Text

Issuance of Amendment Adoption of TSTF-551 Revise Secondary Containment Surveillance Requirements
	ML18291B214
Person / Time
Site:	Monticello
Issue date:	11/26/2018
From:	Robert Kuntz; Plant Licensing Branch III
To:	Church C; Northern States Power Company, Minnesota
	Kuntz R
References
	EPID L-2018-LLA-0192, TSTF-551
	Download: ML18291B214 (16)
	v • d • e

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. Christopher Church Site Vice President Monticello Nuclear Generating Plant November 26, 2018 Northern States Power Company - Minnesota (NSPM) 2807 West County Road 75 Monticello, MN 55362-9637

SUBJECT:

MONTICELLO NUCLEAR GENERATING PLANT-ISSUANCE OF AMENDMENT RE: ADOPTION OF TSTF-551, "REVISE SECONDARY CONTAINMENT SURVEILLANCE REQUIREMENTS" (EPID L-2018-LLA-0192)

Dear Mr. Church:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 199 to Renewed Facility Operating License No. DPR-22 for the Monticello Nuclear Generating Plant.

The amendment consists of changes to the technical specifications (TSs) in response to your application dated July 3, 2018.

The amendment changes the TSs to adopt Technical Specifications Task Force (TSTF) traveler, TSTF-551, Revision 3, "Revise Secondary Containment Surveillance Requirements."

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

Docket No. 50-263

Enclosures:

1. Amendment No. 199 to DPR-22

2. Safety Evaluation cc: ListServ Sincerely, Robert F. Kuntz, Senior Project Manager Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 NORTHERN STATES POWER COMPANY DOCKET NO. 50-263 MONTICELLO NUCLEAR GENERATING PLANT AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 199 License No. DPR-22

1.

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

A.

The application for amendment by Northern States Power Company (NSPM, the licensee), dated July 3, 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 Renewed Facility Operating License No. DPR-22 is hereby amended to read as follows:

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 199, are hereby incorporated in the license. NSPM 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 90 days.

Attachment:

Changes to the Renewed Operating License No. DPR-22 and Technical Specifications FOR THE NUCLEAR REGULATORY COMMISSION OJ 9 V David J. Wrona, Chief Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance:

November 2 6, 2 O 1 8

ATTACHMENT TO LICENSE AMENDMENT NO. 199 MONTICELLO NUCLEAR GENERATING PLANT RENEWED FACILITY OPERATING LICENSE NO. DPR-22 DOCKET NO. 50-263 Replace the following page of Renewed Facility Operating License DPR-22 with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.

REMOVE 3

INSERT 3

Replace the following page of Appendix A, Technical Specifications, with the attached revised page. The revised page is identified by amendment number and contain marginal lines indicating the areas of change.

REMOVE 3.6.4.1-2 INSERT 3.6.4.1-2

2.

Pursuant to the Act and 10 CFR Part 70, NSPM to receive, possess, and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operations, as described in the Final Safety Analysis Report, as supplemented and amended, and the licensee's filings dated August 16, 197 4 (those portions dealing with handling of reactor fuel);

3.

Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NSPM to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required;

4.

Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NSPM to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and

5.

Pursuant to the Act and 10 CFR Parts 30 and 70, NSPM to possess, but not separate, such byproduct and special nuclear material as may be produced by operation of the facility.

C.

This renewed operating license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations 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 NSPM is authorized to operate the facility at steady state reactor core power levels not in excess of 2004 megawatts (thermal).

2.

Technical Specifications

3.

The Technical Specifications contained in Appendix A, as revised through Amendment No. 199, are hereby incorporated in the license. NSPM shall operate the facility in accordance with the Technical Specifications.

Physical Protection NSPM shall implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Renewed License No. DPR-22 Amendment No. 199

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 Monticello SURVEILLANCE

NOTE--------------------------------

Not required to be met for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months if analysis demonstrates one standby gas treatment (SGT}

subsystem is capable of establishing the required secondary containment vacuum.

FREQUENCY Verify secondary containment vacuum is 2'. 0.25 inch 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of vacuum water gauge.

Verify all secondary containment equipment hatches are closed and sealed.

Verify one secondary containment access door in each access opening is closed, except when the access opening is being used for entry and exit.

Verify the secondary containment can be maintained 2'. 0.25 inch of vacuum water gauge for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months using one SGT subsystem at a flow rate

4000 cfm.

31 days 31 days 24 months on a STAGGERED TEST BASIS for each SGT subsystem 3.6.4.1-2 Amendment No. 448, 199 I

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 199 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-22 NORTHERN STATES POWER COMPANY MONTICELLO NUCLEAR GENERATING PLANT DOCKET NO. 50-263

1.0 INTRODUCTION

By application dated July 3, 2018, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML18187A400, Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (the licensee) requested changes to the technical specifications {TSs) for Monticello Nuclear Generating Plant (MNGP). Specifically, the licensee requested changes to the TS to 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 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 TS to permit secondary containment access opening 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 MNGP, the SGT must be able to establish the required vacuum within 5 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 TS Changes

The proposed changes 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 ~ 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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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 The SR 3.6.4.1.3 requires verification that one secondary containment access 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."

2.3 Regulatory Requirements and Guidance The regulation at 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), when an LCO of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the condition can be met.

The regulation at 10 CFR 50.36(c)(3) requires TSs to include items in the category of 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 LCOs 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" (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 alternate source term (AST) amendment requests.

SRP 15.0.1 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).

The RG 1.183 provides acceptable methodology for analyzing the radiological consequences of several design basis accidents to show compliance with 10 CFR 50.67. 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.

10 CFR 50.67(b)(2), "Accident source term," states:

(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 (25 rem)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 Commission 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 10 CFR 50.54.

The regulation at 10 CFR 50.36(a)(1) states, in part: "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." Accordingly, along with the proposed TS changes, the licensee also submitted TS Bases changes that correspond to the proposed TS changes for information only.

3.1 Proposed Change to SR 3.6.4.1.1 A note is being added to SR 3.6.4.1.1. The note allows the SR to not be met for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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 degradations of the secondary containment boundary or of the ability of the SGT system to perform its specified safety function.

The note provides 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 note allowing a 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months exception to meeting the SR acceptance criterion.

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.

While the duration of occurrences in which the secondary containment vacuum limit is not met is anticipated to be very brief, the allowance is permitted for a maximum of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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. SR 3.6.4.1.4 requires verification that the secondary containment can be maintained ;:: 0.25 inch of vacuum water gauge for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months using one SGT subsystem at a flow rate s 4000 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, non-safety 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, 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: (1) the secondary containment pressure can be maintained 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 is maintained, 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 requires 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 SR 3.6.4.1.3 The NRC staff review was limited to the licensee's request to provide an allowance for the brief simultaneous opening of redundant secondary containment access doors during normal entry and exit conditions. Normal entry and exit conditions do not include 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.

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 DBA that takes credit for the secondary containment, and is possibly impacted by the brief simultaneous opening of both an inner and outer access door during normal entry and exit conditions, 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 5 minutes. TS SR 3.6.4.1 verifies each SGT subsystem can maintain greater than or equal to 0.25 inches of vacuum water gauge in the secondary containment at a flow rate of less than or equal to 4000 cubic feet per minute.

Conservatively, the DBA LOCA radiological consequence analysis in the licensee's Updated Safety Analysis Report (USAR), Chapter 14, 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 5 minutes. The license assumes that releases into the secondary containment prior to the 5 minute draw down time leak directly to the environment as a ground level release with no filtration. After the assumed 5 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 5 minutes from the start of the DBA LOCA. Margin exists to ensure that the secondary containment can be reestablished during a brief 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 result. The brief, simultaneous opening of the secondary containment access doors does not impact the design bases and will not result in an increase in any on-site or off-site 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 148 for MNGP and most recently updated the radiological dose consequences analyses for DBAs via license amendment 188. The NRC staff reviewed the impact of the proposed changes to MNGP TS, on all DBAs currently analyzed in the MNGP USAR that could have the potential for significant dose consequences. Chapter 14 of the MNGP USAR describes the DBAs and their radiological consequence analysis results.

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 change to the TSs will continue to comply with these criteria and that that the licensee's estimate of the dose consequence 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 the postulated DBAs.

3.3 Variation from the Approved Traveler The licensee is proposing the following variations from the TS changes described in TSTF-551 or the applicable parts of TSTF-551 or 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 MNGP TS do not contain an SR equivalent to SR 3.6.4.1.4 modified by TSTF-551.

Therefore, the editorial change to SR 3.6.4.1.4 is not applicable.

The Traveler discusses the applicable regulatory requirements and guidance, including the 10 CFR 50, Appendix A, General Design Criteria (GDC). MNGP was not licensed to the 10 CFR 50, Appendix A, GDC. MNGP was designed and constructed to comply with the intent of the 70 GDCs for Nuclear Power Plant Construction Permits proposed by the Atomic Energy Commission in July 1967. The MNGP equivalents of the specific principal design criteria to the referenced GDCs are discussed in MNGP USAR Subsection 1.2.4. These differences do not alter the conclusion that the proposed change is applicable to MNGP.

The final model SE for TSTF-551 discusses that the NRC staff review determined that there are two design basis accidents that take credit for the secondary containment and are possibly impacted by the brief and simultaneous opening of both an inner and outer access door during normal entry and exit conditions: the LOCA and the fuel handling

3.4 accident (FHA) in secondary containment. The MNGP FHA does not credit the secondary containment or SGT system for mitigation of fuel handling accidents greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following reactor shutdown. Because the MNGP 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 and simultaneous opening of both an inner and outer access door during normal entry and exit conditions. This difference does not alter the conclusion that the proposed change is applicable to MNGP.

Summary The NRC staff reviewed the proposed changes and determined that changes to the TS meet the standards for TS in 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 changes to the TS 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 change 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 Minnesota State official was notified of the proposed issuance of the amendment on October 17, 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 the surveillance requirements. The 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 August 28, 2018 (83 FR 43906).

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: Kristy Bucholtz, NRR Date of issuance: November 26, 2018

ML18291B214 OFFICE NRR/DORULPL3/PM NAME RKuntz DATE 10/22/18 OFFICE OGCNLO NAME KGamin**

DATE 11/08/18 NRR/DORULPL3/LA SRohrer 10/22/18 NRR/DORULPL3/BC DWrona 11/26/18

v t e Letter Sequence Approval
EPID:L-2018-LLA-0192, TSTF-551 (Approved, Closed)
Initiation Request Acceptance Administration Questions Answers Results Approval Other:
MONTHYEAR2018-07-03 [Table View]

ML18291B214