ML18221A107
| ML18221A107 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 09/06/2018 |
| From: | Klos L Plant Licensing Branch IV |
| To: | Sawatzke B Energy Northwest |
| Klos L, NRR/DORL/LPL4-1, 301-415-5136 | |
| References | |
| EPID L-2017-LLA-0433, TSTF-551, Rev 3 | |
| Download: ML18221A107 (17) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. Bradley J. Sawatzke Chief Executive Officer Energy Northwest 76 North Power Plant Loop P.O. Box 968 (Mail Drop 1023)
Richland, WA 99352-0968 September 6, 2018
SUBJECT:
COLUMBIA GENERATING STATION-ISSUANCE OF AMENDMENT RE:
ADOPTION OF TECHNICAL SPECIFICATIONS TASK FORCE (TSTF),
TRAVELER TSTF-551, REVISION 3, "REVISE SECONDARY CONTAINMENT SURVEILLANCE REQUIREMENTS" (EPID L-2017-LLA-0433)
Dear Mr. Sawatzke:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 250 to Renewed Facility Operating License No. NPF-21 for the Columbia Generating Station. The amendment consists of changes to the technical specifications (TS) in response to your application dated December 12, 2017.
The amendment revises TS 3.6.4.1, "Secondary Containment," Surveillance Requirement (SR) 3.6.4.1.1, to allow the secondary containment vacuum limit to not be met provided the standby gas treatment system remains capable of establishing the required secondary containment vacuum. In addition, SR 3.6.4.1.3, is revised to allow for the temporary opening of the inner and outer doors of secondary containment for the purpose of entry and exit. Also, SR 3.6.4.1.4 is revised to add an initialism. These changes are consistent with Technical Specifications Task Force (TSTF) Traveler TSTF-551, Revision 3, "Revise Secondary Containment Surveillance Requirements," dated October 3, 2016.
A copy of the related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
Docket No. 50-397
Enclosures:
- 1. Amendment No. 250 to NPF-21
- 2. Safety Evaluation cc: Listserv Sincerely, L. JJ:troj (ger Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 ENERGY NORTHWEST DOCKET NO. 50-397 COLUMBIA GENERA TING STATION AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 250 License No. NPF-21
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Energy Northwest (the licensee), dated December 12, 2017, 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 Renewed Facility Operating License No. NPF-21 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. 250 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 3.
The license amendment is effective as of its date of issuance and shall be implemented within 60 days from the date of issuance/
I I
Attachment:
Changes to the Renewed Facility Operating License No. NPF-21 and Technical Specifications FOR THlt NGCLEAR REGULATORY COMMISSION
\\P/v{
~~- Pascf relli. Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: September 6, 2018
ATTACHMENT TO LICENSE AMENDMENT NO. 250 COLUMBIA GENERA TING STATION RENEWED FACILITY OPERATING LICENSE NO. NPF-21 DOCKET NO. 50-397 Replace the following pages of the Renewed Facility Operating License No. NPF-21 and Appendix A, Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
Renewed Facility Operating License REMOVE INSERT Technical Specification REMOVE 3.6.4.1-2 INSERT 3.6.4.1-2 (2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 250 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- a. For Surveillance Requirements (SRs) not previously performed by existing SRs or other plant tests, the requirement will be considered met on the implementation date and the next required test will be at the interval specified in the Technical Specifications as revised in Amendment No. 149.
(3)
Deleted.
( 4)
Deleted.
(5)
Deleted.
(6)
Deleted.
(7)
Deleted.
(8)
Deleted.
(9)
Deleted.
(10)
Deleted.
( 11)
Shield Wall Deferral (Section 12.3.2, SSER #4, License Amendment #7)
The licensee shall complete construction of the deferred shield walls and window as identified in Attachment 3, as amended by this license amendment.
(12)
Deleted.
(13)
Deleted.
- The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed.
Renewed License No. NPF-21 Amendment No. 250
Secondary Containment 3.6.4.1 SURVEILLANCE REQUIREMENTS SR 3.6.4.1.1 SR 3.6.4.1.2 SR 3.6.4.1.3 SR 3.6.4.1.4 SR 3.6.4.1.5 SURVEILLANCE
NOTE-------------------------------
Not required to be met for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> if analysis demonstrates one standby gas treatment (SGT) subsystem is capable of establishing the required secondary containment vacuum.
Verify secondary containment vacuum is 2 0.25 inch of vacuum water gauge.
Verify all secondary containment equipment hatches are closed and sealed.
Verify each secondary containment access inner door or each secondary containment access outer door in each access opening is closed, except when the access opening is being used for entry and exit.
Verify each SGT subsystem will draw down the secondary containment to 2 0.25 inch of vacuum water gauge in s 120 seconds.
Verify each SGT subsystem can maintain 2 0.25 inch of vacuum water gauge in the secondary containment for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at a flow rate s 2240 cfm.
FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Columbia Generating Station 3.6.4.1-2 Amendment No. 469,499~~ 250
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 250 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-21 ENERGY NORTHWEST COLUMBIA GENERA TING STATION DOCKET NO. 50-397
1.0 INTRODUCTION
By application dated December 12, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML173468280), Energy Northwest (the licensee) requested changes to the technical specifications (TSs) for the Columbia Generating Station (Columbia).
The amendment would adopt Technical Specifications Task Force (TSTF} traveler, TSTF-551, Revision 3: "Revise Secondary Containment Surveillance Requirements," dated October 3, 2016 (ADAMS Accession No. ML16277A226). The U.S. Nuclear Regulatory Commission (NRC or the Commission) approved the traveler on September 21, 2017 (ADAMS Package Accession No. ML17236A365).
The proposed changes would allow the secondary containment vacuum limit to not be met provided that the standby gas treatment (SGT) system remains capable of establishing the required secondary containment vacuum and revises the TSs to permit secondary containment access openings to be open to permit entry and exit.
2.0 REGULATORY EVALUATION
2.1 SYSTEM DESCRIPTION The secondary containment is a structure that encloses the primary containment, including components that may contain primary system fluid. The safety function of the secondary containment is to contain, dilute, and hold up fission products that may leak from primary containment following a design-basis accident (OBA) to ensure the control room operator and offsite doses are within the regulatory limits. There is no redundant train or system that can perform the secondary containment function should the secondary containment be inoperable.
The secondary containment boundary is the combination of walls, floor, roof, ducting, doors, hatches, penetrations and equipment that physically form the secondary containment.
Routinely used secondary containment access openings contain at least one inner and one outer door in an airlock configuration. In some cases, secondary containment access openings are shared such that there are multiple inner or outer doors. All secondary containment access doors are normally kept closed, except when the access opening is being used for entry and exit of personnel, equipment, or material.
Secondary containment operability is based on its ability to contain, dilute, and hold up fission products that may leak from primary containment following a DBA. To prevent ground level exfiltration of radioactive material while allowing the secondary containment to be designed as a mostly conventional structure, the secondary containment requires support systems to maintain the pressure at less than atmospheric pressure. During normal operation, nonsafety-related systems are used to maintain the secondary containment at a slight negative pressure to ensure any leakage is into the building and that any secondary containment atmosphere exiting is via a pathway monitored for radioactive material. However, during normal operation it is possible for the secondary containment vacuum to be momentarily less than the required vacuum for a number of reasons, such as during wind gusts or swapping of the normal ventilation subsystems.
During emergency conditions, the SGT system is designed to be capable of drawing down the secondary containment to a required vacuum within a prescribed time and continue to maintain the negative pressure as assumed in the accident analysis. For Columbia, the SGT must be able to establish the required vacuum within 20 minutes. The leak tightness of the secondary containment together with the SGT system ensure that radioactive material is either contained in the secondary containment or filtered through the SGT system filter trains before being discharged to the outside environment via the elevated release point.
2.2 Proposed Technical Specification Changes The proposed changes to the Columbia TS would allow the secondary containment vacuum limit to not be met provided the SGT system remains capable of establishing the required secondary containment vacuum. The proposed changes would also allow for the temporary opening of the inner and outer doors of secondary containment for the purpose of entry and exit (i.e., normal opening and prompt closure of a door for transit).
2.2.1 Revision to Surveillance Requirement 3.6.4.1.1 Surveillance requirement (SR) 3.6.4.1.1 requires verification that secondary containment vacuum is greater than or equal to (~) 0.25 inch of vacuum water gauge. This SR would be modified by a note that states:
Not required to be met for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> if analysis demonstrates one standby gas treatment (SGT) subsystem is capable of establishing the required secondary containment vacuum.
2.2.2 Revision to Surveillance Requirement 3.6.4.1.3 SR 3.6.4.1.3 requires verification that one secondary containmentaccess door in each access opening is closed. This SR would be modified by adding the following phrase to the end of the SR statement, "... except when the access opening is being used for entry and exit."
2.2.3 Revision to Surveillance Requirement 3.6.4.1.4 An editorial change would be made to SR 3.6.4.1.4 in which the words "standby gas treatment" are replaced with the initialism "SGT."
2.3 Regulatory Requirements and Guidance The regulation under Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36(a)(1 ),
requires an applicant for an operating license to include in the application proposed TS in accordance with the requirements of 10 CFR 50.36. The applicant must include in the application, "[a] summary statement of the bases or reasons for such specifications, other than those covering administrative controls.... " However, per 10 CFR 50.36(a)(1), these TS bases "shall not become part of the technical specifications."
Additionally, 10 CFR 50.36(b) requires:
Each license authorizing operation of a... utilization facility... will include technical specifications. The technical specifications will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to [10 CFR] 50.34 ["Contents of applications; technical information"]. The Commission may include such additional technical specifications as the Commission finds appropriate.
The categories of items required to be in the TSs are provided in 10 CFR 50.36(c). As required by 10 CFR 50.36(c)(2)(i), the TSs will include limiting conditions for operation (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), "[w]hen a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met."
The regulation at 10 CFR 50.36(c)(3) requires TSs to include items in the category of SRs, 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 limiting conditions for operation will be met."
The NRC staff's guidance for review of TSs is in Chapter 16, "Technical Specifications," of NUREG-0800, Revision 3, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition" (SRP), dated March 2010 (ADAMS Accession No. ML100351425).
NUREG-0800, SRP Section 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms," Revision 0, dated July 2000 (ADAMS Accession No. ML003734190), provides guidance to the NRC staff for the review of alternative source term (AST) amendment requests.
It states that the NRC reviewer should evaluate the proposed change against the guidance in Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," Revision 0, dated July 2000 (ADAMS Accession No. ML003716792).
RG 1.183 provides an acceptable methodology for analyzing the radiological consequences of several design basis accidents to show compliance with 10 CFR 50.67, "Accident source term."
RG 1.183 provides guidance to licensees on acceptable application of AST (also known as the accident source term) submittals, including acceptable radiological analysis assumptions for use in conjunction with the accepted AST.
The regulations under 10 CFR 50.67(b )(2) states:
The NRC may issue the amendment only if the applicant's analysis demonstrates with reasonable assurance that:
(i)
An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 0.25 Sv [Sievert] (25 rem [roentgen equivalent man])1 total effective dose equivalent (TEDE).
(ii)
An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), would not receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE).
(iii)
Adequate radiation protection is provided to permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) total effective dose equivalent (TEDE) for the duration of the accident.
3.0 TECHNICAL EVALUATION
The NRC staff evaluated the licensee's application to determine if the proposed changes are consistent with the guidance, regulations, and licensing information discussed in Section 2.3 of this safety evaluation (SE) and the approved traveler TSTF-551, Revision 3. In determining whether an amendment to a license will be issued, the NRC is guided by the considerations that govern the issuance of initial licenses to the extent applicable and appropriate. In making its determination as to whether to amend the license, the NRC staff considered those regulatory requirements that are automatically conditions of the license through 1 O CFR 50.54, "Conditions of licenses."
3.1 Proposed Change to Surveillance Requirement 3.6.4.1.1 A note is being added to SR 3.6.4.1.1. The note would allow the SR to not be met for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> if an analysis demonstrates that one SGT subsystem is capable of establishing the required secondary containment vacuum. During normal operation, conditions may occur that result in SR 3.6.4.1.1 not being met for short durations. For example, wind gusts that lower external pressure or loss of the normal ventilation system that maintains secondary containment vacuum may affect secondary containment vacuum. These conditions may not be indicative of 1 The use of 0.25 Sv (25 rem) TEDE is not intended to imply that this value constitutes an acceptable limit for emergency doses to the public under accident conditions. Rather, this 0.25 Sv (25 rem) TEDE value has been stated in this section as a reference value, which can be used in the evaluation of proposed design basis changes with respect to potential reactor accidents of exceedingly low probability of occurrence and low risk of public exposure to radiation.
degradations of the secondary containment boundary or of the ability of the SGT system to perform its specified safety function.
The proposed note would provide an allowance for the licensee to confirm secondary containment operability by confirming that one SGT subsystem is capable of performing its specified safety function. This confirmation is necessary to apply the exception to meeting the SR acceptance criterion. While the duration of these occurrences is anticipated to be very brief, the allowance is permitted for a maximum of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, which is consistent with the time permitted for secondary containment to be inoperable per Condition A of LCO 3.6.4.1.
The NRC staff has evaluated the impact of this note on the licensee's design basis radiological consequence analyses to ensure that the proposed change will not result in an increase in the dose consequences and that the resulting calculated doses remain within the current radiological consequence analyses.
The proposed addition of the note to SR 3.6.4.1.1 does not change the TS requirement to meet SR 3.6.4.1.4 and SR 3.6.4.1.5. The SR 3.6.4.1.4 requires verification that the secondary containment can be drawn down to ~ 0.25 inch of vacuum water gauge in less than or equal to (S) 120 seconds using one SGT subsystem. The SR 3.6.4.1.5 requires verification that the secondary containment can be maintained ~ 0.25 inch of vacuum water gauge for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> using one SGT subsystem at a flow rate s 2240 cubic feet per minute. In addition, TS LCO 3.6.4.3, "Standby Gas Treatment (SGT) System," must be met; otherwise the licensee shall shut down the reactor or follow any remedial action permitted by TSs until the condition can be met.
As discussed above, secondary containment operability is based on its ability to contain, dilute, and hold up fission products that may leak from primary containment following a DBA. To prevent ground level exfiltration of radioactive material the secondary containment pressure must be maintained at a pressure that is less than atmospheric pressure. The secondary containment requires support systems to maintain the control volume pressure less than atmospheric pressure. Following an accident, the SGT system ensures the secondary containment pressure is less than the external atmospheric pressure. During normal operation, nonsafety-related systems are used to maintain the secondary containment at a negative pressure. However, during normal operation it is possible for the secondary containment vacuum to be momentarily less than the required vacuum for a number of reasons. These conditions may not be indicative of degradations of the secondary containment boundary or of the ability of the SGT system to perform its specified safety function. Since the licensee meets the requirements of SR 3.6.4.1.4, SR 3.6.4.1.5, meets the LCO or is following the Actions of TS LCO 3.6.4.3, and the licensee's analysis confirms secondary containment operability by confirming that one SGT subsystem is capable of performing its specified safety function, then there is reasonable assurance that the secondary containment and SGT subsystem will maintain the vacuum requirements during a DBA.
Therefore, the NRC staff has determined that if the conditions do not affect (1) the ability to maintain the secondary containment pressure during an accident, at a vacuum that is consistent with the accident analyses, and (2) the time assumed in the accident analyses to draw down the secondary containment pressure, then the secondary containment can perform its safety function and may be considered TS operable. This is evident by being able to successfully perform and meet SR 3.6.4.1.4 and SR 3.6.4.1.5. These SRs require the SGT system to establish and maintain the required vacuum in the secondary containment as assumed in the accident analyses.
Furthermore, because the specified safety functions of the secondary containment and SGT subsystem can be performed in the time assumed in the licensee's accident analysis, then the fission products that bypass or leak from primary containment, or are released from the reactor coolant pressure boundary components located in secondary containment prior to release to the environment, will be contained and processed as assumed in the licensee's design basis radiological consequence dose analyses. The NRC staff finds that the proposed change does not affect the current radiological consequence analyses and concludes that the proposed change is acceptable with respect to the radiological consequences of DBAs.
3.2 Proposed Change to Surveillance Requirement 3.6.4.1.3 The NRC staff review was limited to the licensee's request to provide an allowance for the brief, inadvertent, simultaneous opening of redundant secondary containment access doors during normal entry and exit conditions. Planned activities that could result in the simultaneous opening of redundant secondary containment access openings, such as maintenance of a secondary containment personnel access door or movement of large equipment through the openings that would take longer than the normal transit time, will be considered outside the scope of the NRC staff's review.
The NRC staff reviewed the proposed changes to SR 3.6.4.1.3. The NRC staff determined that the SR continues to provide appropriate confirmation that secondary containment boundary doors are properly positioned and capable of performing their function in preserving the secondary containment boundary. The NRC staff determined that the SRs continue to appropriately verify the operability of the secondary containment and provide assurance that the necessary quality of systems and components are maintained in accordance with 10 CFR 50.36(c)(3).
Additionally, the NRC staff evaluated the impact of modifying the licensee's TS to allow secondary containment access openings to be open for entry and exit on the licensee's design basis radiological consequence dose analyses to ensure that the modification will not result in an increase in the radiation dose consequences and that the resulting calculated radiation doses will remain within the design criteria specified in the current radiological consequence analyses. The NRC staff review of these DBAs determined that there is one OBA that takes credit for the secondary containment, and is possibly impacted by the brief, inadvertent, simultaneous opening of both an inner and outer access door during normal entry and exit conditions and the loss-of-coolant accident (LOCA).
3.2.1 LOCA Following a LOCA, the secondary containment structure is maintained at a negative pressure ensuring that leakage from primary containment to secondary containment can be collected and filtered prior to release to the environment. The SGT system performs the function of maintaining a negative pressure within the secondary containment, as well as collecting and filtering the leakage from primary containment. The licensee credits the SGT system for mitigation of the radiological releases from the secondary containment. In the LOCA analysis, the secondary containment draw down analysis assumes that SGT system can draw down the secondary containment within 20 minutes. TS SR 3.6.4.1.4 requires one SGT subsystem to draw down the secondary containment, to.:: 0.25 inches of vacuum water gauge in a maximum allowable time of 120 seconds.
Conservatively, the DBA LOCA radiological consequence analysis in the Columbia Final Safety Analysis Report (FSAR), Amendment 64, Chapter 15, Section 15.6.5, "Loss-of-Coolant Accidents (Resulting from Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary)- Inside Containment" (ADAMS Accession No. ML17355A670), assumes that following the start of a DBA LOCA the secondary containment pressure of 0.25 inches of vacuum water gauge is achieved at approximately 20 minutes. The licensee assumes that releases into the secondary containment prior to the 20-minute draw down time leak directly to the environment as a ground level release with no filtration. After the assumed 20-minute draw down these releases are filtered by the SGT system and released via the SGT system exhaust vent.
Based on this information, the NRC staff concludes that the licensee's DBA LOCA analysis has sufficient conservatism by assuming a draw down time of 20 minutes from the start of the DBA LOCA. Margin exists to ensure that the secondary containment can be reestablished during a brief, inadvertent, simultaneous opening of the inner and outer doors, and there is reasonable assurance that a failure of a safety system needed to control the release of radioactive material to the environment will not occur. The brief, inadvertent, simultaneous opening of the secondary containment access doors does not impact the design bases and will not result in an increase in any onsite or offsite dose.
Based on the above discussion, the NRC staff finds that the licensee's proposed change to the TSs does not impact the licensee's design basis LOCA radiological consequence analysis and will not result in an increase in any onsite or offsite dose. Therefore, the NRC staff concludes that this change is acceptable with respect to the radiological consequences of the DBAs.
The licensee was approved for AST methodology and the radiological dose consequences analyses for DBAs via License Amendment No. 199 for Columbia, dated November 27, 2006 (ADAMS Accession No. ML062610440). The NRC staff reviewed the impact of the proposed changes to the Columbia TSs, on all DBAs currently analyzed in the Columbia FSAR Chapter 15 that could have the potential for significant dose consequences.
The NRC staff's review was limited to the licensee's request to provide an allowance for the brief, inadvertent, simultaneous opening of redundant secondary containment access doors during normal entry and exit conditions. Planned activities that could result in the simultaneous opening of redundant secondary containment access openings, such as maintenance of a secondary containment personnel access door or movement of large equipment through the openings that would take longer than the normal transit time, will be considered outside the scope of the NRC staff's review.
3.2.2 Conclusion As described above, the NRC staff reviewed the technical basis provided by the licensee to assess the radiological impacts of the changes to the secondary containment in the licensee's TSs. The NRC staff finds that the licensee's proposed change to SR 3.6.4.1.3 is consistent with regulatory requirements and guidance identified in Section 2.3 of this SE. The NRC staff finds, with reasonable assurance that the licensee's proposed change to the TSs will continue to comply with these criteria and that that the licensee's estimates of the dose consequences of a design basis LOCA will comply with the requirements of the current radiological consequence analyses. Therefore, the proposed changes are acceptable with regard to the radiological consequences of all the postulated DBAs discussed in Columbia's FSAR Chapter 15.
3.3 Proposed Change to Surveillance Requirement 3.6.4.1.4 The proposed change to SR 3.6.4.1.4 is editorial only and does not change any technical aspects of SR 3.6.4.1.4. The NRC staff determined that the change is acceptable.
3.4 Variations from the Approved Traveler The licensee is proposing the following variations from the TS changes described in TSTF-551, Revision 3, or the applicable parts of the NRC staff's SE.. These variations do not affect the applicability of TSTF-551 or the NRC staffs SE to the proposed license amendment.
The Columbia TS wording for SR 3.6.4.1.3 is different than wording in NUREG-1433, "Standard Technical Specifications - General Electric Plants (BWR/4) Plants," Revision 4, Volume 1, (ADAMS Accession No. ML12104A192). Columbia's SR 3.6.4.1.3 states, "Verify each secondary containment access inner door or each secondary containment access outer door in each access opening is closed." SR 3.6.4.1.3 in NUREG-1433 states, "Verify one [secondary]
containment access door in each access opening is closed." Both wordings are equivalent and this difference is administrative and does not affect the applicability of TSTF-551 to the Columbia TSs.
The Columbia TS wording for SR 3.6.4.1.4 is different than the wording in NUREG-1433. The Columbia SR 3.6.4.1.4 states, "Verify each standby gas treatment (SGT) subsystem will draw down the secondary containment to 2:: 0.25 inch of vacuum water gauge ins 120 seconds."
SR 3.6.4.1.4 in NUREG-1433 states, "Verify [secondary] containment can be drawn down to 2:: [0.25] inch of vacuum water gauge ins [120] seconds using one standby gas treatment (SGT) subsystem." Both wordings are equivalent and this difference is administrative and does not affect the applicability of TSTF-551 to the Columbia TSs.
The final model SE for TSTF-551 discusses that the NRC staff review determined that there are two DBAs (the LOCA and the fuel handling accident (FHA)) that take credit for the secondary containment and are possibly impacted by the brief, inadvertent, and simultaneous opening of both an inner and outer access door during normal entry and exit conditions: The Columbia FHA does not credit the secondary containment or SGT system for mitigation of FHAs.
Because the Columbia FHA radiological consequence analysis does not credit the secondary containment or the SGT system, the FHA in secondary containment analysis is not impacted by the brief, inadvertent, and simultaneous opening of both an inner and outer access door during normal ent.ry and exit conditions. This difference does not alter the conclusion that the proposed change is applicable to Columbia.
3.5 Summary The NRC staff reviewed the proposed changes to the TSs and determined that they meet the standards for TSs as stated at 10 CFR 50.36(b ). The proposed SRs 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, and satisfy 10 CFR 50.36(c)(3). Additionally, the proposed changes to the TSs were reviewed for technical clarity and consistency with customary terminology and format in accordance with SRP Chapter 16.
Additionally, the NRC staff has evaluated the impact of the proposed changes on the design basis radiological consequence analyses against the regulatory requirements and guidance identified in Section 2.3 of this SE. The NRC staff finds, with reasonable assurance that the licensee's proposed changes to the TSs will continue to comply with the requirements of the current radiological consequence analyses. Therefore, the proposed changes are acceptable with regard to the radiological consequences of the postulated DBAs.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Washington State official was notified of the proposed issuance of the amendment on August 8, 2018. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes SRs.
The NRC staff has determined that the 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. The Commission has previously issued a proposed finding that the amendment involves 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 10915). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 1 O 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: K. Bucholtz Date: September 6, 2018
ML18221A107 OFFICE NRR/DORL/LPL4/PM NRR/DORL/LPL4/LA NRR/DSS/STSB/BC NAME JKlos PB1echman VCusumano DATE 8/17/2018 8/16/2018 8/28/2018 OFFICE OGC*
NRR/DORL/LPL4/BC NRR/DORL/LPL4/PM NAME DRoth RPascarelli (LRegner for)
JKlos DATE 8/29/2018 9/6/2018 9/6/2018