Text

Issuance of Amendment Nos. 324 and 305 Revise Technical Specification Actions for Rod Position Indicators
	ML18085B198
Person / Time
Site:	Salem
Issue date:	04/18/2018
From:	James Kim; Plant Licensing Branch 1
To:	Sena P; Public Service Enterprise Group
	Kim J, NRR/DORL/LPLI, 415-4125
References
	EPID L-2018-LLA-0038
	Download: ML18085B198 (22)
	v • d • e

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O.C. 20555-0001 April 18, 2018 Mr. Peter P. Sena, Ill President and Chief Nuclear Officer PSEG Nuclear LLC - N09 Salem Nuclear Generating Station P.O. Box 236 Hancocks Bridge, NJ 08038

SUBJECT:

SALEM NUCLEAR GENERATING STATION, UNIT NOS. 1 AND 2- ISSUANCE OF AMENDMENT NOS. 324 AND 305 RE: REVISE TECHNICAL SPECIFICATION ACTIONS FOR ROD POSITION INDICATORS (EPID L-2018-LLA-0038)

Dear Mr. Sena:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment Nos. 324 and 305 to Renewed Facility Operating License Nos. DPR-70 and DPR-75 for the Salem Nuclear Generating Station, Unit Nos. 1 and 2, respectively. These amendments consist of changes to the Technical Specifications (TSs) in response to your application dated February 8, 2018.

The amendments modify the TS-allowed outage time for more than one inoperable analog rod position indicator from 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and change the basis for entry into the TS actions for inoperable rod position indicators from "per bank" to "per group." The amendments also separate existing TS 3.1.3.2.1, Action a.1, into two separate actions and remove the duplicative Action b (Unit No. 1 only).

A copy of the related safety evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely,

~F-James Kim, Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-272 and 50-311

Enclosures:

1. Amendment No. 324 to DPR-70

2. Amendment No. 305 to DPR-75

3. Safety Evaluation cc w/enclosures: Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O.C. 20555-0001 PSEG NUCLEAR LLC EXELON GENERATION COMPANY, LLC DOCKET NO. 50-272 SALEM NUCLEAR GENERATING STATION, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 324 Renewed License No. DPR-70

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment filed by PSEG Nuclear LLC, acting on behalf of itself and Exelon Generation Company, LLC (the licensees) dated February 8, 2018, 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 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. 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; D. 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 E. 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.

Enclosure 1

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the atta,ehment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-70 is hereby amended to read as follows:

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 324, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. PSEG Nuclear LLC shall operate the facility in accordance with the Technical Specifications, and the Environmental Protection Plan.

3. This license amendment is effective as of its date of issuance and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION Ja s G. Danna, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License and Technical Specifications Date of Issuance: Apr i 1 1 8 , 2 O1 8

ATTACHMENT TO LICENSE AMENDMENT NO. 324 SALEM NUCLEAR GENERATING STATION, UNIT NO. 1 RENEWED FACILITY OPERATING LICENSE NO. DPR-70 DOCKET NO. 50-272 Replace the following page of Renewed 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 Insert Page 3 Page 3 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 3/4 1-19 3/4 1-19 3/4 1-19a 3/4 1-19a

instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5) PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6) PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30 and 70, to possess but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level PSEG Nuclear LLC is authorized to operate the facility at a steady state reactor core power level not in excess of 3459 megawatts (one hundred percent of rated core power).

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 324, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. PSEG Nuclear LLC shall operate the facility in accordance with the Technical Specifications, and the Environmental Protection Plan.

(3) Deleted Per Amendment 22, 11-20-79 (4) Less than Four Loop Operation PSEG Nuclear LLC shall not operate the reactor at power levels above P-7 (as defined in Table 3.3-1 of Specification 3.3.1.1 of Appendix A to this renewed license) with less than four (4) reactor coolant loops in operation until safety analyses 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 renewed license.

(5) PSEG Nuclear LLC shall implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Renewed License No. DPR-70 Amendment No. 324

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

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

All Shutdown Banks: +/- 18 steps at s 85% reactor power or if reactor power is > 85%

RATED THERMAL POWER +/- 12 steps of the group demand counters for withdrawal ranges of 0-30 steps and 200-230 steps.

Control Bank A: +/- 18 steps at s 85% reactor power or if reactor power is > 85%

RATED THERMAL POWER +/- 12 steps of the group demand counters for withdrawal ranges of 0-30 steps and 200-230 steps.

Control Bank B: +/- 18 steps at s 85% reactor power or if reactor power is > 85%

RATED THERMAL POWER +/- 12 steps of the group demand counters for withdrawal ranges of 0-30 steps and 160-230 steps.

Control Bank C and D: +/- 18 steps at s 85% reactor power or if reactor power is > 85%

RATED THERMAL POWER +/- 12 steps of the group demand counters for withdrawal ranges of 0-230 steps.

b. Group demand counters; +/- 2 steps of the pulsed output of the Slave Cycler Circuit over the withdrawal range of 0-230 steps.

APPLICABILITY: MODES 1 and 2.

ACTION:

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

1. Determine the position of the non-indicating rod(s) indirectly using the power distribution monitoring system (if power is above 25% RTP) or using the movable incore detectors (if power is less than 25% RTP or the power distribution monitoring system is inoperable) at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , or

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

SALEM - UNIT 1 3/4 1-19 Amendment No. 324

REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

b. With two or more analog rod position indicators per group inoperable:

1. Immediately place the control rods in manual control, and

2. Monitor and record Reactor Coolant System T avg once every hour, and

3. Verify the position of the rods with inoperable position indicators indirectly using the power distribution monitoring system (if power is above 25% RTP) or using the movable incore detectors (if power is less than 25% RTP or the power distribution monitoring system is inoperable) at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , and

4. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months restore the inoperable rod position indicators to OPERABLE status such that a maximum of one rod position indicator per group is inoperable, or

5. Be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c. When one or more rods with inoperable position indicators have been moved in excess of 24 steps in one direction since the last determination of the rod's position:

1. Determine the position of the non-indicating rod(s) indirectly using the power distribution monitoring system (if power is above 25% RTP) or using the movable incore detectors (if power is less than 25% RTP or the power distribution monitoring system is inoperable) within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , or

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

d. With a maximum of one group demand position indicator per bank inoperable either:

1. 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 ::; 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

2. 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::; 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 1 3/4 1-19a Amendment No. 324

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 PSEG NUCLEAR LLC EXELON GENERATION COMPANY, LLC DOCKET NO. 50-311 SALEM NUCLEAR GENERATING STATION, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 305 Renewed License No. DPR-75

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment filed by PSEG Nuclear LLC, acting on behalf of itself and Exelon Generation Company, LLC (the licensees) dated February 8, 2018, 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 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. 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; D. 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 E. 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.

Enclosure 2

2. 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 Renewed 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 Appendix A, as revised through Amendment No. 305, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. PSEG Nuclear LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. This license amendment is effective as of its date of issuance and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION 9~

James G. Danna, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License and Technical Specifications Date of Issuance: April 1 8, 2 O1 8

ATTACHMENT TO LICENSE AMENDMENT NO. 305 SALEM NUCLEAR GENERATING STATION, UNIT NO. 2 RENEWED FACILITY OPERATING LICENSE NO. DPR-75 DOCKET NO. 50-311 Replace the following page of Renewed Facility Operating License No. DPR-75 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 Insert Page 3 Page 3 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 3/4 1-16 3/4 1-16 3/4 1-16a 3/4 1-16a

(4) PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use at any time any byproduct, source or special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration and as fission detectors in amounts as required; (5) PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6) PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This renewed license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level PSEG Nuclear LLC is authorized to operate the facility at steady state reactor core power levels not in excess of 3459 megawatts (thermal).

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 305, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. PSEG Nuclear LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Renewed License No. DPR-75 Amendment No. 305

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

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

All Shutdown Banks: +/- 18 steps at s 85% reactor power or if reactor power is > 85%

RATED THERMAL POWER +/- 12 steps of the group demand counters for withdrawal ranges of 0-30 steps and 200-230 steps.

a Control Bank A: +/- 1 steps at s 85% reactor power or if reactor power is > 85% RATED THERMAL POWER+/- 12 steps of the group demand counters for withdrawal ranges of 0-30 steps and 200-230 steps.

Control Bank B: +/- 18 steps at s 85% reactor power or if reactor power is > 85% RATED THERMAL POWER +/- 12 steps of the group demand counters for withdrawal ranges of 0-30 steps and 160-230 steps.

Control Banks C and D: +/- 18 steps at s 85% reactor power or if reactor power is > 85%

RATED THERMAL POWER +/- 12 steps of the group demand counters for withdrawal range of 0-230 steps.

b. Group demand counters; +/- 2 steps of the pulsed output of the Slave Cycler Circuit over the withdrawal range of 0-230 steps.

APPLICABILITY: MODES 1 and 2.

ACTION:

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

1. Determine the position of the non-indicating rod(s) indirectly using the power distribution monitoring system (if power is above 25% RTP) or using the movable incore detectors (if power is less than 25% RTP or the power distribution monitoring system is inoperable) at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , or

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

SALEM - UNIT 2 3/4 1-16 Amendment No. 305

REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION <Continued}

b. With two or more analog rod position indicators per group inoperable:

1. Immediately place the control rods in manual control, and

2. Monitor and record Reactor Coolant System Tavg once every hour, and

3. Verify the position of the rods with inoperable position indicators indirectly using the power distribution monitoring system (if power is above 25% RTP) or using the movable incore detectors (if power is less than 25% RTP or the power distribution monitoring system is inoperable) at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , and

4. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months restore the inoperable rod position indicators to OPERABLE status such that a maximum of one rod position indicator per group is inoperable, or

5. Be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c. When one or more rods with inoperable position indicators have been moved in excess of 24 steps in one direction since the last determination of the rod's position:

1. Determine the position of the non-indicating rod(s) indirectly using the power distribution monitoring system (if power is above 25% RTP) or using the movable incore detectors (if power is less than 25% RTP or the power distribution monitoring system is inoperable) within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , or

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

d. With a maximum of one group demand position indicator per bank inoperable either:

1. 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

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

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 2 3/4 1-16a Amendment No. 305

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 324 AND 305 TO RENEWED 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 February 8, 2018 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML18040A301), PSEG Nuclear LLC (PSEG or the licensee) submitted a request for changes to the Salem Nuclear Generating Station (Salem), Unit Nos. 1 and 2, Technical Specifications (TSs). Specifically, the application requests changes to Salem, Unit Nos. 1 and 2, TS 3.1.3.2.1, "Position Indication Systems - Operating," to modify the TS action for more than one inoperable analog rod position indicator from one hour to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months consistent with Technical Specification Task Force (TSTF) traveler TSTF-234-A, Revision 1, "Add Action for More Than One [D]RPI lnoperable," 1 and to align the TS actions with NUREG-1431, Revision 4, "Standard Technical Specifications - Westinghouse Plants" (ADAMS Accession No. ML12100A222).

2.0 REGULATORY EVALUATION

2.1 System Description Salem, Unit Nos. 1 and 2, use an analog rod position indication system that senses and displays control rod position. The licensee describes this system as follows:

An electrical coil stack linear variable differential transformer is placed above the stepping mechanisms of the control rod magnetic jacks external to the rod/reactor coolant system pressure housing. When the associated control rod is at the bottom of the core, the magnetic coupling between primary and secondary windings is small and there is a small voltage induced in the secondary winding.

As the magnetic jacks raise the control rod, the relatively high permeability of the lift rod causes an increase in magnetic coupling. Thus, an analog signal proportional to rod position is derived.

1 EXCEL Services Corporation, 2003, Rockville, MD.

Enclosure 3

Direct, continuous readout of every RCCA [rod cluster control assembly] position is presented to the operator by individual control board meter indications without the need for operator selection or switching to determine rod position. The rod position is also displayed on the plant computer.

Salem, Unit Nos. 1 and 2, utilize 53 RCCAs (rods) separated into shutdown rods and control rods. There are 24 shutdown rods and 29 control rods. The RCCAs are inserted into the fuel assemblies at distinct locations to uniformly control reactivity. The 53 RCCAs are separated into banks and groups as shown below.

Shutdown Rods Bank A Bank B BankC Bank D Group 1 I Group 2 Group 1 I Group 2 Group 1 Group 1 4 I 4 4 I 4 4 4 Control Rods Bank A Bank B BankC Bank D Group 1 Group 2 Group 1 Group 2 Group 1 Group 2 Group 1 Group 2 4- Unit 1 4- Unit 1 2 - Unit 1 2 - Unit 1 4 4 4 5 2- Unit 2 2- Unit 2 4- Unit 2 4- Unit 2 The RCCA operation is described by the licensee as follows.

The four control banks (A, B, C, D) are the only rods that can be operated under automatic control. All RCCAs in a group are paralleled to step simultaneously.

Reactor startup is accomplished by first manually withdrawing the shutdown rod banks to the full out position. The control rod banks are then withdrawn manually and sequentially by the operator. Control rod movement is automatically programmed to withdraw the control rods in a predetermined sequence. The control rod programming is sequenced such that as the first control rod bank being withdrawn reaches a preset position, the second control rod bank begins to move out simultaneously with the first bank. The staggered withdrawal sequence continues until the control rod banks either reach their fully withdrawn position, or reach the desired position to control axial flux. Normally all rods are fully withdrawn at full power. The programmed insertion sequence, manual or automatic, is the opposite of the withdrawal sequence (i.e., the last control rod bank out is the first control rod bank in).

The shutdown rod groups together with the control rod groups are capable of shutting the reactor down under all conditions. They are used in conjunction with the adjustment of soluble boron to provide shutdown margin of at least 1.3% ~k/k following a reactor trip with the most reactive rod in the fully withdrawn position.

During normal power operation, it is desirable to maintain the rods in alignment with their respective banks. This provides consistency with the assumptions of the safety analyses, maintains symmetric neutron flux and power distribution profiles, provides assurance that peaking factors are within acceptable limits and assures adequate shutdown margin.

2.2 Description of Changes Existing TS 3.1.3.2.1 Actions a and b would be revised to change from a "per bank" basis to a "per group" basis. Existing TS 3.1.3.2.1 Action a.1 would be revised to separate the action associated with verifying rod position for a rod with an inoperable position indicator when the rod has moved in excess of 24 steps in one direction to be consistent with NUREG-1431. This separate action would be labeled as new Action c. New Action c would retain the allowance to verify rod position indirectly with the power distribution monitoring system in addition to using the movable incore detectors and would align the time for verification of position indication to eight hours consistent with Action a.1. Existing TS Action b would be revised to incorporate the changes from TSTF-234-A. Existing Action c would be renumbered to new Action d. The duplicate Action b in the existing TS 3.1.3.2.1 for Salem, Unit No. 1, would be deleted.

The proposed changes to Salem, Unit Nos. 1 and 2, TS 3.1.3.2.1, "Position Indication Systems - Operating," are shown below. Deletions are indicated with a double strike through and additions are indicated with underlining. *

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

1. Determine the position of the non-indicating rod(s) indirectly using the power distribution monitoring system (if power is above 25% RTP) or using the movable incore detectors (if power is less than 25% RTP or the power distribution monitoring system is inoperable) at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months a1=1e witAiR e1=1e f;;ie1;1r after a1=1y FRetie1=1 sf tt;;ie 1=1e1=1 i1=1ei0ati1=1s ree wf;;ii0f;;i &)teeees ~4 steps i1=1 e1=1e eire0tie1=1 si1=10e tf;;ie last eeterFRi1=1atie1=1 sf tf;;ie red's pesitie1=1, or

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

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FRay reFRai1=1 i1=1epera0le fellewi1=1s tt;;ie f;;ie1;1r, witf;;i A0tie1=1 (a) a0ev8 00i1=1s appli0a0l8 freFR tt;;ie erisi1=1al e1=1try tiFRe i1=1te tf;;ie bCO.

b. With two or more analog rod position indicators per group inoperable:

1. Immediately place the control rods in manual control. and

2. Monitor and record Reactor Coolant System Tava once every hour. and 3, Verify the position of the rods with inoperable position indicators jndirectly using the power djstribution monitoring system (if power is above 25% RTP} or using the movable incore detectors <if power is less than 25% RTP or the power distribution monitoring system is inoperable} at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . and

4. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months restore the inoperable rod position indicators to OPERABLE status such that a maximum of one rod position indicator per group is inoperable, or

5. Be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c. When one or more rods with inoperable position indicators haye been moved in excess of 24 steps jn one djrectjon since the last determination of the rod's position; 1, Determjne the position of the non-indicating rod<s} indirectly using the power distribution monitoring system Of power is above 25% RTPl or usjng the movable incore detectors Of power is less than 25% RTP or the power distribution monitoring a

system js inoperable} within hours, or

2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within a hours.

eg. With a maximum of one group demand position indicator per bank inoperable either:

1. 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

2. 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 .

2.3 Regulatory Review The categories of items required to be in the TSs are provided in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36(c). As required by 10 CFR 50.36(c)(2)(i), the TSs will include limiting conditions for operations (LCOs), which are the lowest functional capability or performance levels of equipment required for safe operation of the facility. Per 10 CFR 50.36(c)(2)(i), when an LCO of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the condition can be met. The regulation at 10 CFR 50.36(c)(3) requires TSs to include items in the category of surveillance requirements, which 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 LCOs will be met. Also, 10 CFR 50.36(a)(1) states that a summary statement of the bases or reasons for such specifications, other than those covering administrative controls, shall also be included in the application, but shall not become part of the TSs.

The U.S. Nuclear Regulatory Commission (NRC) staff's guidance for the review of TSs is in Chapter 16, "Technical Specifications," of NUREG-0800, Revision 3, "Standard Review Plan

[SRP] for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition," March 2010 (ADAMS Accession No. ML100351425). As described therein, as part of the regulatory standardization effort, the NRC staff has prepared Standard Technical Specifications (STS) for each of the light-water reactor nuclear designs. The STS for Westinghouse-designed plants is provided in NUREG-1431.

3.0 TECHNICAL EVALUATION

The NRC staff reviewed the proposed changes to the TSs for continued compliance with the requirements of 10 CFR 50.36 and for consistency with conventional terminology and with the format and usage rules embodied in the STS. The NRC staff also reviewed the technical justification for the proposed changes provided in the license amendment request to ensure that

the reasoning was logical, complete, and clearly written as described in SRP Chapter 16.

The proposed change to TS 3.1.3.2.1, Actions a and b, from "per bank" to "per group" is consistent with NUREG-1431. The licensee stated that the current Salem, Unit Nos. 1 and 2, TS were developed based upon NUREG-0452, "Standard Technical Specifications for Westinghouse Pressurized Water Reactors." When the current Westinghouse STS in NUREG-1431 were developed, the NRC revised the action associated with rod position indication from "per bank" to "per group."

LCO 3.1.3.2.1 ensures OPERABILITY of the analog rod position indicators to determine the position of the rods in each group and thereby ensure compliance with group alignment and bank insertion limits. Individual rods in a group all receive the same demand signal to move and should, therefore, all be at the same position indicated by the group step counter for that group.

All RCCAs in a group are paralleled to step simultaneously. Therefore, the staff finds this change acceptable because the basis for entry into the TS Actions for inoperable analog position indicators is appropriately "per group."

The proposed change to TS 3.1.3.2.1, Action a.1, in order to be consistent with NUREG-1431, is to separate the action associated with verifying rod position for a rod with an inoperable position indicator when the rod has moved in excess of 24 steps in one direction. This separate action would be labeled as new Action c. New Action c would maintain the wording of the original Action a to verify rod position indirectly with either the power distribution monitoring system (if power is above 25 percent rated thermal power (RTP)) or the movable incore detectors (if power is less than 25 percent RTP or the power distribution monitoring system is inoperable) and would maintain the time for verification of position indication to eight hours consistent with original Action a.1. Therefore, the staff finds this change acceptable.

The proposed change to TS 3.1.3.2.1, Action b, is to incorporate the changes from TSTF-234-A, which allows for more than one rod position indication (RPI) to be inoperable for a maximum of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , given that other indirect means of monitoring changes in rod position are available.

This provides sufficient time to restore operability while minimizing shutdown transients during the time that the position indication is degraded. The additional time to restore an inoperable RPI is appropriate because the proposed Action b would require that the control rods be under manual control, that reactor coolant system average temperature be monitored and recorded hourly, and that rod position be verified indirectly every eight hours, thereby assuring that the rod alignment and rod insertion LCOs are met. Therefore, the required shutdown margin will be maintained. Given the alternate position monitoring requirement, and other indirect means of monitoring changes in rod position (e.g., alarms on average versus reference temperature deviation), a 24-hour completion time to restore all but one RPI per group provides sufficient time to restore operability while minimizing shutdown transients during the time that the position indication system is degraded. Based on the above, the proposed TS 3.1.3.2.1, Action b, is acceptable.

Existing TS 3.1.3.2.1, Action c, would be renumbered as new Action d. This is an administrative/editorial change and is, therefore, acceptable.

For Salem, Unit No. 1, the proposed change would also eliminate the duplication of TS 3.1.3.2.1, Action b, on pages 3/4 1-19 and 3/4 1-19a. The duplication of TS 3.1.3.2.1,

Action b, was introduced during the issuance of Salem, Unit No. 1, TS Amendment No. 299.

This change is administrative/editorial and is, therefore, acceptable.

The regulations at 10 CFR 50.36 require that TSs include items in specified categories, including LCOs. The proposed changes would modify the LCOs, Conditions, Required Actions, and Completion Times applicable to the control rod position indication systems. The TSs would continue to specify the LCOs and specify the remedial measures to be taken if one of these requirements is not satisfied. The NRG staff finds that the proposed changes to TS 3.1.3.2.1 meet the requirements of 10 CFR 50.36(c)(2) and 50.36(c)(3) because the minimum performance level of equipment needed for safe operation of the facility is contained in the LCO and the appropriate remedial measures are specified if the LCO is not met.

4.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION

The NRC's regulation at 10 CFR 50.92(c) states that the NRG may make a final determination under the procedures in 10 CFR 50.91 that a license amendment involves no significant hazards consideration if operation of the facility, in accordance with the amendment, would not:

(1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

An evaluation of the issue of no significant hazards consideration is presented below:

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

Response: No.

Rod position indication instrumentation is not an accident initiator, providing indication only of the control and shutdown rods positions.

Normal operation, abnormal occurrences and accident analyses assume the rods are at certain positions within the reactor core. The proposed changes modify the time that rod position indication may be inoperable and provide appropriate actions to compensate for that inoperability.

Thus, these changes do not involve a significant increase in the probability of an accident.

Extending the allowed outage time to restore inoperable rod position indicators does not affect the operability of the shutdown or control rods.

With rod position indicators inoperable, the position of non-indicating rods is required to be verified using the movable incore detectors or the power distribution monitoring system. Thus, inoperable rod position indication instrumentation does not involve an increase in the consequences of an accident.

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

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

Response: No.

The proposed change does not alter the design, function, or operation of any plant component and does not install any new or different equipment.

The proposed changes will not impose any new or different requirement or introduce a new accident initiator, accident precursor, or malfunction mechanism.

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

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

Response: No.

Loss of rod position indication does not cause a rod to be misaligned.

With rod position indicators inoperable, the position of non-indicating rods is required to be verified using the movable incore detectors or the power distribution monitoring system. The proposed changes will not affect the ability of the shutdown or control rods to perform their required function.

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

Based on the above evaluation, the NRC staff concludes that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff has made a final determination that no significant hazards consideration is involved for the proposed amendments and that the amendments should be issued as allowed by the criteria contained in 10 CFR 50.91.

5.0 STATE CONSULTATION

In accordance with the Commission's regulations, the New Jersey State official was notified of the proposed issuance of the amendments on March 6, 2018. The State official had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

The amendments change requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. 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

(83 FR 8904; March 1, 2018). Accordingly, the amendments meet 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.

7.0 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) there is reasonable assurance that 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 Contributor: R. Beaton, NRR Date: April 18, 2018

SUBJECT:

SALEM NUCLEAR GENERATING STATION, UNIT NOS. 1 AND 2- ISSUANCE OF AMENDMENT NOS. 324 AND 305 RE: REVISE TECHNICAL SPECIFICATION ACTIONS FOR ROD POSITION INDICATORS (EPID L-2018-LLA-0038) DATED APRIL 18, 2018 DISTRIBUTION:

Public PM File Copy RidsACRS_MailCTR Resource RidsNrrDssStsb Resource RidsNrrDorlLpl1 Resource RidsRgn1 MailCenter Resource RidsNrrDssSrxb Resource RidsNrrLAIBetts Resource RidsNrrPMSalem Resource RBeaton, NRR MChernoff, NRR ADAMS Access1on No.: ML180858198 *b>Y memo **via

. ema,*1 OFFICE DORL/LPL 1/PM DORL/LPL 1/LA DSS/SRXB/BC (A)* DSS/STSB/BC*

NAME JKim IBetts JWhitman VCusumano DATE 04/03/18 04/03/18 03/22/18 03/20/18 OFFICE OGC-NLO** DORL/LPL 1/BC DORL/LPL 1/BC NAME JWachutka JDanna JKim DATE 04/18/18 04/13/18 04/18/18 OFFICIAL RECORD COPY

v t e Letter Sequence Approval
EPID:L-2018-LLA-0038, TSTF-425, TSTF-234 (Approved, Closed)
Initiation Request Acceptance, Acceptance Administration Questions Answers Results Approval Other: ML18045A568, ML18045A655, NRC-2018-0039, Transmittal Letter for Individual FRN for Salem Rod Position Indication
MONTHYEAR2018-02-08 [Table View]

ML18085B198

Text

SUBJECT:

Dear Mr. Sena:

Enclosures:

Attachment:

Attachment:

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

4.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION

5.0 STATE CONSULTATION

6.0 ENVIRONMENTAL CONSIDERATION

7.0 CONCLUSION

SUBJECT:

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