal connections at the battery. In accordance with the Surveillance Frequency Control Program by verifying that the battery charger will supply;::

150 amperes at ;:: 28 volts for;:: 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . In accordance with the Surveillance Frequency Control Program, during shutdown, by verifying that the battery capacity is adequate to supply and maintain in OPERABLE status all of the actual or simulated emergency loads for the design duty cycle when the battery is subjected to a battery service test. In accordance with the Surveillance Frequency Control Program, during shutdown, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test. Satisfactory completion of this performance discharge test shall also satisfy the requirements of Specification 4.8.2.5.2.f if the performance discharge test is conducted during a shutdown where that test and the battery service test would both be required. At least once per 12 months, during shutdown, if the battery shows signs of degradation OR has reached 85% of the service life with a capacity less than 100% of manufacturers rating, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test. Degradation is indicated when the battery capacity drops more than 10% of rated capacity from its capacity on the previous performance test, or is below 90% of the manufacturer's rating. At least once per 24 months, during shutdown, if the battery has reached 85% of the service life with capacity greater than or equal to 100% of manufacturers rating, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test. SALEM -UNIT 3/48-14a Amendment No. 282 ELECTRICAL POWER SYSTEMS 28-VOL T D.C. DISTRIBUTION

-SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.2.6 As a minimum, the following D.C. electrical equipment and bus shall be energized and OPERABLE: 1 28-volt D.C. bus, and 1 28-volt battery and at least one full capacity charger associated with the above D.C. bus.

MODES 5 and 6. During movement of irradiated fuel assemblies.

ACTION: With less than the above complement of D.C. equipment and busses OPERABLE, immediately declare the affected required features inoperable, or suspend all operations involving CORE ALTERATIONS, positive reactivity changes, and movement or irradiated fuel assemblies until the minimum required 28 Volt D.C. electrical power sources are restored to OPERABLE status. SURVEILLANCE REQUIREMENTS 4.8.2.6.1 The above required 28-volt D.C. bus shall be determined OPERABLE and energized in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignment and voltage on the bus. 4.8.2.6.2 The above required 28-volt batteries and charger shall be demonstrated OPERABLE per Surveillance Requirement 4.8.2.5.2.

SALEM -UNIT 3/48-15 Amendment No. 282 ELECTRICAL POWER 3/4.8.3 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES LIMITING CONDITION FOR OPERATION 3.8.3.1 All containment penetration conductor overcurrent protective devices required provide thermal protection of penetrations shall be APPLICABILITY:

MODES 1, 2, 3 and With one or more of the containment penetration conductor overcurrent protective inoperable:

a. Restore the protective device(s) to OPERABLE status or de-energize the circuit(s) by tripping either the primary or backup protective device, or racking out or removing the primary or backup device within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , declare the affected system or component inoperable, and verify the primary or backup protective device to be tripped, or the primary or backup device racked out or removed at least once per 7 days thereafter; or b. Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . SURVEILLANCE REQUIREMENTS 4.8.3.1 All required containment penetration conductor overcurrent protective devices shall be demonstrated OPERABLE: In accordance with the Surveillance Frequency Control Program: For 4.16 KV reactor coolant pump circuits, by performance of: A CHANNEL CALIBRATION of the associated protective relays, and An integrated system functional test which includes simulated automatic actuation of the system and verifying that each relay and associated circuit breakers and control circuits function as designed.

SALEM -UNIT 3/48-16 Amendment No. 282 ELECTRICAL POWER By verifying the OPERABILITY of the required molded case and lower voltage circuit breakers, by selecting and functionally testing a representative sample of at least 10% of all the circuit breakers of that type. Circuit breakers selected for functional testing shall be selected on a rotating basis. The functional test shall consist of injecting a current input at the specified setpoint to each selected circuit breaker and verifying that each circuit breaker functions as designed.

Circuit breakers found inoperable during functional testing shall be restored to OPERABLE status prior to resuming operation.

For each circuit breaker found inoperable during the functional tests, an additional representative sample of at least 10% of all the circuit breakers of the inoperable type shall also be functionally tested until no more failures are found or all circuit breakers of that type have been functionally tested. In accordance with the Surveillance Frequency Control Program by subjecting each circuit breaker to an inspection and preventive maintenance in accordance with procedures prepared in conjunction with its manufacturer's recommendations.

SALEM -UNIT 3/48-17 Amendment No. 282 3/4.9 REFUELING OPERATIONS BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.1 The boron concentration of the Reactor Coolant System, the fuel storage pool, the refueling canal, and the refueling cavity shall be maintained within the limit specified in the CORE OPERATING LIMITS REPORT (COLR).

MODE 6 (Only applicable to the refueling canal, the fuel storage pool and refueling cavity when connected to the Reactor Coolant System) ACTION: With the requirements of the above specification not satisfied, immediately

a. Suspend CORE ALTERATIONS and b. Suspend positive reactivity additions and c. Initiate action to restore boron concentration to within limit specified in the COLR. d. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.1 Verify the boron concentration is within the limit of the COLR in accordance with the Surveillance Frequency Control Program. SALEM -UNIT 3/49-1 Amendment No. 282 REFUELING OPERATIONS 3/4.9.2 INSTRUMENTATION LIMITING CONDITION FOBmQPERATION 3.9.2 As a minimum, two source range neutron flux monitors shall be operating, each with continuous visual indication in the control room and one with audible indication in the containment and control room. APPLICABILITY:

MODE 6. ACTION: a. With one of the above required monitors inoperable, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes. b. With both of the required monitors inoperable, determine the boron concentration of the reactor coolant system at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . c. The provisions of Specification 3.0.3 are not applicable.

4.9.2 Each source range neutron flux monitor shall be demonstrated OPERABLE by performance of: a. A CHANNEL CHECK in accordance with the Surveillance Frequency Control Program, and b. A CHANNEL FUNCTIONAL TEST in accordance with the Surveillance Frequency Control Program. SALEM -UNIT 2 3/4 9-2 Amendment No. 282 REFUELING OPERATIONS CONTAINMENT BUILDING PENETRATIONS 3.9.4 The containment building penetrations shall be in the following status: a. The equipment hatch inside door is capable of being closed and held in place by a minimum of four bolts, or an equivalent closure device installed and capable of being closed, b. A minimum of one door in each airlock is capable of being closed c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either: 1 closed by a manual or automatic isolation valve, blind flange, or equivalent, or 2. capable of being closed by the Containment Purge and Pressure-Vacuum Relief Isolation System. Note: Penetration flow path(s) providing direct access from the containment atmosphere to the outside atmosphere may be un isolated under administrative controls.

APPLICABILITY:

During movement of irradiated fuel within the containment.

ACTION: With the requirements of the above specification not satisfied, immediately suspend all operations involving movement of irradiated fuel in the containment building.

The provisions of Specification 3.0.3 are not applicable.

4.9.4.1 Each of the above required containment building penetrations shall be determined to be either in its required condition or capable of being closed by a manual or automatic containment isolation valve in accordance with the Surveillance Frequency Control Program. 4.9.4.2 Once per refueling prior to the start of movement of irradiated fuel assemblies within the containment building, verify the capability to close, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , the equipment hatch inside door or an equivalent closure device. Applicable only when the equipment hatch is open during movement of irradiated fuel in the containment building.

4.9.4.3 Verify, in accordance with the Surveillance Frequency Control Program, each required containment purge isolation valve actuates to the isolation position on a manual actuation signal. SALEM -UNIT 2 3/4 9-4 Amendment No. 282 REFUELING 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION ALL WATER LEVELS LIMITING CONDITION FOR OPERATION 3.9.8.1 At least one residual heat removal loop shall be in APPLICABILITY:

MODE a. With less than one residual heat removal loop in operation, except as provided in b. below, suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System. Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . b. The residual heat removal loop may be removed from operation for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period during the performance of CORE ALTERATIONS in the vicinity of the reactor pressure vessel hot legs. c. The provisions of Specification 3.0.3 are not applicable.

4.9.8.1 In accordance with the Surveillance Frequency Control Program one RHR loop shall be verified in operation and circulating coolant at a flow rate of: a. greater than or equal to 1000 gpm, and b. sufficient to maintain the RCS temperature at less than or equal to 140°F. SALEM -UNIT 2 3/4 9-8 Amendment No. 282 REFUELING 3/4.9.10 WATER LEVEL -REACTOR VESSEL LIMITING CONDITION FOR OPERATION 3.9.10 At least 23 feet of water shall be maintained over the top of the reactor pressure vessel flange.

During movement of fuel assemblies or control rods within the reactor pressure vessel when either the fuel assemblies being moved or the fuel assernblies seated within the reactor pressure vessel are irradiated.

ACTION: With the requirements of the above specification not satisfied, suspend all operations involving movement of fuel assemblies or control rods within the pressure vessel. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.10 The water level shall be determined to be at least its minimum required depth within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to the start of and in accordance with the Surveillance Frequency Control Program thereafter during movements of fuel assemblies or control rods. SALEM -UNIT 3/49-11 Amendment No. 282 REFUELING 3/4,9,11 STORAGE POOL WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.11 At least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks, APPLICABILITY:

Whenever irradiated fuel assemblies are in the storage pool. ACTION: With the requirements of the specification not satisfied, suspend all movement of fuel assemblies and crane operations with loads in the fuel storage areas and restore water level to within its limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The provisions of Specification 3.0.3 are not applicable, SURVEILLANCE REQUIREMENTS 4.9.11 The water level in the storage pool shall be determined to be at least its minimum required depth in accordance with the Surveillance Frequency Control Program when irradiated fuel assemblies are in the fuel storage pool. SALEM -UNIT 2 3/49-12 Amendment No. 282 REFUELING 3/4.9.12 FUEL HANDLING AREA VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.9.12 The Fuel Handling Area Ventilation System shall be OPERABLE with: Two exhaust fans and one supply fan OPERABLE and operating, and b. Capable of maintaining slightly negative pressure in the Fuel Handling Building.

APPLICABILITY:

During movement of irradiated fuel within the Fuel Handling Building ACTION: With no Fuel Handling Area Ventilation System OPERABLE, suspend all operations involving movement of fuel within the storage pool until the Fuel Handling Area Ventilation System is restored to OPERABLE status. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.12 The above required ventilation system shall be demonstrated OPERABLE: In accordance with the Surveillance Frequency Control Program by verifying that the Fuel Handling Building is maintained at a slightly negative pressure with respect to atmospheric pressure. In accordance with the Surveillance Frequency Control Program by verifying both exhaust fans and one supply fan start and operate for at least 15 minutes, if not operating already. In accordance with the Surveillance Frequency Control Program by verifying a system flowrate of 19,490 cfm +/- 10% during system operation.

SALEM -UNIT 3/49-13 Amendment No. 282 3/4.10 SPECIAL TEST EXCEPTIONS SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of control rod worth and shutdown margin provided the reactivity equivalent to at least the highest estimated control rod worth is available for trip insertion from OPERABLE control rod(s). APPLICABILITY:

MODE 2. ACTION: a. With any full length control rod not fully inserted and with less than the above reactivity equivalent available for trip insertion, immediately initiate and continue boration at 33 gpm of a solution containing 6,560 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.

b. With all full length control rods inserted and the reactor subcritical by less than the above reactivity equivalent, immediately initiate and continue boration at 33 gpm of a solution containing 6,560 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.

SURVEILLANCE REQUIREMENTS 4.10.1.1 The position of each full length rod either partially or FULLY WITHDRAWN shall be determined in accordance with the Surveillance Frequency Control Program. 4.10.1.2 Each full length rod not fully inserted shall be demonstrated capable of full insertion when tripped from at least the 50% withdrawn position within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to reducing the SHUTDOWN MARGIN to less than the limits of Specification 3.1.1.1. SALEM -UNIT 2 3/4 10-1 Amendment No. 282 SPECIAL TEST GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION 3.10.2 The group height, insertion and power distribution limits of Specifications 3.1.3.1, 3.1.3.4, 3.1.3.5, 3.2.1, and 3.2.4 may be suspended during the performance of PHYSICS TESTS provided: The THERMAL POWER is maintained less than or equal to 85%/ of RATED THERMAL POWER, and The limits of Specifications 3.2.2 and 3.2.3 are maintained and determined at the frequencies specified in Specification 4.10.2.2 below. APPLICABILITY:

MODE 1. With any of the limits of Specifications 3.2.2 or 3.2.3 being exceeded while the requirements Specifications 3.1.3.1, 3.1.3.4, 3.1.3.5, 3.2.1 and 3.2.4 are suspended, Reduce THERMAL POWER sufficient to satisfy the ACTION requirements of Specifications 3.2.2 and 3.2.3, or Be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . SURVEILLANCE REQUIREMENTS 4.10.2.1 The THERMAL POWER shall be determined to be less than or equal to 85% of RATED THERMAL POWER in accordance with the Surveillance Frequency Control Program during PHYSICS TESTS. 4.10.2.2 The below listed surveillance requirements shall be performed in accordance with the Surveillance Frequency Control Program during PHYSICS TESTS: Surveillances 4.2.2.2 and 4.2.2.3. Surveillances 4.2.3.1 and 4.2.3.2. # See page 3/4 10-3 SALEM -UNIT 2 3/4 10-2 Amendment No. 282 SPECIAL TEST PHYSICS TESTS LIMITING CONDITION FOR OPERATION 3.10.3 The limitations of Specifications 3.1.1.3, 3.1.1.4, 3.1.3.1, 3.1.3.4 and 3.1.3.5 may be suspended during the performance of PHYSICS TESTS provided: The THERMAL POWER does not exceed 5% of RATED THERMAL POWER, The reactor trip setpoints on the OPERABLE Intermediate and Power Range Nuclear Channels are set at less than or equal to 25% of RATED THERMAL POWER, and The Reactor Coolant System lowest operating loop temperature (Tavg) is greater than or equal to 531 OF. APPLICABILITY:

MODE 2. ACTION: With the THERMAL POWER greater than 5% of RATED THERMAL POWER, immediately open the reactor trip breakers. With a Reactor Coolant System operating loop temperature (Tavg) less than 531°F, restore Tavg to within its limit within 15 minutes or be in at least HOT STANDBY within the next 15 minutes. SURVEILLANCE REQUIREMENTS 4.10.3.1 The THERMAL POWER shall be determined to be less than or equal to 5% of RATED THERMAL POWER in accordance with the Surveillance Frequency Control Program during PHYSICS TESTS. 4.10.3.2 Each Intermediate and Power Range Channel shall be subjected to a CHANNEL FUNCTIONAL TEST prior to initiating PHYSICS TESTS. 4.10.3.3 The Reactor Coolant System temperature (Tavg) shall be determined to be greater than or equal to 531 OF in accordance with the Surveillance Frequency Control Program during PHYSICS TESTS. SALEM -UNIT 3/4 10-4 Amendment No. 282 SPECIAL TEST NO FLOW TESTS LIMITING CONDITION FOR OPERATION 3.10.4 The limitations of Specification 3.4.1.1 may be suspended during the performance of startup and PHYSICS TESTS, provided:

a. The THERMAL POWER does not the P-7 Interlock Setpoint, and The Reactor Trip Setpoints on the OPERABLE Intermediate and Power Range Channels are set less than or equal to 25% of RATED THERMAL POWER APPLICABILITY:

During operation below the P-7 Interlock Setpoint. With the THERMAL POWER greater than the P-7 Interlock Setpoint, immediately open reactor trip 4.10.4.1 The THERMAL POWER shall be determined to be less than P-7 Interlock Setpoint in accordance with the Surveillance Frequency Control Program during startup and PHYSICS TESTS. 4.10.4.2 Each Intermediate, Power Range Channel and P-7 Interlock shall be subjected to a CHANNEL FUNCTIONAL TEST prior to initiating startup or PHYSICS TESTS. SALEM -UNIT 3/4 10-5 Amendment No. 282 RADIOACTIVE EFFLUENTS LIQUID HOLDUP TANKS* LIMITING CONDITION FOR OPERATION 3.11.1.4 The quantity of radioactive material contained in each outdoor temporary tank shall be limited to less than or equal to 10 curies, excluding tritium and dissolved or entrained noble gases. APPLICABILITY:

At all times. ACTION: With the quantity of radioactive material in any of the above listed tanks exceeding the above limit, immediately suspend all additions of radioactive material to the tank and within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limit. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.1.4 The quantity of radioactive material contained in each outdoor temporary tank shall be determined to be within the above limit by analyzing a representative sample of the tank's contents in accordance with the Surveillance Frequency Control Program when radioactive materials are being added to the tank. Tanks included in this Specification are those outdoor temporary tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system. SALEM -UNIT 3/4 11-7 Amendment No. 282 RADIOACTIVE EXPLOSIVE GAS 3.11.2.5 The concentration of oxygen in the waste gas holdup system shall be limited to less than or equal to 2% by volume. APPLICABILITY:

At all times.

ACTION: With the concentration of oxygen in the waste gas holdup system greater than 2% by volume but less than or equal 4% by volume, reduce the oxygen concentration to the above limits within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . With the concentration of oxygen in the waste gas holdup system greater than 4% by volume immediately suspend all additions of waste gases to the system and reduce the concentration of oxygen to less than or equal to 2% by volume without delay. The provisions of Specification 3.0.3 are not applicable.

4.11.2.5 The concentration of oxygen in the waste gas holdup system shall be determined to be within the above limits by continuously**

monitoring the waste gases in the waste gas holdup system with the oxygen monitor. If hydrogen is not measured, the concentration of hydrogen shall be assumed to exceed 4% by volume. Not applicable to portions of the Waste Gas System removed from service for maintenance provided that, the portions removed for maintenance are isolated, and purged of hydrogen to less than 4% by volume. If the oxygen monitoring instrumentation is inoperable, operation of the waste gas holdup system may continue provided grab samples are collected in accordance with the Surveillance Frequency Control Program and analyzed within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . SALEM -UNIT 3/411-15 Amendment No. 282 ADMINISTRATIVE 6.8.4.1 Surveillance Frequency Control Program This program provides controls for Surveillance Frequencies.

The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met. The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1. The provisions of Surveillance Requirements 4.0.2 and 4.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program. SALEM -UNIT 6-19g Amendment No. 282 UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555*0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 299 AND 282 TO FACILITY OPERATING LICENSE NOS. DPR-70 AND DPR-75 PSEG NUCLEAR, LLC EXELON GENERATION COMPANY, LLC SALEM NUCLEAR GENERATING STATION, UNIT NOS. 1 AND 2 DOCKET NOS. 50-272 AND 50-311

1.0 INTRODUCTION

By letter dated March 23, 2010, as supplemented by letters dated November 19, 2010, January 31,2011, and February 23,2011 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML 100910154, ML 103230482, ML 110320094, and ML 110540567 respectively), PSEG Nuclear LLC (PSEG or the licensee) submitted a request for changes to the Salem Nuclear Generating Station, Unit Nos. 1 and 2, (Salem) Technical Specifications (TSs). The supplements dated November 19, 2010, January 31, 2011, and February 23, 2011, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission (NRC or the Commission) staff's original proposed no significant hazards consideration determination as published in the Federal Register (FR) on June 15, 2010 (75 FR 33843). The requested change is the adoption of NRC-approved TS Task Force (TSTF) change TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control -RITSTF [Risk*lnformed TSTF] Initiative 5b" (Reference 1). When implemented, TSTF-425 relocates most periodic frequencies of TS surveillances to a licensee-controlled program, the Surveillance Frequency Control Program (SFCP), and provides requirements for the new program in the Administrative Controls section of the TSs. All surveillance frequencies can be relocated except: Frequencies that reference other approved programs for the specific interval (such as the Inservice Testing Program or the Primary Containment Leakage Rate Testing Program); Frequencies that are purely event-driven (e.g., "each time the control rod is withdrawn to the 'full out' position"); Frequencies that are event-driven, but have a time component for performing the surveillance on a one-time basis once the event occurs (e.g., "within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after thermal power reaching 2: 95% RTP [rated thermal power]");

and Enclosure

-2 Frequencies that are related to specific conditions (e.g., battery degradation, age and capacity) or conditions for the performance of a surveillance requirement (SR) (e.g., "drywell to suppression chamber differential pressure decrease").

A new program would be added to the Administrative Controls in TS Section 6.0 as TS 6.8.4.1. The new program is called the SFCP and describes the requirements for the program to control changes to the relocated surveillance frequencies.

The proposed licensee changes to the Administrative Controls of the TSs to incorporate the SFCP include a specific reference to Nuclear Energy Institute (NEI) 04-10, "Risk-Informed Technical Specifications Initiative 58, Informed Method for Control of Surveillance Frequencies," Revision 1 (Reference 2), as the basis for making any changes to the surveillance frequencies once they are relocated out of the TSs. In a letter dated September 19,2007 (Reference 9), the NRC staff approved NEI 04-10, Revision 1, as acceptable for referencing by licensees proposing to amend their TSs to establish a SFCP, to the extent specified and under the limitations delineated in NEI 04-10, and in the NRC staff's safety evaluation (SE) providing the basis for its acceptance of NEI 04-10. The NRC staff issued a Notice of Availability for TSTF-425, Revision 3, in the Federal Register on July 6,2009 (74 FR 31996). The notice included a model SE. In its application dated March 23,2010, the licensee stated that "PSEG has concluded that the justifications presented in the TSTF proposal and the safety evaluation prepared by the NRC staff are applicable to Salem Units 1 and 2." The SE that follows is based, in large part, on the model SE for TSTF-425.

2.0 REGULATORY EVALUATION

In the "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors," published in the Federal Register on July 22, 1993 (58 FR 39132), the NRC addressed the use of Probabilistic Safety Analysis (PSA, currently referred to as Probabilistic Risk Assessment (PRA>> in determining the content of the TSs. On page 39135 of this FR publication, the Commission stated, in part, that: The Commission believes that it would be inappropriate at this time to allow requirements which meet one or more of the first three criteria [in Title 10 of the Code of Federal Regulations (10 CFR), Section 50.36(c)(2)(ii)]

to be deleted from Technical Specifications based solely on PSA (Criterion 4). However, if the results of PSA indicate that Technical Specifications can be relaxed or removed, a deterministic review will be performed

..... The Commission Policy in this regard is consistent with its Policy Statement on "Safety Goals for the Operation of Nuclear Power Plants," 51 FR 30028, published on August 21, 1986. The Policy Statement on Safety Goals states in part, "* *

probabilistic results should also be reasonably balanced and supported through use of deterministic arguments.

In this way, judgments can be made * *

about the degree of confidence to be given these [probabilistic]

estimates and assumptions.

This is a key part of the process of determining the degree of regulatory conservatism that may be warranted for particular decisions.

-3 This defense-in-depth approach is expected to continue to ensure the protection of public health and safety." .... The Commission will continue to use PSA. consistent with its policy on Safety Goals. as a tool in evaluating specific line-item improvements to Technical Specifications.

new requirements, and industry proposals for risk-based Technical Specification changes. Approximately 2 years later, the NRC provided additional detail concerning the use of PRA in the "Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory Activities; Final Policy Statement," published in the Federal Register on August 16, 1995 (60 FR 42622). On page 42627 of this FR publication, the Commission stated. in part, that: PRA addresses a broad spectrum of initiating events by assessing the event frequency.

Mitigating system reliability is then assessed.

including the potential for multiple and common-cause failures.

The treatment therefore goes beyond the single failure requirements in the deterministic approach.

The probabilistic approach to regulation is, therefore, considered an extension and enhancement of traditional regulation by considering risk in a more coherent and complete manner. On pages 42628 and 42629 of this FR publication.

the Commission provided its policy on the use of PRA which states: Although PRA methods and information have thus far been used successfully in nuclear regulatory activities.

there have been concerns that PRA methods are not consistently applied throughout the agency. that sufficient agency PRAlstatistics expertise is not available.

and that the Commission is not deriving full benefit from the large agency and industry investment in the developed risk assessment methods. Therefore, the Commission believes that an overall policy on the use of PRA in nuclear regulatory activities should be established so that the many potential applications of PRA can be implemented in a consistent and predictable manner that promotes regulatory stability and efficiency.

This policy statement sets forth the Commission's intention to encourage the use of PRA and to expand the scope of pRA applications in all nuclear regulatory matters to the extent supported by the state-of-the-art in terms of methods and data. Implementation of the policy statement will improve the regulatory process in three areas: Foremost, through safety decision making enhanced by the use of pRA insights; through more efficient use of agency resources; and through a reduction in unnecessary burdens on licensees.

Therefore, the Commission adopts the following policy statement regarding the expanded NRC use of PRA: (1) The use of PRA technology should be increased in all regulatory matters to the extent supported by the state-of-the-art in PRA methods and data and in a manner that complements the NRC's deterministic approach and supports the NRC's traditional defense-in-depth philosophy.

(2) PRA and associated analyses (e.g., sensitivity studies, uncertainty analyses, and importance measures) should be used in regulatory matters, where practical within the bounds of the state-of-the-art, to reduce unnecessary conservatism associated with current regulatory requirements, regulatory guides, license commitments, and staff practices.

Where appropriate, PRA should be used to support the proposal for additional regulatory requirements in accordance with 10 CFR 50.109 (8ackfit Rule). Appropriate procedures for including PRA in the process for changing regulatory requirements should be developed and followed.

It is, of course, understood that the intent of this policy is that existing rules and regulations shall be complied with unless these rules and regulations are revised. (3) PRA evaluations in support of regulatory decisions should be as realistic as practicable and appropriate supporting data should be publicly available for review. (4) The Commission's safety goals for nuclear power plants and subsidiary numerical objectives are to be used with appropriate consideration of uncertainties in making regulatory judgments on the need for proposing and backfitting new generic requirements on nuclear power plant licensees.

The Commission's regulatory requirements related to the content of the TSs are set forth in 10 CFR 50.36, 'Technical specifications." This regulation requires that the TSs include items in the following five specific categories:

(1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation; (3) SRs; (4) design features; and (5) administrative controls.

The regulation does not specify the particular requirements to be included in a plant's TSs. As stated in 10 CFR 50.36(c)(3), "Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met." To meet this requirement, the SR must specify an adequate test, calibration, or inspection and an appropriate frequency of performance.

The licensee has proposed to implement changes to surveillance frequencies in the SFCP using the methodology in NEI 04-10, which includes qualitative considerations, results of risk analyses, sensitivity studies and any bounding analyses, recommended monitoring of structures, systems, and components (SSCs), and documentation of the evaluation.

Furthermore, changes to frequencies are subject to regulatory review and oversight of the SFCP implementation through the rigorous NRC review of safety-related SSC performance provided by the reactor oversight process (ROP). Licensees are required by the TSs to perform surveillance test, calibration, or inspection on specific safety-related system equipment (e.g., reactivity control, power distribution, electrical, and instrumentation) to verify system operability.

Surveillance frequencies, currently identified in the TSs, are based primarily upon deterministic methods such as engineering judgment, operating experience, and manufacturer's recommendations.

The licensee's use of approved methodologies identified in NEI 04-10 provides a way to establish risk-informed

-5 surveillance frequencies that complement the deterministic approach and support the NRC's traditional defense-in-depth philosophy.

The licensee's SFCP is intended to ensure that SRs specified in the TSs are performed at intervals sufficient to assure the above regulatory requirements are met. Existing regulatory requirements, such as 10 CFR 50.65, "Requirements for monitoring the effectiveness of maintenance at nuclear power plants," and Appendix B to 10 CFR Part 50 require licensee monitoring of surveillance test failures and implementation of corrective actions to address such failures.

One of these actions may be to consider increasing the frequency at which a surveillance test is performed.

In addition, the SFCP implementation guidance in NEI 04-10 requires monitoring the performance of SSCs for which surveillance frequencies are decreased to assure reduced testing does not adversely impact the SSCs. These requirements, and the monitoring required by NEI 04-10, are intended to ensure that surveillance frequencies are sufficient to assure that the requirements of 10 CFR 50.36 are satisfied and that any performance deficiencies will be identified and appropriate corrective actions taken. Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Informed Decisions on Plant-Specific Changes to the Licensing Basis" (Reference 5), describes a risk-informed approach, acceptable to the NRC, for assessing the nature and impact of proposed licensing-basis changes by considering engineering issues and applying risk insights.

This regulatory guide also provides risk acceptance guidelines for evaluating the results of such evaluations.

RG 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking:

Technical Specifications" (Reference 3), describes an acceptable risk-informed approach specifically for assessing proposed TS changes. RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities" (Reference 4), describes an acceptable approach for determining whether the quality of the PRA, in total or the parts that are used to support an application, is sufficient to provide confidence in the results, such that the PRA can be used in regulatory decision making for light-water reactors.

3.0 TECHNICAL EVALUATION

The licensee's adoption of TSTF-425 for Salem provides for administrative relocation of applicable surveillance frequencies, and provides for the addition of the SFCP to the Administrative Controls section of the TSs. TSTF-425 also requires the application of NEI 04-10 for any changes to surveillance frequencies within the SFCP. The licensee's application for the changes proposed in TSTF-425 included documentation regarding the PRA technical adequacy consistent with the requirements of RG 1.200, Revision 1 (Reference 4). In accordance with NEI 04-10, PRA methods are used, in combination with plant performance data and other considerations, to identify and justify modifications to the surveillance frequencies of equipment at nuclear power plants. This is consistent with the guidance provided in RG 1.174 (Reference

5) and RG 1.177 (Reference 3).

-6 3.1 RG 1.177 Five Key Safety Principles RG 1.177 identifies five key safety principles required for risk-informed changes to the TSs. Each of these principles is addressed by the industry methodology document, NEI 04-10, and is evaluated below in SE Sections 3.1.1 through 3.1.5 with respect to the proposed amendment.

3.1.1 The Proposed Change Meets Current Regulations The first key safety principle in RG 1.177 is that the proposed change meets the current regulations.

Paragraph (c)(3) in 10 CFR 50.36 requires that TSs will include SRs which are "requirements relating to test, calibration, or inspection to assure that necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met." The proposed amendment would relocate most periodic SR frequencies, currently shown in the Salem TSs, to a licensee-controlled program (i.e., the SFCP). The SRs themselves would remain in the TSs, as required by 10 CFR 50.36(c)(3).

The requirements for the SFCP would be added to new TS 6.8.4.1. In accordance with TS 6.8.4.1, any changes to the SR frequencies would be made in accordance with NEI 04-10, Revision 1. By letter dated September 19, 2007 (Reference 9), the NRC staff found that the methodology in NEI 04-10, Revision 1, met NRC regulations, specifically 10 CFR 50.36(c)(3), and was an acceptable program for controlling changes to surveillance frequencies.

Based on the above considerations, the NRC staff concludes that the proposed change is consistent with the requirements in 10 CFR 50.36(c)(3).

Therefore, the proposed change meets the first key safety principle of RG 1.177. 3.1.2 The Proposed Change Is Consistent With the Defense-in-Depth Philosophy Consistency with the defense-in-depth philosophy, the second key safety principle of RG 1.177, is met if: A reasonable balance is preserved among prevention of core damage, prevention of containment failure, and consequence mitigation. Over-reliance on programmatic activities to compensate for weaknesses in plant design is avoided. System redundancy, independence, and diversity are preserved commensurate with the expected frequency, consequences of challenges to the system, and uncertainties (e.g., no risk outliers). Defenses against potential common cause failures are preserved, and the potential for the introduction of new common cause failure mechanisms is assessed. Independence of barriers is not degraded. Defenses against human errors are preserved. The intent of the General Design Criteria in 10 CFR Part 50, Appendix A, is maintained.

TSTF-425 requires the application of NEI 04-10 for any changes to surveillance frequencies within the SFCP. NEI 04-10 uses both the core damage frequency (CDF) and the large early release frequency (LERF) metrics to evaluate the impact of proposed changes to surveillance frequencies.

The guidance of RG 1.174 and RG 1.177 for changes to CDF and LERF is achieved by evaluation using a comprehensive risk analysis, which assesses the impact of proposed changes including contributions from human errors and common cause failures.

Defense-in-depth is also included in the methodology explicitly as a qualitative consideration outside of the risk analysis, as is the potential impact on detection of component degradation that could lead to an increased likelihood of common cause failures.

The NRC staff concludes that both the quantitative risk analysis and the qualitative considerations assure that a reasonable balance of defense-in-depth is maintained.

Therefore, the proposed change meets the second key safety principle of RG 1.177. The Proposed Change Maintains Sufficient Safety Margins The engineering evaluation that wi" be conducted by the licensee under the SFCP, when frequencies are revised, wi" assess the impact of the proposed frequency change in accordance with the principle that sufficient safety margins are maintained.

The guidelines used for making that assessment will include ensuring the proposed surveillance test frequency change is not in conflict with approved industry codes and standards or adversely affects any assumptions or inputs to the safety analysis, or, if such inputs are affected, justification is provided to ensure sufficient safety margin will continue to exist. The design, operation, testing methods, and acceptance criteria for SSCs, specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plant licensing basis (including the Updated Final Safety Analysis Report and bases to the TSs), since these are not affected by changes to the surveillance frequencies.

Similarly, there is no impact to safety analysis acceptance criteria as described in the plant licensing basis. Based on the above considerations, the NRC staff concludes that there is reasonable assurance that safety margins will be maintained through use of the SFCP methodology.

Therefore, the proposed change meets the third key safety principle of RG 1.177. When Proposed Changes Result in an Increase in Core Damage Frequency or Risk, the Increases Should Be Small and Consistent With the Intent of the Commission's Safety Goal Policy Statement RG 1.177 provides a framework for evaluating the risk impact of proposed changes to surveillance frequencies.

This requires the identification of the risk contribution from impacted surveillances, determination of the risk impact from the change to the proposed surveillance frequency, and performance of sensitivity and uncertainty evaluations.

TSTF-425 requires application of NEI 04-10 in the SFCP. NEI 04-10 satisfies the intent of RG 1.177 requirements for evaluating the change in risk, and for assuring that such changes are small. 3.1.4.1 Quality of the PRA The quality of the Salem PRA is compatible with the safety implications of the proposed TS change and the role the PRA plays in justifying the change. That is, the more the potential

-change in risk or the greater the uncertainty in that risk from the requested TS change, or both, the more rigor that must go into ensuring the quality of the PRA. The licensee used RG 1.200 to address the technical adequacy of the Salem PRA. RG 1.200 is NRC's developed regulatory guidance which, in Revision 1, endorsed with comments and qualifications the use of the American Society of Mechanical Engineers (ASME) RA-Sb-200S, "Addenda to ASME RA-S-2002 Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications," (Reference 6), NEI 00-02, "PRA Peer Review Process Guidelines," (Reference

7) and NEI OS-04, "Process for Performing Follow-On PRA Peer Reviews Using the ASME PRA Standard" (Reference 8). The licensee has performed an assessment of the PRA models used to support the SFCP against the requirements of RG 1.200 to assure that the PRA models are capable of determining the change in risk due to changes to surveillance frequencies of SSCs, using plant-specific data and models. Capability category II of ASME RA-Sb-200S was applied as the standard, and any identified deficiencies to those requirements are assessed further to determine any impacts to proposed decreases to surveillance frequencies, including by the use of sensitivity studies where appropriate.

The NRC staff notes that in Revision 2, RG 1.200 endorsed with comments and qualifications an updated combined standard which includes requirements for fire, seismic, and other external events PRA models. The existing internal events standard was subsumed into the combined standard, but the technical requirements are essentially unchanged.

Since NEI 04-10 Rev. 1 specifically identified the use of RG 1.200 Revision 1 to assess the internal events standard, the licensee's approach is reasonable and consistent with the approved methodology.

The licensee used a full scope peer review of its internal events PRA model. The NRC staff reviewed the licensee's assessment of the Salem PRA and the remaining open deficiencies that do not conform to capability category II of the ASME PRA standard (Table 2-1 of Attachment 2 of the licensee's application dated March 23, 2011). The staffs assessment of the remaining open "gaps," to assure that they may be addressed and dis positioned for each surveillance frequency evaluation per the NEI 04-10 methodology, is provided below. IE-A3a-01, IE-A4-01, I E-A6-0 1 , IE-A7 -01, IE-83-01, IE-C1 b-01, SC-8S-01, SC-C3-02, SY-A4-01, SY-A 19-01, SY-BS-01, DA-E2-01, QU-BS-01, QU-B9-01, QU-D1 b-01, QU-D3-01, QU-D4-01, QU-F2-01, LE-F3-01, LE-GS-01, LE-G6-01:

The peer review identified 21 findings for which the deficiency was identified as inadequate documentation.

These findings do not impact the technical adequacy of the PRA model with regards to risk calculations to support implementation of the SFCP, and therefore can be addressed per the methodology of NEI 04-10. IE-A 1-01: The loss of an alternating current (AC) bus is not modeled nor is it documented as to its plant impact (Le., a reactor trip or a forced shutdown).

The licensee considers the omission to be a minimal impact because the event is bounded by other modeled trip initiators.

The NRC staff concurs that a qualitative consideration of initiator impact can be addressed per the methodology of NEI 04-10. IE-A3-01:

Support system initiating events fault tree include a multiplier which should be replaced with the newer methodology of Electric Power and Research Institute (EPRI)-1013490, "Support System Initiating Events: Identification and Quantification Guideline." The licensee identified that this gap is important for changes involving the support systems and will be addressed using the EPRI methodology.

The NRC staff concurs that this deficiency can be addressed per the methodology of NEI 04-10.

-I E-AS-O1 : Failure to consider plant-specific events occurring at plant conditions other than power operations, as well as controlled shutdown events, was identified as potentially resulting in exclusion of valid initiators.

The licensee characterized this finding as related only to documentation, and further confirmed that all plant shutdowns were reviewed to support the development of the initiating event scope. The NRC staff, therefore, accepts that this deficiency can be addressed per the methodology of NEI 04-10. AS-A7-01:

Non-minimal sequences are subsumed in the single-top fault tree model and could result in the loss of risk insights or masking of important sequences.

The licensee identified that the issue was determined to be related to inclusion of containment isolation status, which is now addressed in separate logic associated with the level two PRA model. The licensee thus characterized this finding as related only to documentation, thereby indicating that masking of risk insights is not occurring.

The NRC staff, therefore, accepts that this deficiency can be addressed per the methodology of NEI 04-10. AS-A7-02:

Flooding of plant areas due to leakage through residual heat removal (RHR) components prior to isolation of the interfacing systems loss-of-coolant accident (ISLOCA) by the operator has not been considered.

The licensee identified evaluations performed which considered the water volume involved and the potential for flooding damage beyond the immediate area of the pipe break, which show that there is no potential for additional damage to plant equipment.

The licensee characterized this finding as related only to documentation.

The NRC staff, therefore, accepts that this deficiency can be addressed per the methodology of NEI04-10.

AS-AS-01:

Issues affecting sequences involving successful recovery of offsite power were identi'fied involving recovery of mitigating systems, mission times, and possible double crediting of offsite power recovery.

The licensee identified that the issues of mission times and offsite power recovery credit have been resolved.

The issue of not evaluating the status of mitigation equipment failures after an assumed recovery of offsite power has been qualitatively evaluated by the licensee as an insignificant contributor to risk which will be addressed using sensitivity studies as needed. The NRC staff, therefore, accepts that this deficiency can be addressed per the methodology of NEI 04-10. SY-AS-01, SY-AS-01, SY-A10-01, SY-A12-01, and SY-A13-01:

These findings collectively address deficiencies in system boundaries and apparently missing failure modes. The licensee provided details of the findings and its review, which characterized these findings as related only to documentation, thereby indicating that the technical deficiencies are not valid, but simply not adequately documented.

The NRC staff, therefore, accepts that this deficiency can be addressed per the methodology of NEI 04-10. SY-A1S-01:

The basis for the time available to recover a service water train using a header crosstie valve is based on room heatup rather than recovery of diesel generator cooling for loss of offsite power sequences.

This impacts one input to a single human reliability event, and can be addressed by sensitivity studies if necessary for specific surveillance frequency change evaluations.

The NRC staff concurs that this deficiency can be addressed per the methodology of NEI 04-10. SY-A21-01:

The correlation of component failure data is not correctly addressed due to the use of system-specific coding of the model. The licensee identified that the scope of the issue is limited to component failure modes with inadequate data sources, requiring that data sources applicable to a similar component failure mode are applied. The NRC staff concurs that this deficiency can be addressed per the methodology of NEI 04-10.

-10 SY-B11-01:

Some auxiliary feedwater (AFW) actuation signals are not modeled, and no justification is provided for their exclusion.

The licensee identified that the actuation logic modeled reflects those with the most impact, and that there is no technical deficiency.

The NRC staff notes that it is typical of PRA models to only include representative actuation logic, and that the exclusion of redundant actuation logic is conservative.

The staff concurs that this deficiency can be addressed per the methodology of NEI 04-10. HR-C3-01:

Miscalibration of standby systems was not considered in the model. The licensee stated that such failure modes would be added to the model to account for these failure modes if the surveillance frequency of the instrumentation was being evaluated.

The NRC staff concurs that this deficiency can be addressed per the methodology of NEI 04-10. DA-A 1 a-01: Component boundaries are not clearly defined, which can cause duplication or omission of failures.

The licensee stated that the process used to define boundaries was consistent with failure data sources and that unavailability data is consistent with the system analysis definition.

The supporting requirement is met and the issue is related only to the documentation of the boundaries.

The NRC staff concurs that this deficiency can be addressed per the methodology of N E I 04-1 O. DA-C1-01:

Generic unavailability data applied does not have a justification that the data is consistent with the test and maintenance practices at the plant. The licensee identified that for the limited scope of components using this generic data, either there was zero plant-specific unavailability (which would make the generic data conservative), or a determination of applicability was completed.

The NRC staff concurs that this deficiency can be addressed per the methodology of NEI 04-10. DA-C2-01:

Plant-specific data was not applied to all risk-significant components.

The licensee stated that the majority of high importance systems have been updated with recent plant-specific data, and that failure rate sensitivity analyses will be performed to address this deficiency.

The NRC staff concurs that this deficiency can be addressed per the methodology of NEI 04-10. DA-C6-01:

Demand data for standby components was not documented.

The licensee stated that the demand data was based on a review of surveillance test and operational data, and is documented, but that the applicable references are not identified completely in the data notebook.

Therefore, this deficiency is in documentation and not a technical issue. The NRC staff concurs that this deficiency can be addressed per the methodology of NEI 04-10. DA-C1 0-01: Documentation of the process for review of test procedures for surveillance test data could not be identified.

The licensee stated that sensitivity analyses would be performed on failure probabilities.

The NRC staff concurs that this deficiency can be addressed per the methodology of NEI 04-10. DA-D4-01:

No documentation of the comparison of the generic and plant-specific data values was found for Bayes updates to confirm expected values. The licensee stated that comparisons were made to ensure updated values were reasonable, and the deficiency is in the documentation of the results. The NRC staff concurs that this deficiency can be addressed per the methodology of NEI 04-10. DA-D6-01:

No comparison of data sources (common cause failures) was identified.

The licensee identified the sources of generic common cause factors used in its models, and stated that there was no plant-specific evidence which would invalidate their use, and that the deficiency was in the documentation.

The NRC staff concurs that this deficiency can be addressed per the methodology of NEI 04-10.

-11 LE-C8a-01:

There was no evidence of consideration of equipment survivability and human actions under adverse environmental conditions.

The licensee stated that no credit is taken, and that this is a documentation issue. The NRC staff concurs that this deficiency can be addressed per the methodology of NEI 04-10. LE-F1 b-01: There are no checks on the reasonableness of the results of the LERF contributors.

The licensee stated that the LERF results and distributions are reasonable and consistent with typical results for similar plants with large dry containments.

The licensee identifies this deficiency as a documentation issue. The NRC staff concurs that this deficiency can be addressed per the methodology of NEI 04-10. Based on the licensee's assessment using the applicable PRA standard and RG 1.200, the level of PRA quality, combined with the proposed evaluation and disposition of the identified deficiencies, is sufficient to support the evaluation of changes proposed to surveillance frequencies within the SFCP, and is consistent with Regulatory Position 2.3.1 of RG 1.177. 3.1.4.2 Scope of the PRA The licensee is required to evaluate each proposed change to a relocated surveillance frequency using the guidance contained in NEI 04-10 to determine its potential impact on risk, due to impacts from internal events, fires, seismic, other external events, and from shutdown conditions.

Consideration is made of both CDF and LERF metrics. In cases where a PRA of sufficient scope or where quantitative risk models were unavailable, the licensee uses bounding analyses, or other conservative quantitative evaluations.

A qualitative screening analysis may be used when the surveillance frequency impact on plant risk is shown to be negligible or zero. The Individual Plant Examination of External Events (IPEEE) fire-induced vulnerability evaluation analysis, and the IPEEE seismic margins analysis, will be used to provide insights for fires and seismic events. Other external hazards were screened during the IPEEE assessment, and will, therefore, be qualitatively assessed for this application.

The licensee's evaluation methodology is sufficient to ensure the scope of the risk contribution of each surveillance frequency change is properly identified for evaluation, and is consistent with Regulatory Position 2.3.2 of RG 1.177. 3.1.4.3 PRA Modeling The licensee will determine whether the SSCs affected by a proposed change to a surveillance frequency are modeled in the PRA. Where the SSC is directly or implicitly modeled, a quantitative evaluation of the risk impact may be carried out. The methodology adjusts the failure probability of the impacted SSCs, including any impacted common cause failure modes, based on the proposed change to the surveillance frequency.

Where the SSC is not modeled in the PRA, bounding analyses are performed to characterize the impact of the proposed change to the surveillance frequency.

Potential impacts on the risk analyses due to screening criteria and truncation levels are addressed by the requirements for PRA technical adequacy consistent with guidance contained in RG 1.200, and by sensitivity studies identified in NEI 04-10.

-12 The licensee will perform quantitative evaluations of the impact of selected testing strategy (I.e., staggered testing or sequential testing) consistent with the guidance of NUREG/CR-6141 and NUREG/CR-S497, as discussed in NEI 04-10. Thus, through the application of NEI 04-10, the Salem PRA modeling is sufficient to ensure an acceptable evaluation of risk for the proposed changes in surveillance frequency, and is consistent with Regulatory Position 2.3.3 of RG 1.177. 3.1.4.4 Assumptions for Time-Related Failure Contributions The failure probabilities of SSCs modeled in the Salem PRA include a standby time-related contribution and a cyclic demand-related contribution.

NEI 04-10 criteria adjust the time-related failure contribution of SSCs affected by the proposed change to surveillance frequency.

This is consistent with RG 1.177 Section 2.3.3 which permits separation of the failure rate contributions into demand and standby for evaluation of SRs. If the available data do not support distinguishing between the time-related failures and demand failures, then the change to surveillance frequency is conservatively assumed to impact the total failure probability of the SSC, including both standby and demand contributions.

The SSC failure rate (per unit time) is assumed to be unaffected by the change in test frequency, and will be confirmed by the required monitoring and feedback implemented after the change in surveillance frequency is implemented.

The process requires consideration of qualitative sources of information with regard to potential impacts of test frequency on SSC performance, including industry and specific operating experience, vendor recommendations, industry standards, and code-specified test intervals.

Thus, the process is not reliant upon risk analyses as the sole basis for the proposed changes. The potential beneficial risk impacts of reduced surveillance frequency, including reduced downtime, lesser potential for restoration errors, reduction of potential for test caused transients, and reduced test-caused wear of equipment, are identified qualitatively.

Thus, through the application of NEI 04-10, the licensee has employed reasonable assumptions with regard to extensions of surveillance test intervals, and its approach is consistent with Regulatory Position 2.3.4 of RG 1.177. 3.1.4.S Sensitivity and Uncertainty Analyses NEI 04-10 requires sensitivity studies to assess the impact of uncertainties from key assumptions of the PRA, uncertainty in the failure probabilities of the affected SSCs, impact to the frequency of initiating events, and of any identified deviations from capability category II of ASME PRA Standard ASME RA-Sb-200S (Reference 6). Where the sensitivity analyses identify a potential impact on the proposed change, revised surveillance frequencies are considered, along with any qualitative considerations that may bear on the results of such sensitivity studies. Required monitoring and feedback of SSC performance, once the revised surveillance frequencies are implemented, will also be performed.

Thus, through the application of NEI 04-10, the licensee has appropriately considered the possible impact of PRA model uncertainty and sensitivity to key assumptions and model limitations, and its approach is consistent with Regulatory Position 2.3.S of RG 1.177.

-3.1.4.6 Acceptance Guidelines The licensee will quantitatively evaluate the change in total risk (including internal and external events contributions) in terms of CDF and LERF for both the individual risk impact of a proposed change in surveillance frequency and the cumulative impact from all individual changes to surveillance frequencies using the guidance contained in NRC-approved NEI 04-10 in accordance with the TS SFCP. Each individual change to surveillance frequency must show a risk impact below 1 E-6 per year for change to CDF, and below 1 E-7 per year for change to LERF. These criteria are consistent with the limits of RG 1.174 for very small changes in risk. Where the RG 1.174 limits are not met, the process either considers revised surveillance frequencies which are consistent with RG 1.174 or the process terminates without permitting the proposed changes. Where quantitative results are unavailable to permit comparison to acceptance guidelines, appropriate qualitative analyses are required to demonstrate that the associated risk impact of a proposed change to surveillance frequency is negligible or zero. Otherwise, bounding quantitative analyses are required which demonstrate the risk impact is at least one order of magnitude lower than the RG 1.174 acceptance guidelines for very small changes in risk. In addition to assessing each individual SSC surveillance frequency change, the cumulative impact of all changes must result in a risk impact below 1 E-S per year for change to CDF, and below 1 E-6 per year for change to LERF, and the total CDF and total LERF must be reasonably shown to be less than 1 E-4 per year and 1 E-S per year, respectively.

These are consistent with the limits of RG 1.174 for acceptable changes in risk, as referenced by RG 1.177 for changes to surveillance frequencies.

The NRC staff interprets this assessment of cumulative risk as a requirement to calculate the change in risk from a baseline model utilizing failure probabilities based on the surveillance frequencies prior to implementation of the SFCP, compared to a revised model with failure probabilities based on changed surveillance frequencies.

The NRC staff further notes that the licensee includes a provision to exclude the contribution to cumulative risk from individual changes to surveillance frequencies associated with insignificant risk increases (less than SE-8 CDF and SE-9 LERF) once the baseline PRA models are updated to include the effects of the revised surveillance frequencies. The quantitative acceptance guidance of RG 1.174 is supplemented by qualitative information to evaluate the proposed changes to surveillance frequencies, including industry and plant-specific operating experience, vendor recommendations, industry standards, the results of sensitivity studies, and SSC performance data and test history. The final acceptability of the proposed change is based on all of these considerations and not solely on the PRA results compared to numerical acceptance guidelines.

Post implementation performance monitoring and feedback are also required to assure continued reliability of the components.

The NRC staff concludes that the licensee's application of NEI 04-10 provides reasonable acceptance guidelines and methods for evaluating the risk increase of proposed changes to surveillance frequencies, consistent with Regulatory Position 2.4 of RG 1.177. Therefore, the proposed change satisfies the fourth key safety principle of RG 1.177 by assuring that any increase in risk is small consistent with the intent of the Commission's Safety Goal Policy Statement.

-The Impact of the Proposed Change Should Be Monitored Using Performance Measurement Strateg ies The licensee's adoption of TSTF-425 requires application of NEI 04-10 in the SFCP. NE104-10 requires performance monitoring of SSCs whose surveillance frequency has been revised as part of a feedback process to assure that the change in test frequency has not resulted in degradation of equipment performance and operational safety. The monitoring and feedback includes consideration of maintenance rule monitoring of equipment performance.

In the event of degradation of SSC performance, the surveillance frequency will be reassessed in accordance with the methodology, in addition to any corrective actions which may apply as part of the maintenance rule requirements.

The NRC staff concludes that the performance monitoring and feedback specified in NEI 04-10 is sufficient to reasonably assure acceptable SSC performance and is consistent with Regulatory Position 3.2 of RG 1.177. Therefore, the proposed change meets the fifth key safety principle of RG 1.177. Addition of Surveillance Frequency Control Program to Administrative Controls The proposed amendment would add the SFCP to the Administrative Controls section of the Salem TSs. Specifically, new TS 6.8.4.1, "Surveillance Frequency Control Program," would read as follows: This program provides controls for Surveillance Frequencies.

The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure that the associated Limiting Conditions for Operation are met. The Surveillance Frequency Control Program shall contain a list of Frequencies of the Surveillance Requirements for which the Frequency is controlled by the program. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1. The provisions of Surveillance Requirements 4.0.2 and 4.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program. The NRC staff concludes that the proposed addition to the Administrative Controls section of the TSs adequately identifies the scope of the SFCP and defines the methodology to be used in a revision of SR frequencies.

Therefore, the proposed TS change is acceptable. TS Bases Changes PSEG's application dated March 23, 2010, provided proposed changes to the TS Bases to be implemented with the associated TS changes. These pages were provided for information only and will be revised in accordance with the Salem TS Bases Control Program.

-15 3.4 Technical Evaluation Conclusion The NRC staff has reviewed the licensee's proposed relocation of certain surveillance frequencies to a new licensee-controlled program, the SFCP, and its proposal to control changes to surveillance frequencies in accordance with the new program. Based on the above considerations, the NRC staff concludes that the proposed amendment is acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the New Jersey State Official was notified of the proposed issuance of amendments.

The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION The amendments change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes SRs. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure.

The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (75 FR 33843). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (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 amendments will not be inimical to the common defense and security or to the health and safety of the public. Principal Contributors:

A. Howe R. Ennis Date: March 21, 2011

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7.0 REFERENCES

TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control-RITSTF Initiative 5b," March 18, 2009 (ADAMS Accession Number: ML090850642). NEI 04-10, Revision 1, "Risk-Informed Technical Specifications Initiative 5B, Risk-Informed Method for Control of Surveillance Frequencies," April 2007 (ADAMS Accession Number: ML071360456). Regulatory Guide 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking:

Technical Specifications," August 1998 (ADAMS Accession Number: ML003740176). Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 1, January 2007 (ADAMS Accession Number: ML070240001). Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Informed Decisions on Plant-Specific Changes to the Licensing Basis," NRC, Revision 1, November 2002 (ADAMS Accession Number: ML023240437). ASME PRA Standard ASME RA-Sb-2005, Addenda to ASME RA-S-2002, "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Application." NEI 00-02, Revision 1 "Probabilistic Risk Assessment (PRA) Peer Review Process Guidance," Revision 1, May 2006 (ADAMS Accession Number: ML061510621). NEI 05-04, "Process for Performing Follow-On PRA Peer Reviews Using the ASME PRA Standard," Revision 0, August 2006. Letter from Ho. K. Nieh (NRC) to Biff Bradley (NEI), "Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) 04-10, Revision 1, Risk-Informed Technical Specification Initiative 5-B, Risk-Informed Method for Control of Surveillance Frequencies," dated September 19, 2007 (ADAMS Accession No. ML072570267).

March 21, 2011 Mr. Thomas Joyce President and Chief Nuclear Officer PSEG Nuclear P.O. Box 236, N09 Hancocks Bridge, NJ 08038 SALEM NUCLEAR GENERATING STATION, UNIT NOS. 1 AND 2, ISSUANCE OF AMENDMENTS RE: RELOCATION OF SPECIFIC SURVEILLANCE FREQUENCIES TO A LICENSEE-CONTROLLED PROGRAM BASED ON TECHNICAL SPECIFICATION TASK FORCE (TSTF) CHANGE TSTF-425 (TAC NOS. ME3574 AND ME3575)

Dear Mr. Joyce:

The Commission has issued the enclosed Amendment Nos. 299 and 282 to Facility Operating License Nos. DPR-70 and DPR-75 for the Salem Nuclear Generating Station, Unit Nos. 1 and 2. These amendments consist of changes to the Technical Specifications (TSs) and Facility Operating Licenses in response to your application dated March 23, 2010, as supplemented by letters dated November 19, 2010, January 31,2011, and February 23,2011. The amendments modify the TSs by relocating specific surveillance frequencies to a controlled program. The changes are based on Nuclear Regulatory Commission-approved TS Task Force (TSTF) change TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control-RITSTF [Risk-Informed TSTF] Initiative 5b." A copy of our safety evaluation is also enclosed.

Notice of Issuance will be included in the Commission's biweekly Federal Register notice. Sincerely, Ira! Richard B. Ennis, Senior Project Manager Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-272 and 50-311

Enclosures:

1. Amendment No. 299 to License No. DPR-70 2. Amendment No. 282 to License No. DPR-75 3. Safety Evaluation cc w/encls: See next page DISTRI BUTION: PUBLIC LPLI-2 R/F RidsAcrsAcnw_MailCTR Resource RidsNrrDirsltsb Resource RidsNrrDorlDpr Resource RidsNrrDorlLpl1-2 Resource RidsNrrPMSalem Resource RidsOgcRp Resource RidsNrrLAABaxter Resource RidsRgn1 MailCenter Resource RidsNrrDraApla Resource GWaig, NRRlDIRS/ITSB AHowe, NRR/DRAIAPLA Accession No.: ML

via SE dated 2/3/11 OGC OFFICE LPL1-2/PM APLAlBC* NAME REnnis DHarrison STurk DATE 3/10/11 2/23/11 3/1111 2/3/11 3/9/11 OFFICIAL RECORD COPY

v t e Letter Sequence Approval
TAC:ME3574, Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5B (Approved, Closed) TAC:ME3575, Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5B (Approved, Closed)
Initiation Request Acceptance Supplement Administration Questions 1 September 2010 20 December 2010 Answers Results Approval Other:
MONTHYEAR2010-09-01 [Table View]

ML110410691

Text

Dear Mr. Joyce:

Enclosures:

Attachment:

Attachment:

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

4.0 STATE CONSULTATION

6.0 CONCLUSION

7.0 REFERENCES

Dear Mr. Joyce:

Enclosures:

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