Letter Sequence Response to RAI |
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Category:Letter type:L
MONTHYEARL-MT-23-054, Subsequent License Renewal Application Supplement 82024-01-11011 January 2024
Subsequent License Renewal Application Supplement 8
L-MT-23-047, License Amendment Request: Revision to the MNGP Pressure Temperature Limits Report to Change the Neutron Fluence Methodology and Incorporate New Surveillance Capsule Data2023-12-29029 December 2023
License Amendment Request: Revision to the MNGP Pressure Temperature Limits Report to Change the Neutron Fluence Methodology and Incorporate New Surveillance Capsule Data
L-MT-23-056, Subsequent License Renewal Application Response to Request for Additional Information and Request for Confirmation of Information - Set 1 Part 22023-12-18018 December 2023
Subsequent License Renewal Application Response to Request for Additional Information and Request for Confirmation of Information - Set 1 Part 2
L-MT-23-042, 2023 Annual Report of Changes in Emergency Core Cooling System Evaluation Models Pursuant to 10 CFR 50.462023-12-11011 December 2023
2023 Annual Report of Changes in Emergency Core Cooling System Evaluation Models Pursuant to 10 CFR 50.46
L-MT-23-052, Subsequent License Renewal Application Supplement 72023-11-30030 November 2023
Subsequent License Renewal Application Supplement 7
L-MT-23-051, Update to the Technical Specification Bases2023-11-28028 November 2023
Update to the Technical Specification Bases
L-MT-23-049, Subsequent License Renewal Application Response to Request for Additional Information and Request for Confirmation of Information - Set 12023-11-21021 November 2023
Subsequent License Renewal Application Response to Request for Additional Information and Request for Confirmation of Information - Set 1
L-MT-23-043, 10 CFR 50.55a(z)(1) Request Regarding OMN-17, Revision 1. VR-092023-11-13013 November 2023
10 CFR 50.55a(z)(1) Request Regarding OMN-17, Revision 1. VR-09
L-MT-23-038, License Amendment Request to Revise Monticello Technical Specification Surveillance Requirement 3.8.6.62023-11-10010 November 2023
License Amendment Request to Revise Monticello Technical Specification Surveillance Requirement 3.8.6.6
L-MT-23-046, Subsequent License Renewal Application Response to Request for Additional Information Round 2 - Set 12023-11-0909 November 2023
Subsequent License Renewal Application Response to Request for Additional Information Round 2 - Set 1
L-MT-23-041, Subsequent License Renewal Application Response to Request for Confirmation of Information Set 22023-10-0303 October 2023
Subsequent License Renewal Application Response to Request for Confirmation of Information Set 2
L-MT-23-037, Subsequent License Renewal Application Response to Request for Additional Information Set 32023-09-22022 September 2023
Subsequent License Renewal Application Response to Request for Additional Information Set 3
L-MT-23-036, Subsequent License Renewal Application Response to Request for Additional Information Set 2 and Supplement 62023-09-0505 September 2023
Subsequent License Renewal Application Response to Request for Additional Information Set 2 and Supplement 6
L-MT-23-035, Subsequent License Renewal Application Supplement 52023-08-28028 August 2023
Subsequent License Renewal Application Supplement 5
L-MT-23-034, Subsequent License Renewal Application Response to Request for Additional Information Set 12023-08-15015 August 2023
Subsequent License Renewal Application Response to Request for Additional Information Set 1
L-MT-23-028, 2023 Refueling Outage 90-Day Inservice Inspection (ISI) Summary Report2023-07-31031 July 2023
2023 Refueling Outage 90-Day Inservice Inspection (ISI) Summary Report
L-MT-23-032, 10 CFR 50.55a(z)(2) Request Regarding MO-2397, VR-112023-07-31031 July 2023
10 CFR 50.55a(z)(2) Request Regarding MO-2397, VR-11
L-MT-23-031, Subsequent License Renewal Application Supplement 4 and Responses to Request for Confirmation of Information - Set 12023-07-18018 July 2023
Subsequent License Renewal Application Supplement 4 and Responses to Request for Confirmation of Information - Set 1
L-MT-23-030, Subsequent License Renewal Application Supplement 32023-07-0404 July 2023
Subsequent License Renewal Application Supplement 3
L-MT-23-025, Subsequent License Renewal Application Supplement 22023-06-26026 June 2023
Subsequent License Renewal Application Supplement 2
L-MT-23-019, Submittal of 2022 Annual Radiological Environmental Operating Report2023-05-10010 May 2023
Submittal of 2022 Annual Radiological Environmental Operating Report
L-MT-23-020, Submittal of 2022 Annual Radioactive Effluent Release Report2023-05-10010 May 2023
Submittal of 2022 Annual Radioactive Effluent Release Report
L-MT-23-021, Core Operating Limits Report (COLR) for the Monticello Nuclear Generating Plant for Cycle 322023-05-0202 May 2023
Core Operating Limits Report (COLR) for the Monticello Nuclear Generating Plant for Cycle 32
L-MT-23-017, 2022 Annual Report of Individual Monitoring for the Monticello Nuclear Generating Plant (MNGP)2023-04-18018 April 2023
2022 Annual Report of Individual Monitoring for the Monticello Nuclear Generating Plant (MNGP)
L-MT-23-010, Subsequent License Renewal Application Supplement 12023-04-0303 April 2023
Subsequent License Renewal Application Supplement 1
L-MT-23-013, Core Operating Limits Report (COLR) for Cycle 31, Revision 32023-03-28028 March 2023
Core Operating Limits Report (COLR) for Cycle 31, Revision 3
L-MT-23-012, Core Operating Limits Report (COLR) for Monticello Nuclear Generating Plant Cycle 31, Revision 22023-03-17017 March 2023
Core Operating Limits Report (COLR) for Monticello Nuclear Generating Plant Cycle 31, Revision 2
L-MT-23-008, 10CFR50.55a Request to Use Later Edition of ASME Section XI for ISI Code of Record (RR-003)2023-02-0707 February 2023
10CFR50.55a Request to Use Later Edition of ASME Section XI for ISI Code of Record (RR-003)
L-MT-23-004, CFR 50.55a Request RR-001, Request to Use a Provision of a Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI for the Monticello Third Interval Containment Inservice Inspection Program2023-01-23023 January 2023
CFR 50.55a Request RR-001, Request to Use a Provision of a Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI for the Monticello Third Interval Containment Inservice Inspection Program
L-MT-23-005, Response to the NRC Request for Additional Information Regarding the 50.55a Request Pr 08, HPCI Pump Quarterly Testing (EPID Number L-2022-LLR-0088)2023-01-0606 January 2023
Response to the NRC Request for Additional Information Regarding the 50.55a Request Pr 08, HPCI Pump Quarterly Testing (EPID Number L-2022-LLR-0088)
L-MT-22-049, Industry Groundwater Protection Initiative Special Report2022-12-15015 December 2022
Industry Groundwater Protection Initiative Special Report
L-MT-22-052, L-MT-22-052 Monticello Nuclear Generating Plant 10 CFR 50.55a Request No. Pr 08, Request for HPCI Pump Quarterly Alternative2022-12-15015 December 2022
L-MT-22-052 Monticello Nuclear Generating Plant 10 CFR 50.55a Request No. Pr 08, Request for HPCI Pump Quarterly Alternative
L-MT-22-046, 2022 Annual Report of Changes in Emergency Core Cooling System Evaluation Models Pursuant to 10 CFR 50.462022-12-13013 December 2022
2022 Annual Report of Changes in Emergency Core Cooling System Evaluation Models Pursuant to 10 CFR 50.46
L-MT-22-048, Update to the Monticello Technical Specification Bases2022-11-28028 November 2022
Update to the Monticello Technical Specification Bases
L-MT-22-047, Withdrawal of Request for Relief from ASME OM Code for the Sixth Inservice Testing Interval2022-11-10010 November 2022
Withdrawal of Request for Relief from ASME OM Code for the Sixth Inservice Testing Interval
L-MT-22-045, Letter Submitting Post-Exam Package2022-11-0404 November 2022
Letter Submitting Post-Exam Package
L-MT-22-030, Sixth Interval Inservice Testing (1ST) Plan2022-09-0606 September 2022
Sixth Interval Inservice Testing (1ST) Plan
L-MT-22-037, Supplement to 10 CFR 50.55a Request Associated with the Monticello Sixth Inservice Testing Ten-Year Interval, Alternative VR-10, Excess Flow Check Valve Testing Frequency2022-08-29029 August 2022
Supplement to 10 CFR 50.55a Request Associated with the Monticello Sixth Inservice Testing Ten-Year Interval, Alternative VR-10, Excess Flow Check Valve Testing Frequency
L-MT-22-007, Response to Request for Additional Information for the Monticello Nuclear Generating Plant Alternative Request VR-08 (EPID: L-MT-22-007)2022-07-22022 July 2022
Response to Request for Additional Information for the Monticello Nuclear Generating Plant Alternative Request VR-08 (EPID: L-MT-22-007)
L-MT-22-026, Changes to the Emergency Plan2022-07-19019 July 2022
Changes to the Emergency Plan
L-MT-22-024, Response to a Request for Additional Information for the Monticello Nuclear Generating Plant Related to the Amendment to Adopt Advanced Framatome Methodologies2022-06-0606 June 2022
Response to a Request for Additional Information for the Monticello Nuclear Generating Plant Related to the Amendment to Adopt Advanced Framatome Methodologies
L-MT-22-022, Response to a Request for Additional Information: Monticello Alternative VR-10, Excess Flow Check Valve Testing Frequency2022-05-25025 May 2022
Response to a Request for Additional Information: Monticello Alternative VR-10, Excess Flow Check Valve Testing Frequency
L-MT-22-017, 2021 Annual Radiological Environmental Operating Report2022-05-11011 May 2022
2021 Annual Radiological Environmental Operating Report
L-MT-22-018, 2021 Annual Radioactive Effluent Release Report2022-05-11011 May 2022
2021 Annual Radioactive Effluent Release Report
L-MT-22-016, 2021 Annual Report of Individual Monitoring2022-04-28028 April 2022
2021 Annual Report of Individual Monitoring
L-MT-22-019, Withdrawal of Requests for Relief from ASME OM Code for the Sixth Inservice Testing Interval2022-04-18018 April 2022
Withdrawal of Requests for Relief from ASME OM Code for the Sixth Inservice Testing Interval
L-MT-22-010, License Amendment Request to Revise Technical Specification 3.6.1.8 Residual Heat Removal (RHR) Drywell Spray Header and Nozzle Surveillance Frequency2022-03-18018 March 2022
License Amendment Request to Revise Technical Specification 3.6.1.8 Residual Heat Removal (RHR) Drywell Spray Header and Nozzle Surveillance Frequency
L-MT-22-012, Special Report for the Bypass of the Offgas Treatment Storage System2022-03-15015 March 2022
Special Report for the Bypass of the Offgas Treatment Storage System
L-MT-22-008, 10 CFR 50.55a Request Associated with the Monticello Sixth Inservice Testing Ten-Year Interval Alternative Related to Excess Flow Check Valve Testing Frequency (L-MT-22-008)2022-03-0707 March 2022
10 CFR 50.55a Request Associated with the Monticello Sixth Inservice Testing Ten-Year Interval Alternative Related to Excess Flow Check Valve Testing Frequency (L-MT-22-008)
L-MT-22-006, 10 CFR 50.55a Request Associated with the Monticello Sixth Inservice Testing Ten-Year Interval OMN-26 (L-MT-22-006)2022-02-18018 February 2022
10 CFR 50.55a Request Associated with the Monticello Sixth Inservice Testing Ten-Year Interval OMN-26 (L-MT-22-006)
2024-01-11
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Text
(l Xcel Energy Monticello Nuclear Generating Plant 2807 W County Road 75 Monticello, MN 55362 July 3, 2014 L-MT-14-057 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Monticello Nuclear Generating Plant Docket 50-263 Renewed Facility Operating License No. DPR-22 Response to Requests for Additional Information for the License Amendment Request to Reduce the Reactor Steam Dome Pressure Specified in the Reactor Core Safety Limits (TAG No. MF1054)
References:
- 1) NSPM to NRC, "License Amendment Request: Reduce the Reactor Steam Dome Pressure Specified in the Reactor Core Safety Limits,"
(L-MT-13-01 0) dated March 11, 2013.
- 2) NRC to NSPM, "Monticello Nuclear Generating Plant- Draft Requests for Additional Information re : Amendment to Reduce Reactor Steam Dome Pressure Safety Limit (TAG No. MF1054)," dated May 27 , 2014.
On March 11, 2013, in accordance with 10 CFR 50.90, the Northern States Power Company- Minnesota (NSPM), doing business as Xcel Energy, Inc., submitted a License Amendment Request (LAR) (Reference 1) proposing to reduce the reactor steam dome pressure specified within Reactor Core Safety Limits Specification 2.1.1, in the Technical Specifications (TSs). This change will resolve a 10 CFR Part 21 condition concerning a potential to momentarily violate Reactor Core Safety Limit 2.1.1.1 during a Pressure Regulator Failure Maximum Demand (Open) transient.
On May 27, 2014, the U.S. Nuclear Regulatory Commission (NRC) requested additional information (RAI) from NSPM (Reference 2) to complete their review. Enclosure 1 provides the requested information.
In accordance with 10 CFR 50.91, a copy of this response, with enclosure, is being provided to the designated Minnesota Official.
Document Control Desk L-MT-14-057 Page 2 of 2 If you have any questions or require additional information, please contact Mr. Richard Loeffler at (763) 295-1247.
Summary of Commitments This letter proposes no new commitments and does not revise any existing commitments.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on July )' , 2014.
~~J~ ~/-;/-
Karen D. Fili, Site Vice-President, Monticello Nuclear Generating Plant Northern States Power Company- Minnesota Enclosure cc: Regional Administrator, Region Ill, USNRC Project Manager, Monticello Nuclear Generating Plant, USNRC Resident Inspector, Monticello Nuclear Generating Plant, USNRC State of Minnesota
ENCLOSURE 1 MONTICELLO NUCLEAR GENERATING PLANT RESPONSE TO REQUESTS FOR ADDITIONAL INFORMATION FOR THE LICENSE AMENDMENT REQUEST TO REDUCE THE REACTOR STEAM DOME PRESSURE SPECIFIED IN THE REACTOR CORE SAFETY LIMITS (3 Pages Follow)
L-MT-14-057 Enclosure Page 1 of 3 RESPONSE TO REQUESTS FOR ADDITIONAL INFORMATION FOR THE LICENSE AMENDMENT REQUEST TO REDUCE THE REACTOR STEAM DOME PRESSURE SPECIFIED IN THE REACTOR CORE SAFETY LIMITS On May 27, 2014, the U.S. Nuclear Regulatory Commission (NRC) transmitted requests for additional information (RAis) to Northern States Power Company- Minnesota (NSPM). The RAis are provided in italic type with the NSPM responses for each portion of the requests immediately following.
SRXB (Reactor Systems Branch) RAJ# 1 Since the proposed approach for MNGP is a plant-specific resolution of the 10 CFR Part 21 issue as discussed in your submittal (Reference 1), please provide the following additional information:
- 1. Discuss how Northern States Power Company- Minnesota (NSPM) plans to address the Part 21 issue when the MNGP core may be a mixed-core design consisting of more than one fuel design whose critical power ratio (CPR) correlations have different lower bound pressures.
NSPM Response There are two cases, a transition from one fuel vendor to another, and/or utilizing a fuel design with a different CPR lower bound pressure. The licensee/fuel vendor reload safety analysis process applies in either case.
For a fuel transition, a mixed-core (different vendor fuel designs) will result for several cycles. A license amendment is required for a fuel vendor transition.
For the Monticello AREVA fuel transition, NSPM submitted a license amendment request to adopt the AREVA safety analysis methodology (Reference 2) . As discussed in Enclosure 1, Section 3.4.1 of Reference 2, the proposed AREVA correlations and methodologies will be valid within the proposed safety limits such that any mixed core will be operated within the bounds of an NRC approved methodology. When a fuel design from the same vendor with a different CPR lower bound pressure is to be loaded, the reload safety analysis will again apply. The reload safety analysis process includes identifying various design inputs- including Technical Specification (TS) requirements. The CPR correlations are inherent in the design of the fuel and are considered as part of the reload safety analysis process.
L-MT-13-057 Page 2 of 3
- 2. Describe the current MNGP core design, including the fuel types in use.
NSPM Response The MNGP reactor core consists of 484 fuel assemblies of [General Electric]
GE14 fuel. The current core design was provided in the Supplemental Reload Licensing Report (SRLR) which was submitted in a letter dated December 20, 2013 (Reference 3) .
When the MNGP core transitions to a fuel design whose lower bound pressure for the CPR correlation is higher (or lower) than that of the current CPR correlation, discuss how NSPM intends to address the change. The discussion should include whether it will require a Technical Specification (TS) license amendment to address the change.
NSPM Response As indicated previously, there are two general cases, a transition from one fuel vendor to another, and/or the utilization of a fuel design with a different CPR lower bound pressure. The licensee/fuel vendor reload safety analysis process as discussed in the response to SRXB RAI # 1 applies in either case.
As long as the lower bound of the fuel's CPR correlation is less than the reactor steam dome pressure specified in the TS Reactor Core Safety Limits (686 psig was specified in the license amendment request (Reference 1) based on the core of GE14 fuel), no TS change is necessary. The AREVA fuel designs CPR correlations (lower bound) are less than the proposed reactor steam dome pressure value. If the fuel design(s) CPR correlation is not less than the TS specified value a license amendment would be necessary.
SRXBRA/#2 In Reference 1, page 2 of 15, it was stated, While this condition had been determined by GE to not involve an actual safety hazard, the potential for violation of a Reactor Core Safety Limit had been identified and restoration to comply with the safety limit is required."
The U.S. Nuclear Regulatory Commission (NRC) staff understands that although this issue may not be an actual safety hazard, by lowering the dome pressure in TS 2.1.1 from the current value of 785 psig to the proposed value of 686 psig may prevent
L-MT-13-057 Page 3 of 3 unnecessary reactor shutdowns as required by TS 2.2.2 if dome pressure goes below 785 psig during a transient.
Please discuss if there are any other operational and/or safety benefits of lowering dome pressure from the current value of 785 psig to 686 psig.
NSPM Response Lowering the value of reactor steam dome pressure in the TS has no physical effect on plant equipment and therefore, no impact on the course of plant transients. The change is an analytical exercise to demonstrate the applicability of correlations and methodologies. There are no known operational or safety benefits.
REFERENCES
- 1. NSPM to NRC, "License Amendment Request: Reduce the Reactor Steam Dome Pressure Specified in the Reactor Core Safety Limits," (L-MT-13-01 0),
dated March 11, 2013.
- 2. NSPM to NRC, "License Amendment Request for Transition to AREVA ATRIUM 1OXM Fuel and AREVA Safety Analysis Methodology," (L-MT-13-055), dated July 15, 2013.
- 3. NSPM to NRC, "Maximum Extended Load Line Limit Analysis Plus: Cycle 27 Safety Reload Licensing Report and Request for Additional Information Response," (L-MT-13-126), dated December 20, 2013. (ADAMS Accession No. ML13358A372)
