ML15121A589
| ML15121A589 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 06/30/2015 |
| From: | Robert Gladney Plant Licensing Branch 1 |
| To: | Bryan Hanson Exelon Nuclear |
| Gladney R, NRR/DORL/LPLI-2, 415-1022 | |
| References | |
| GL 2008-01, TAC MF4435 | |
| Download: ML15121A589 (15) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 June 30, 2015
SUBJECT:
THREE MILE ISLAND NUCLEAR STATION, UNIT 1 - ISSUANCE OF AMENDMENT REGARDING ADOPTION OF TECHNICAL SPECIFICATIONS TASK FORCE (TSTF) TRAVELER-523, "GENERIC LETTER 2008-01, MANAGING GAS ACCUMULATION" (TAC NO. MF4435)
Dear Mr. Hanson:
The Commission has issued the enclosed Amendment No. 285 to Renewed Facility Operating License No. DPR-50 for the Three Mile Island Nuclear Station, Unit 1 (TMl-1 ), in response to your application dated July 10, 2014, as supplemented by letter dated May 7, 2015.
The amendment modifies the Technical Specification (TS) requirements for TMl-1 to address the concerns discussed in Generic Letter 2008-01, "Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems," dated January 11, 2008.
The TS changes are based on TS Task Force (TSTF) Traveler TSTF-523, Revision 2, "Generic Letter 2008-01, Managing Gas Accumulation," dated February 21, 2013. The Federal Register notice published on January 15, 2014 (79 FR 2700), announced the availability of this TS improvement.
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,. q/};.. .,
-/ifJ/~;-
11 Robert L. Gladney, Project Manager Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-289
Enclosures:
- 1. Amendment No. 285 to DPR-50
- 2. Safety Evaluation cc w/enclosures: Distribution via Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY, LLC DOCKET NO. 50-289 THREE MILE ISLAND NUCLEAR STATION, UNIT NO. 1 AMENDMENT TO F~.CILITY OPERATING LICENSE Amendment No. 285 Renewed License No. DPR-50
- 1. The Nuclear Regulatory Commission (the Commission or NRC) has found that:
A The application for amendment by Exelon Generation Company, LLC (the licensee), dated July 10, 2014, as supplemented by letter dated May 7, 2015, 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; 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 attachment to this license amendment, and paragraph 2.c.(2) of Renewed Facility Operating License No. DPR-50 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 285 are hereby incorporated in the license. The Exelon Generation Company shall operate the facility in accordance with the Technical Specifications.
- 3. This license amendment is effective as of its date of issuance and shall be implemented within 120 days of issuance.
FOR THE NUCLEAR REGULA TORY COMMISSION CCJ,;:A tihtllL Douglas A. Broaddus, Chief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: June 30, 2015
ATTACHMENT TO LICENSE AMENDMENT NO. 285 RENEWED FACILITY OPERATING LICENSE NO. DPR-50 DOCKET NO. 50-289 Replace the following page of the Facility Operating License with the revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.
Remove Insert 4 4 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Remove Insert 4-41 4-41 4-42 4-42 4-43 4-43 4-52a 4-52a
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 285 are hereby incorporated in the license. The Exelon Generation Company shall operate the facility in accordance with the Technical Specifications.
(3) Physical Protection Exelon Generation Company shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822), and the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans 1 , submitted by letter dated May 17, 2006, is entitled: "Three Mile Island Nuclear Station Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan, Revision 3." The set contains Safeguards Information protected under 10 CFR 73.21.
Exelon Generatior:i Company shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The Exelon Generation Company CSP was approved by License Amendment No. 275.
(4) Fire Protection Exelon Generation Company shall implement and maintain in effect all provisions of the Fire Protection Program as described in the Updated FSAR for TMl-1.
Changes may be made to the Fire Protection Program without prior approval by the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire. Temporary changes to specific fire protection features which may be necessary to accomplish maintenance or modifications are acceptable provided that interim compensate measures are implemented.
(5) The licensee shall implement a secondary water chemistry monitoring program to inhibit steam generator tube degradation. This program shall include:
- a. Identification of a sampling schedule for the critical parameters and control points for these parameters;
- b. Identification of the procedures used to measure the values of the critical parameters;
- c. Identification of process sampling points;
- d. Procedure for the recording and management of data; 1
The Training and Qualification Plan and Safeguards Contingency Plan are Appendices to the Security Plan.
Amendment No. 285 Renewed Operating License No. DPR-50
4.5.2 EMERGENCY CORE COOLING SYSTEM Applicability: Applies to periodic testing requirement for emergency core cooling systems.
Objective: To verify that the emergency core cooling systems are operable.
Specification 4.5.2.1 High Pressure Injection
- a. At the frequency specified in the Surveillance Frequency Control Program and following maintenance or modification that affects system flow characteristics, system pumps and system high point vents shall be vented, and a system test shall be conducted to demonstrate that the system is operable.
- b. The test will be considered satisfactory if the valves (MU-V-14A/B
& 16A/B/C/D) have completed their travel and the make-up pumps are running as evidenced by system flow. Minimum acceptable injection flow must be greater than or equal to 431 gpm per HPI pump when pump discharge pressure is 600 psig or greater (the pressure between the pump and flow limiting device) and when the RCS pressure is equal to or less than 600 psig.
- c. Testing which requires HPI flow thru MU-V16A/B/C/D shall be conducted only under either of the following conditions:
- 1) Indicated RCS temperature shall be greaterthan 313°F.
- 2) Head of the Reactor Vessel shall be removed.
- d. At the frequency specified in the Surveillance Frequency Control Program, verify High Pressure Injection locations susceptible to gas accumulation are sufficiently filled with water.
4.5.2.2 Low Pressure Injection
- a. At the frequency specified in the Surveillance Frequency Control Program and following maintenance or modification that affects system flow characteristics, system pumps and high point vents shall be vented, and a system test shall be conducted to demonstrate that the system is operable. The auxiliaries required for low pressure injection are all included in the emergency loading sequence test specified in 4.5.1.
- b. The test will be considered satisfactory if the decay heat pumps have been successfully started and the decay heat injection valves and the decay heat supply valves have completed their travel as evidenced by the control board component operating lights. Flow shall be verified to be equal to or greater than the flow assumed in the Safety Analysis for the single corresponding RCS pressure used in the test.
4-41
- c. When the Decay Heat System is required to be operable, the correct position of DH-V-19A/B shall be verified by observation within four hours of each valve stroking operation or valve maintenance which affects the position indicator.
- d. At the frequency specified in the Surveillance Frequency Control Program, verify Low Pressure Injection locations susceptible to gas accumulation are sufficiently filled with water.
4.5.2.3 Core Flooding
- a. At the frequency specified in the Surveillance Frequency Control Program, a system test shall be conducted to demonstrate proper operation of the system. Verification shall be made that the check and isolation valves in the core cooling flooding tank discharge lines operate properly.
- b. The test will be considered satisfactory if control board indication of core flooding tank level verifies that all valves have opened.
- c. At the frequency specified in the Surveillance Frequency Control Program, verify Core Flooding locations susceptible to gas accumulation are sufficiently filled with water.
4.5.2.4 Component Tests
- a. At the frequency specified in the Surveillance Frequency Control Program, the components required for emergency core cooling will be tested.
- b. The test will be considered satisfactory if the pumps and fans have been successfully started and the valves have completed their travel as evidenced by the control board component operating lights, and a second means of verification, such as: the station computer, verification of pressure/flow, or control board indicating lights initiated by separate limit switch contacts.
Bases The emergency core cooling systems (Reference 1) are the principal reactor safety features in the event of a loss of coolant accident. The removal of heat from the core provided by these systems is designed to limit core damage.
The low pressure injection pumps are tested singularly for operability by opening the borated water storage tank outlet valves and the bypass valves in the borated water storage tank fill line. This allows water to be pumped from the borated water storage tank through each of the injection lines and back to the tank.
The minimum acceptable HPl/LPI flow assures proper flow and flow split between injection legs.
With the reactor shutdown, the valves in each core flooding line are checked for operability by reducing the reactor coolant system pressure until the indicated level in the core flood tanks verify that the check and isolation valves have opened.
ECCS piping and components have the potential to develop voids and pockets of entrained gases. Preventing and managing gas intrusion and accumulation is necessary for proper operation of the ECCS and may also prevent water hammer, pump cavitation, and pumping of noncondensible gas into the reactor vessel.
4-42 Amendment No. a+,e8,..t49,4&7,+e+,22&,2-74, 285
4.5.3 REACTOR BUILDING COOLING AND ISOLATION SYSTEM Applicability Applies to testig of the reactor building cooling and isolation systems.
Objective To verify that the reactor building cooling systems are operable.
Specification 4.5.3.1 System Tests
- a. Reactor Building Spray System
- 1. At the frequency specified in the Surveillance Frequency Control Program and simultaneously with the test of the emergency loading sequence, a Reactor Building 30 psi high pressure test signal will start the spray pump. Except for the spray pump suction valves, all engineered safeguards spray valves will be closed.
Water will be circulated from the borated water storage tank through the reactor building spray pumps and returned through the test line to the borated water storage tank.
The operation of the spray valves will be verified during the component test of the R. B. Cooling and Isolation System.
The test will be considered satisfactory if the spray pumps have been successfully started.
- 2. Compressed air will be introduced into the spray headers to verify each spray nozzle is unobstructed at the frequency specified in the Surveillance Frequency Control Program.
- 3. At the frequency specified in the Surveillance Frequency Control Program, verify Reactor Building Spray locations susceptible to gas accumulation are sufficiently filled with water.
- b. Reactor Building Cooling and Isolation Systems
- 1. At the frequency specified in the Surveillance Frequency Control Program, a system test shall be conducted to demonstrate proper operation of the system.
- 2. The test will be considered satisfactory if measured system flow is greater than accident design flow rate.
4-43 Amendment No. +a+,4-93,~.~.2-74.~. 285
4.9 DECAY HEAT REMOVAL (OHR) CAPABILITY-PERIODIC TESTING (Continued) 4.9.1.6 Acceptance Criteria These tests shall be considered satisfactory if control board indication and visual observation of the equipment demonstrates that all components have operated properly except for the tests required by Specification 4.9.1.1.
4.9.2 RCS Temperature less than or equal to 250 degrees F.*
4.9.2.1 At the frequency specified in the Surveillance Frequency Control Program, verify operability of the means for OHR required by Specification 3.4.2 by observation of console status indication.
N 0 TE----------------------------------------------
Entry into 4.9.2.2. below is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after RCS temperature is less than or equal to 250 degrees F.
4.9.2.2 At the frequency specified in the Surveillance Frequency Control Program, verify required OHR loop locations susceptible to gas accumulation are sufficiently filled with water.
- These requirements supplement the requirements of Specifications 4.5.2.2 and 4.5.4.
The ASME Code specifies requirements and acceptance standards for the testing of nuclear safety related pumps. The EFW Pump test frequency specified by the ASME Code will be sufficient to verify that the turbine-driven and both motor-driven EFW Pumps are operable.
Compliance with the normal acceptance criteria assures that the EFW Pumps are operating as expected. The surveillance requirements ensure that the overall EFW System functional capability is maintained.
Deferral of the requirement to perform IST on the turbine-driven EFW Pump is necessary to assure sufficient OTSG pressure to perform the test using Main Steam.
Periodic verification of the operability of the required means for OHR ensures that sufficient OHR capability will be maintained.
OHR System piping and components have the potential to develop voids and pockets of entrained gases. Preventing and managing gas intrusion and accumulation is necessary for proper operation of the required OHR loop(s) and may also prevent water hammer, pump cavitation, and pumping of noncondensible gas into the reactor vessel.
Selection of OHR System locations susceptible to gas accumulation is based on a review of system design information, including piping and instrumentation drawings, isometric drawings, plan and elevation drawings, and calculations. The design review is supplemented by system walk downs to validate the system high points and to confirm the 4-52a Amendment No. +s,44-9.~.4-72.~,200.2-74, 285
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 285 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-50 EXELON GENERATION COMPANY, LLC THREE MILE ISLAND NUCLEAR STATION. UNIT 1 DOCKET NO. 50-289
1.0 INTRODUCTION
By application dated July 10, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML14191A059), as supplemented by letter dated May 7, 2015 (ADAMS Accession No. ML15127A537) Exelon Generation Company, LLC (the licensee), requested changes to the Technical Specifications (TSs) for Three Mile Island Nuclear Station, Unit 1 (TMl-1). Specifically, the licensee requested to adopt U.S. Nuclear Regulatory Commission (NRC)-approved TSs Task Force (TSTF) Standard Technical Specifications (STS) Change Traveler TSTF-523, Revision 2, "Generic Letter 2008-01, Managing Gas Accumulation" (ADAMS Accession No. ML13053A075), dated February 21, 2013.
The supplement dated May 7, 2015, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register on September 2, 2014 (79 FR 52063).
The proposed change would add surveillance requirements (SRs) related to gas accumulation for the emergency core coolil'lg systems (ECCS). The proposed change would also add new SRs related to gas accumulation for the decay heat removal (OHR) and the reactor building cooling and isolation systems. Bases changes associated with these SRs would also be made.
The licensee stated that the license amendment request is consistent with NRG-approved Traveler TSTF-523. The availability of this TS improvement was announced in the Federal Register on January 15, 2014 (79 FR 2700) as part of the consolidated line item improvement process.
2.0 REGULATORY EVALUATION
2.1 Background Gas accumulation in reactor systems can result in water hammer, pump cavitation, and pumping of non-condensable gas into the reactor vessel. These effects may result in the subject system being unable to perform its specified safety function. The NRC issued Generic Letter (Gl) 2008-01, "Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems," in January 2008 to address the issue of gas accumulation in ECCS, DHR, and containment spray systems (ADAMS Accession No. ML072910759). The industry and NRC staff agreed that a change to the STS and plant-specific TS would be necessary to address some issues discussed in GL 2008-01. TSTF-523 contains changes to the TS SRs and Bases to address some of the concerns in GL 2008-01. The licensee proposed amending the TMl-1 TS using a plant-specific adoption of the TSTF-523 changes.
2.2 Technical Specification Changes The additions of new SRs 4.S.2.1 d, 4.5.2.2d, 4.5.2.3c, 4.5.3.1 a.3, and 4.9.2.2 to TS 4.5.2, "Emergency Core Cooling System," TS 4.5.3, "Reactor Building Cooling and Isolation System,"
and TS 4.9, "Decay Heat Removal (OHR) Capability - Periodic Testing," respectively, were proposed. Associated Bases changes were submitted for information only for the respective SR additions.
2.3 Regulatory Review The regulations in Appendix 6 to Title 1O of the Code of Federal Regulations (10 CFR) Part 50, the TSs, and the licensee quality assurance programs provide operating requirements. For TMl-1, plant-specific principal design criteria provide design requirements and are described in Chapter 1 of the Updated Final Safety Analysis Report (UFSAR).
The construction permit for TMl-1 was issued by the Atomic Energy Commission (AEC) on May 18, 1968, and an operating license was issued on April 19, 1974. The plant design approval for the construction phase was based on the proposed General Design Criterion (GDC) published by the AEC in the Federal Register (32 FR 10213) on July 11, 1967 (hereinafter referred to as "draft GDC"). The AEC published the final rule that added Appendix A to 10 CFR Part 50, "General Design Criteria for Nuclear Power Plants," in the Federal Register (36 FR 3255) on February 20, 1971 (hereinafter referred to as "final GDC" or just "GDC").
Differences between the draft GDC and final GDC included a consolidation from 70 to 64 criteria. In accordance with an NRC staff requirements memorandum from S. J. Chilk to J. M.
Taylor, "SECY-92-223 - Resolution of Deviations Identified During the Systematic Evaluation Program," dated September 18, 1992 (ADAMS Accession No. ML003763736), the Commission decided not to apply the final GDC to plants with construction permits issued prior to May 21, 1971, which includes TMl-1.
The TMl-1 UFSAR, Section 1.4, provides an evaluation of the design bases of TMl-1 against the draft GDC. The UFSAR evaluation of the draft GDC, specifically Criterion 1, "Quality Standards"; Criterion 44, "Emergency Core Cooling Systems Capability"; Criterion 45,
"Inspection of Emergency Core Cooling Systems"; Criterion 46, "Testing of Emergency Core Cooling Systems Components"; Criterion 47, "Testing of Emergency Core Cooling Systems";
Criterion 48, "Testing of Operational Sequence of Emergency Core Cooling Systems";
Criterion 52, "Containment Heat Removal Systems"; Criterion 58, "Inspection of Containment Pressure-Reducing Systems"; Criterion 59, "Testing of Containment Pressure-Reducing Systems"; Criterion 60, "Testing of Containment Spray Systems"; and Criterion 61, "Testing of Operational Sequence of Containment Pressure-Reducing Systems," reflect design requirements similar to those specified in the final GDCs discussed in TSTF-523. The differences do not alter the conclusion that the proposed change is applicable to TMl-1, since the TMl-1 design criteria provide design requirements similar to the GDC requirements applicable to gas accumulation.
Quality assurance criteria provided in 10 CFR Part 50, Appendix B, that apply to gas management in the subject systems include: Criteria Ill, V, XI, XVI, and XVII. Criteria Ill and V require measures to ensure that applicable regulatory requirements and the design basis, as defined in 10 CFR 50.2, "Definitions," and as specified in the license application, are correctly translated into controlled specifications, drawings, procedures, and instructions. Criterion XI requires a test program to ensure that the subject systems will perform satisfactorily in service and requires that test results shall be documented and evaluated to ensure that test requirements have been satisfied. Criterion XVI requires measures to ensure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances, are promptly identified and corrected, and that significant conditions adverse to quality are documented and reported to management. Criterion XVII requires maintenance of records of activities affecting quality.
The NRC regulatory requirements related to the content of the TSs are contained in 10 CFR 50.36(c). The regulations in 10 CFR 50.36 require that the TSs include items in the following categories: (1) saf~ty limits, limiting safety systems settings, and limiting control settings; (2) limiting conditions for operation (LCO); (3) SRs; (4) design features; and (5) administrative controls. SRs 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. TMl-1 TS Section 6 requires the licensee to establish, implement, and maintain written procedures covering the applicable procedures recommended in Appendix A to Regulatory Guide (RG) 1.33, "Quality Assurance Program Requirements (Operation)." Furthermore, Appendix A to RG 1.33 identifies instructions for filling and venting the ECCS and DHR system, as well as for draining and refilling heat exchangers. Additionally, the regulations in 10 CFR 50.46 provide specified ECCS performance criteria.
3.0 TECHNICAL EVALUATION
The SRs the licensee proposed to add are consistent with the TSTF-523, and are appropriate to address the issues discussed in GL 2008-01. The NRC staff compared the proposed changes to the existing SRs, as well as the regulatory requirements of 10 CFR 50.36(c). The proposed SRs must be sufficient to provide reasonable assurance that the necessary quality of systems and components are maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.
The licensee proposed the following TS changes:
(1) Add SR 4.5.2.1 d, which states, "At the frequency specified in the Surveillance Frequency Control Program, verify High Pressure Injection locations susceptible to gas accumulation are sufficiently filled with water."
(2) Add SR 4.5.2.2d, which states, "At the frequency specified in the Surveillance Frequency Control Program, verify Low Pressure Injection locations susceptible to gas accumulation are sufficiently filled with water."
(3) Add SR 4.5.2.3c, which states, "At the frequency specified in the Surveillance Frequency Control Program, verify Core Flooding locations susceptible to gas accumulation are sufficiently filled with water."
(4) Add SR 4.5.3.1 a.3, which states, "At the frequency specified in the Surveillance Frequency Control Program, verify Reactor Building Spray locations susceptible to gas accumulation are sufficiently filled with water."
(5) Add SR 4.9.2.2, which states, "At the frequency specified in the Surveillance Frequency Control Program, verify required DHR loop locations susceptible to gas accumulation are sufficiently filled with water" with a NOTE that states, "Entry into 4.9.2.2 below is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after RCS temperature is less than or equal to 250 degrees F."
The new language for the SRs was developed using licensee responses to GL 2008-01 and the NRC discussion contained in Task Interface Agreement (TIA) 2008-03, "Emergency Core Cooling System (ECCS) Voiding Relative To Compliance With Surveillance Requirements (SR) 3.5.1.1, 3.5.2.3, and 3.5.3.1" (ADAMS Accession No. ML082560209). Many of the GL 2008-01 responses stated that licensees identified system locations susceptible to gas accumulation. In the TIA, the NRC stated that the intent of the TS SRs, which state, "full of water," may be met if the licensee can establish, through an Operability Determination, that there is a reasonable expectation that the system in question will perform its specified safety function.
Therefore, the phrase, "sufficiently filled with water," was recommended for the proposed TS changes. In the TS, "sufficiently filled with water" is understood to mean "sufficiently filled with water to support Operability."' The regulations in 10 CFR 50.36(c)(3) state that one of the purposes of the SR is to verify that the LCO is met. Therefore, the new SR language, "Verify the
[system name] locations susceptible to gas accumulation are sufficiently filled with water," is acceptable since this language will allow the licensee to make a conclusion as to whether or not a system is operable.
The language for the notes that state that the SR does not have to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after RCS temperature is less than or equal to 250 degrees F is acceptable because the note provides a limited time to perform the Surveillance after entering the Applicability of the LCO; however, the requirement to manage gas accumulation is not affected. Licensees must have confidence that the SR can be met or the LCO must be declared not met.
The NRC staff found that the proposed SRs meet the regulatory requirements of 10 CFR 50.36 because they provide assurance that the necessary quality of systems and components will be maintained and that the LCOs will be met. Therefore, the NRC staff finds the proposed change acceptable.
Since TMl-1 has Bases pages printed on the same page as some of its TS pages, some of the revised TS pages also contain Bases revisions associated -with the proposed changes. However, the regulations in 10 CFR 50.36(a)(1) state, in part, that: "A summary statement of the bases or reasons for such specifications ... shall also be included in the application, but shall not become part of the technical specifications."
Therefore, the NRC staff is not making a determination on the acceptability of the Bases revisions.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Pennsylvania State official was notified of the proposed issuance of the amendment. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes 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 surveillance requirements (SRs). The NRC staff has determined that the amendment involves 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. By Federal Register notice 79 FR 52063 dated September 2, 2014, the Commission previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on these findings.
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 amendment.
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) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor: Caroline E. Tilton Date: June 30, 2015
Mr. Bryan C. Hanson President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555
SUBJECT:
THREE MILE ISLAND NUCLEAR STATION, UNIT 1 - ISSUANCE OF AMENDMENT REGARDING ADOPTION OF TECHNICAL SPECIFICATIONS TASK FORCE (TSTF) TRAVELER-523, "GENERIC LETTER 2008-01, MANAGING GAS ACCUMULATION" (TAC NO. MF4435)
Dear Mr. Hanson:
The Commission has issued the enclosed Amendment No. 285 to Renewed Facility Operating License No. DPR-50 for the Three Mile Island Nuclear Station, Unit 1 (TMl-1 ), in response to your application dated July 10, 2014, as supplemented by letter dated May 7, 2015.
The amendment modifies the Technical Specification (TS) requirements for TMl-1 to address the concerns discussed in Generic Letter 2008-01, "Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems," dated January 11, 2008.
The TS changes are based on TS Task Force (TSTF) Traveler TSTF-523, Revision 2, "Generic Letter 2008-01, Managing Gas Accumulation," dated February 21, 2013. The Federal Register notice published on January 15, 2014 (79 FR 2700), announced the availability of this TS improvement.
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, IRA/
Robert L. Gladney, Project Manager Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-289
Enclosures:
- 1. Amendment No. 285 to DPR-50
- 2. Safety Evaluation cc w/enclosures: Distribution via Listserv DISTRIBUTION PUBLIC LPLl-2 R/F RidsNrrDorlDpr Resource RidsACRS_MailCTR Resource RidsNrrDirsltsb Resource RidsNrrDorlLpl1-2 Resource RidsNrrPMThreeMilelsland Resource RidsNrrDssSrxb Resource RidsNrrDssStsb Resource RidsNrrLALRonewicz Resource RidsRgn1 MailCenter Resource ADAMS A ccession No.: ML15121A589 *by e-mail OFFICE DORL/LPL 1-2/PM DORL/LPLl-2/LA DSS/SRXB DSS/STSB NAME RGladnev ABaxter w/comments CJackson* RElliott (PSnvder for)
DATE 6/19/2015 6/15/2015 6/15/2015 6/14/2015 OFFICE OGG DORL/LPL 1-2/BC DORL/LPL 1-2/PM NAME DRoth DBroaddus RGladnev DATE 6/15/2015 6/30/2015 6/30/2015 OFFICIAL RECORD COPY