ML18327A005
| ML18327A005 | |
| Person / Time | |
|---|---|
| Site: | Watts Bar  |
| Issue date: | 01/28/2019 |
| From: | Robert Schaaf Plant Licensing Branch II |
| To: | James Shea, Skaggs M Tennessee Valley Authority |
| Jordan N, | |
| References | |
| EPID L-2018-LLA-0002 | |
| Download: ML18327A005 (16) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 January 28, 2019 Mr. Joseph W. Shea Vice President, Nuclear Regulatory Affairs and Support Services Tennessee Valley Authority 1101 Market Street, LP 4A Chattanooga, TN 37402-2801
SUBJECT:
WATTS BAR NUCLEAR PLANT UNITS 1 AND 2-ISSUANCE OF AMENDMENT TO MODIFY TECHNICAL SPECIFICATION SURVEILLANCE REQUIREMENT 3.6.3.5 FOR CONTAINMENT PURGE ISOLATION VALVES (EPID L-2018-LLA-0002)
Dear Mr. Shea:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 123 to Facility Operating License No. NPF-90 and Amendment No. 24 to Facility Operating License No. NPF-96 for Watts Bar Nuclear Plant (WBN) Units 1 and 2, respectively. These amendments are in response to your application dated January 5, 2018.
The amendments revise Technical Specification 3.6.3, "Containment Isolation Valves,"
Surveillance Requirement (SR) 3.6.3.5 and associated Bases. Specifically, the proposed change would revise SR 3.6.3.5 for the containment purge valves to revise the frequency from "184 days AND Within 92 days after opening the valve" to "In accordance with the Containment Leakage Rate Testing Program."
A copy of the related safety evaluation is also enclosed. Notice of issuance will be included in the Commission's biweekly Federal Register notice.
Docket Nos. 50-390 and 50-391
Enclosures:
- 1. Amendment No. 123 to NPF-90
- 2. Amendment No. 24 to NPF-96
- 3. Safety Evaluation cc:
Listserv Sincerely, Jdii Robert G. Schaaf, Senior Projec Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 TENNESSEE VALLEY AUTHORITY DOCKET NOS. 50-390 WATTS BAR NUCLEAR PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 123 License No. NPF-90
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (TVA) dated January 5, 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; 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.
- 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 Facility Operating License No. NPF-90 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. 123 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. TVA shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 3.
This license amendment is effective as of the date of its issuance, and shall be implemented within 30 days.
Attachment:
Changes to Facility Operating License No. NPF-90 and Technical Specifications Date of Issuance: January 28, 2019 FOR THE NUCLEAR REGULATORY COMMISSION J~/u Undine Shoop, Chief ~
Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
ATTACHMENT TO AMENDMENT NO. 123 WATTS BAR NUCLEAR PLANT, UNIT 1 FACILITY OPERATING LICENSE NO. NPF-90 DOCKET NO. 50-390 Replace page 3 of Facility Operating License No. NPF-90 with the attached revised page 3.
Replace the following page of the Appendix A Technical Specifications with the attached revised page. The revised page is identified by amendment number and contains vertical lines indicating the areas of change.
REMOVE 3.6-13 INSERT 3.6-13 (4)
TVA, 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, instrument calibration, or other activity associated with radioactive apparatus or components; and (5)
TVA, 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 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 TVA is authorized to operate the facility at reactor core power levels not in excess of 3459 megawatts thermal.
(2)
Technical Specifications and Environmental Protection Plan (3)
The Technical Specifications contained in Appendix A as revised through Amendment No. 123 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. TVA shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
Safety Parameter Display System (SPDS) (Section 18.2 of SER Supplements 5 and 15)
Prior to startup following the first refueling outage, TVA shall accomplish the necessary activities, provide acceptable responses, and implement all proposed corrective actions related to having the Watts Bar Unit 1 SPDS operational.
(4)
Vehicle Bomb Control Program (Section 13.6.9 of SSER 20)
During the period of the exemption granted in paragraph 2.D.(3) of this license, in implementing the power ascension phase of the approved initial test program, TVA shall not exceed 50% power until the requirements of 10 CFR 73.55(c)(7) and (8) are fully implemented. TVA shall submit a letter under oath or affirmation when the requirements of 73.55(c)(7) and (8) have been fully implemented.
Facility License No. NPF-90 Amendment No. 123
Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.3.4 SR 3.6.3.5 SR 3.6.3.6 SR 3.6.3.7 Watts Bar-Unit 1 SURVEILLANCE Verify the isolation time of each power operated and each automatic containment isolation valve is within limits.
Perform leakage rate testing for containment purge valves with resilient seals.
Verify each automatic containment isolation valve that is not locked, sealed, or otherwise secured in position, actuates to the isolation position on an actual or simulated actuation signal.
Verify each 24 inch containment lower compartment purge supply and exhaust isolation valve is blocked to restrict the valve from opening > 50°.
3.6-13 FREQUENCY In accordance with the lnservice Testing Program or 92 days In accordance with the Containment Leakage Rate Testing Program 18 months 18 months (continued)
Amendment 123
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 TENNESSEE VALLEY AUTHORITY DOCKET NOS. 50-391 WATTS BAR NUCLEAR PLANT. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 24 License No. NPF-96
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (TVA) dated January 5, 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; 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.
- 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 Facility Operating License No. NPF-96 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. 24 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. TVA shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 3.
This license amendment is effective as of the date of its issuance, and shall be implemented within 30 days.
Attachment:
Changes to Facility Operating License No. NPF-96 and Technical Specifications Date of Issuance: January 28, 2019 FOR THE NUCLEAR REGULATORY COMMISSION
~tf'µ Undine Shoop, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
ATTACHMENT TO AMENDMENT NO. 24 WATTS BAR NUCLEAR PLANT, UNIT 2 FACILITY OPERATING LICENSE NO. NPF-96 DOCKET NO. 50-391 Replace page 3 of Facility Operating License No. NPF-96 with the attached revised page 3.
Replace the following page of the Appendix A Technical Specifications with the attached revised page. The revised page is identified by amendment number and contains vertical lines indicating the areas of change.
REMOVE 3.6-13 INSERT 3.6-13
Unit 2 C.
The 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 TVA is authorized to operate the facility at reactor core power levels not in excess of 3411 megawatts thermal.
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A as revised through Amendment No. 24 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. TVA shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
(3)
TVA shall implement permanent modifications to prevent overtopping of the embankments of the Fort Loudon Dam due to the Probable Maximum Flood by June 30, 2018.
(4)
PAD4TCD may be used to establish core operating limits for Cycles 1 and 2 only. PAD4TCD may not be used to establish core operating limits for subsequent reload cycles.
(5)
By December 31, 2017, the licensee shall report to the NRC that the actions to resolve the issues identified in Bulletin 2012-01, "Design Vulnerability in Electrical Power System," have been implemented.
(6)
The licensee shall maintain in effect the provisions of the physical security plan, security personnel training and qualification plan, and safeguards contingency plan, and all amendments made pursuant to the authority of 10 CFR 50. 90 and 50.54(p ).
(7)
TVA 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 TVA approved CSP was discussed in NUREG-0847, Supplement 28, as amended by changes approved in License Amendment No. 7.
(8)
TVA shall implement and maintain in effect all provisions of the approved fire protection program as described in the Fire Protection Report for the facility, as described in NUREG-0847, Supplement 29, subject to the following provision:
Facility Operating License No. NPF-96 Amendment No. 24
Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE SR 3.6.3.5 Perform leakage rate testing for containment purge valves with resilient seals.
SR 3.6.3.6 Verify each automatic containment isolation valve that is not locked, sealed, or otherwise secured in position, actuates to the isolation position on an actual or simulated actuation signal.
SR 3.6.3. 7 Verify each 24 inch containment lower compartment purge supply and exhaust isolation valve is blocked to restrict the valve from opening > 50°.
SR 3.6.3.8 Verify the combined leakage rate for all shield building bypass leakage paths is~ 0.25 La when pressurized to ~ 15.0 psig.
Watts Bar-Unit 2 3.6-13 FREQUENCY In accordance with the Containment Leakage Rate Testing Program 18 months 18 months In accordance with the Containment Leakage Rate Testing Program Amendment 24
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 123 AND 24 TO FACILITY OPERATING LICENSE NOS. NPF-90 AND NPF-96 TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT. UNITS 1 AND 2 DOCKET NOS. 50-390 AND 50-391
1.0 INTRODUCTION
By letter dated January 5, 2018 (Reference 1 ), Tennessee Valley Authority (TVA or the licensee), proposed changes to Watts Bar Nuclear Plant (WBN), Units 1 and 2 Technical Specifications {TSs) Surveillance Requirement (SR) 3.6.3.5. These changes would revise the containment purge valve leakage test interval from "184 days AND Within 92 days after opening the valve" to "In accordance with the Containment Leakage Rate Testing Program." The WBN Containment Leakage Rate Testing Program is described in TS 5.7.2.19, "Containment Leakage Rate Testing Program." The proposed changes would revise WBN TS SR 3.6.3.5 to replace the currently specified test frequency for leak testing containment purge valves with resilient seals to a frequency established in accordance with TS 5. 7.2.19, "Containment Leakage Rate Testing Program."
2.0 REGULATORY EVALUATION
Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix J, Option B, requires the guidance document used to establish the primary containment leakage rate testing program be identified in a plant's TSs. The WBN Containment Leakage Rate Testing Program is described in TS 5.7.2.19, "Containment Leakage Rate Testing Program." WBN TS 5.7.2.19 references Regulatory Guide (RG) 1.163, "Performance-Based Containment Leak Test Program," dated September 1995 as the guidance document for the program. RG 1.163 endorses Nuclear Electric lnstitute's (NEl's) Topical Report 94-01, Revision 0, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J."
Regulatory position C.2 of RG 1.163 limits the leakage test intervals of containment purge and vent valves to 30 months.
Prior to 1995, the regulations in 10 CFR Part 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors," required that the containment isolation valves (Type C), including containment purge and vent valves, and other penetration barriers
{Type B) be subjected to local leakage rate tests at every refueling outage not to exceed 2 years, and that an integrated leakage rate test (ILRT, or Type A) be performed three times in 10 years. Compliance with 10 CFR Part 50, Appendix J, provided assurance that the leakage rate of the containment, including those systems and components which penetrate the containment, does not exceed the allowable leakage rate specified in the TSs. The allowable leakage rate was established so that the leakage rate assumed in the safety (dose) analyses is not exceeded. The regulations in 10 CFR Part 50, Appendix J, did not contain any special requirements for containment purge and vent valves.
However, in the 1970s, early commercial nuclear plant operating experience had shown that containment purge and vent valves, particularly those of large diameter with resilient seals were more susceptible than other containment isolation valves to degradation caused by environmental factors (such as temperature extremes and changes in humidity and barometric pressure) and mechanical factors (such as wear and tear and hardening of resilient seals due to aging and exposure to radiation). Not only could this degradation cause high and rapidly increasing leakage rates, but the radiological consequences of such leaks were more significant than for other valves because of the large diameters of most containment purge and vent valves and the direct connection they provided between the primary containment atmosphere and the outside environment.
For this reason, the U.S. Nuclear Regulatory Commission (NRC) staff determined that containment purge and vent valves were, as a class, a special problem in terms of leakage rate. The NRC established Generic Issue *s-20, "Containment Leakage Due to Seal Deterioration" to study the problem and propose a regulatory resolution. As part of the resolution of Generic Issue B-20 (also known as Multi-Plant Action MPA-B020), the NRC staff required increased frequency of local leakage rate testing of containment purge and vent valves, over and above the frequency required by Appendix J (additional background may be found in Inspection and Enforcement Circular 77-11, "Leakage of Containment Isolation Valves with Resilient Seats," issued September 6, 1977). The purpose was to limit the length of time during which the valves might have an excessive leakage potential and, thereby, make it more likely that a licensee would catch and correct advancing degradation before it could challenge primary containment operability.
Although there was some variation, a typical test frequency was every 6 months for "passive" valves (those not normally opened during plant operation), and every 3 months for "active" valves (those opened during plant operation). However, the NRC staff implemented the increased testing frequencies through individual plant TS without changing the requirements in 1 O CFR Part 50, Appendix J. The current requirements in WBN's TSs for the test interval is 184 days for the valves that have not been opened, and 92 days for valves that have been opened.
These reflect the test frequencies mentioned above.
Subsequent to the problems observed in the 1970s, the industry made considerable progress in correcting the deficiencies of containment purge and vent valves with resilient seals.
Improved seat seal materials, quality control, and modifications of equipment and environmental conditions have largely corrected poor valve leakage performance at most plants. After careful evaluation of the improvements made in containment leakage potential, by the nuclear industry, the NRC revised Appendix J in 1995 to add a performance-based option for leakage testing. This was identified as Option B, whereas the original requirements were subsequently referred to as Option A. The NRC staff also issued RG 1.163, which endorses NEI guidance document NEI 94-01, Revision 0, dated July 26, 1995. NEI 94-01 identified methods acceptable to the NRC staff for complying with Option B. The NRC staff specified that the acceptable use of NEI 94-01 for complying with Option B was subject to four different exceptions that are detailed in RG 1.163. With regard to the exceptions specified in RG 1.163, a licensee could use NEI 94-01, Revision 0, to extend leak test intervals in the following manner:
Type A tests - a maximum of once in 10 years based on two consecutive successful tests and consideration of performance factors described in NEI 94-01, Revision 0.
Type B tests - (local leakage rate tests of containment penetrations whose design incorporates resilient seals, gaskets, or sealant compounds, piping penetrations fitted with expansion bellows, and electrical penetrations) a maximum interval of 120 months based in large part on satisfactory leakage test history.
Type C tests - (local leakage rate tests of containment isolation valves) a maximum interval of 60 months based in large part on satisfactory leakage test history.
Concerning containment purge and vent valves, the NRC staff, in RG 1.163, did not endorse extending the test interval. RG 1.163 endorsed the maximum test interval increase for containment purge and vent valves from 24 months (10 CFR Part 50, Appendix J Option A requirement) to 30 months. This determination took into consideration the past poor operating experience and the safety significance of the associated large diameter and direct containment atmosphere to outside environment pathways.
The NRC staff has granted similar requests pertaining to the containment purge and vent valves, including those from Catawba Nuclear Station (ML051730234), Grand Gulf Nuclear Station (ML021490101 ), River Bend Station (ML070920450), McGuire Nuclear Station (ML022540102), and Waterford Steam Electric Station (ML071290447). However, those TS changes were granted on the basis of plant-specific historical leakage rate testing results showing that their primary containment purge and vent valves have had consistently good leakage performance and are, thus, unlikely to experience enough degradation to jeopardize containment operability during the longer test interval permitted by 10 CFR Part 50, Appendix J, Option B, and RG 1.163.
3.0 TECHNICAL EVALUATION
3.1 Evaluation of the Proposed Technical Specification Changes The license amendment request (LAR) describes the WBN Reactor Building Purge Ventilation system function as supplying outside air into the containment for ventilation and cooling or heating, to equalize internal and external pressures and to reduce the concentration of noble gases within containment prior to and during personnel access. The supply and exhaust lines each contain two isolation valves. Because of their large size and their exposure to higher containment pressure during accident conditions, the containment lower compartment purge isolation valves are physically restricted to less than or equal to 50 degrees open. Furthermore, because the valves used in the Reactor Building Purge Ventilation system are designed to meet the requirements for automatic containment isolation valves, these valves may be opened as needed in Modes 1, 2, 3, and 4.
WBN Unit 1 and 2 Containment Purge Systems each contain ten containment penetrations with each penetration containing redundant containment isolation valves. The valves are 8 inch, 12 inch, or 24 inch diameter pneumatically operated butterfly valves with resilient seals. The Containment Purge Air System is placed in service as needed to conduct containment-purging operations in Modes 1 through 4 and during cold shutdown and refueling activities.
The containment isolation valves form part of the containment pressure boundary and provide a means for fluid penetrations not serving accident consequence limiting systems to be provided with two isolation barriers that are closed on a containment isolation signal or which are normally closed. Two barriers in series are provided for each penetration so that no single credible failure or malfunction of an active component can result in a loss of isolation or leakage that exceeds limits assumed in the safety analyses. Automatic isolation signals are produced during accident conditions. Containment Phase "A" isolation occurs upon receipt of a safety injection signal. The Phase "A" isolation signal isolates nonessential process lines in order to minimize leakage of fission product radioactivity. Containment Phase "B" isolation occurs upon receipt of a containment pressure High-High signal and isolates the remaining process lines, except systems required for accident mitigation. In addition to the isolation signals listed above, the purge and exhaust valves receive an isolation signal on a containment high radiation condition. As a result, the containment isolation valves (and blind flanges) help ensure that the containment atmosphere will be isolated from the environment in the event of a release of fission product radioactivity into the containment atmosphere as a result of a Design-Basis Accident. The operability requirements for the containment isolation valves help ensure that containment is isolated and assumed leak tightness established within the time limits assumed in the safety analyses. Therefore, the operability requirements provide assurance that the containment function assumed in the safety analyses will be maintained.
The LAR contained a Table 1, WBN Unit 1 Containment Purge Air System Containment Isolation Valves Leakage History, and a Table 2, WBN Unit 2 Containment Purge Air System Containment Isolation Valves Leakage History, which provided the previous 10 years of as-found leak rate history of the Unit 1 valves and the available test results for Unit 2 valves since plant startup in 2016. The LAR indicates that the valves are essentially identical between Unit 1 and Unit 2 and that with the same operating conditions and maintenance and testing program, the testing results are expected to remain closely comparable between the units. All the leak test results shown in Table 1 and 2 were acceptable with no apparent trend or pattern and the LAR indicates that no maintenance had been performed during this time. The LAR also included a Table 4, Comparison of the Containment Purge Air System Valve Leakage to Total Combined Leakage, for both the Unit 1 and Unit 2 valves, which showed their combined contribution to the Type Band C test totals to be about 3 percent or less of the as-found 0.6 La (maximum allowable containment leakage rate) performance criterion specified in TS 5.7.2.19.
LAR Table 5, Average Number of Times the Containment Purge Air System Valves have been Cycled per Year, showed a range of 8 to 30, which suggests normal wear and tear would not significantly increase leakage potential over the requested longer test interval. In addition, a number of these cycles occurred during performance of in-service testing program in accordance with TVA surveillance Instructions 1/2-Sl-30-901-A, "Valve Full Stroke Exercising During Plant Operation - Ventilation (Train A)," and 1/2-Sl-30-901-B, "Valve Full Stroke Exercising During Plant Operation - Ventilation (Train B). Although not directly indicative of leakage potential, these tests provide more frequent periodic checks of overall valve condition.
3.2 Technical Conclusion Based on the licensee's information contained in the LAR, the NRC staff has determined that WBN purge system valves have exhibited good low leakage potential performance over the previous 10 years on Unit 1. Similar results were determined for Unit 2 but during a lesser period of operation. However, the valves are of essentially the same make, model, and size, and are in the same operating environments under the same maintenance and testing programs. This suggest that it is likely that similar results will be achieved for Unit 2. Therefore, the NRC staff finds the licensee's request, to revise TS SR 3.6.3.5 to require the test frequency to be "in accordance with the Primary Containment Leakage Rate Testing Program" that is specified in TS 5. 7.2.19, to be acceptable. Furthermore, the NRC staff determined that it will provide reasonable assurance that containment leakage potential will continue to be adequately monitored to allow for timely corrective actions that may be needed. With this change, the maximum leakage test interval for the containment purge system valves will be 30 months.
On the basis of its review, the NRC staff concludes that the proposed amendments are consistent with the guidance discussed in Section 2.0 of this safety evaluation and meet the regulatory requirements set forth in Section 2.0 of this safety evaluation, and are, therefore, acceptable.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Tennessee State official was notified of the proposed issuance of the amendments on November 21, 2018. 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 change 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 previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding published in the Federal Register on March 13, 2018 (83 FR 10924). 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.
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.
7.0 REFERENCE
- 1. Letter dated January 5, 2018, from Tennessee Valley Authority to U.S. Nuclear Regulatory Commission, Application to Modify Technical Specification (TS) 3.6.3, "Containment Isolation Valves," Surveillance Requirement 3.6.3.5 (WBN-TS-17-01 ), Agencywide Documents Access and Management System Accession Number ML18008A257.
Principal Contributors: Jerome 0. Bettle Tarico G. Sweat Robert J. Wolfgang Date: January 28, 2019
- B e-mail
- B SE memo OFFICE NRR/DORL/LPL2-2/PM NRR/DORL/LPL2-2/LA NRR/DSS/SCPB/BC**
NAME NJordan BClayton SAnderson DATE 12/10/2018 12/10/2018 10/3/2018 OFFICE NRR/DSS/STSB/BC*
NRR/DE/EMIB/BC*
NAME VCusamano SBailey KGamin DATE 11/29/2018 11/19/2018 1/2/2019 OFFICE NRR/DORL/LPL2 2/BC NRR/DORL/LPL2 2/PM NAME UShoop RSchaaf DATE 1/28/19 1/28/19 I
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