ML18318A045
| ML18318A045 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 11/19/2018 |
| From: | Robert Kuntz Plant Licensing Branch III |
| To: | Sharp S Northern States Power Co |
| Kuntz R, 415-3733 | |
| References | |
| EPID L-2018-LLA-0261 | |
| Download: ML18318A045 (5) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 November 19, 2018 Mr. Scott M. Sharp Vice President Northern States Power Company - Minnesota Prairie Island Nuclear Generating Plant 1717 Wakonade Drive East Welch, MN 55089-9642
SUBJECT:
PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 -
SUPPLEMENTAL INFORMATION NEEDED FOR ACCEPTANCE OF REQUESTED LICENSING ACTION RE: AMENDMENT TO RECLASSIFY CERTAIN FUEL HANDLING EQUIPMENT (EPID L-2018-LLA-0261)
Dear Mr. Sharp:
By letter dated October 2, 2018, Northern States Power Company (NSPM) submitted a license amendment request for Prairie Island Nuclear Generating Plant, Units 1 and 2. The proposed amendment would allow certain fuel handling equipment to be reclassified from quality assurance (QA) Type I ("safety-related") to QA Type Ill ("non-safety related"). The purpose of this letter is to provide the results of the U.S. Nuclear Regulatory Commission (NRC) staff's acceptance review of this amendment request. The acceptance review was performed to determine if there is sufficient technical information in scope and depth to allow the NRC staff to complete its detailed technical review. The acceptance review is also intended to identify whether the application has any readily apparent information insufficiencies in its characterization of the regulatory requirements or the licensing basis of the plant.
Consistent with Section 50.90 of Title 10 of the Code of Federal Regulations ( 10 CFR), an amendment to the license (including the technical specifications) must fully describe the changes requested, and following as far as applicable, the form prescribed for original applications. Section 50.34 of 10 CFR addresses the content of technical information required.
This section stipulates that the submittal address the design and operating characteristics, unusual or novel design features, and principal safety considerations.
The NRC staff has reviewed your application and concluded that the information delineated in the enclosure to this letter is necessary to enable the staff to make an independent assessment regarding the acceptability of the proposed amendment in terms of regulatory requirements and the protection of public health and safety and the environment.
In order to make the application complete, the NRC staff requests that NSPM supplement the application to address the information requested in the enclosure by December 5, 2018. This will enable the NRC staff to begin its detailed technical review. If the information responsive to the NRC staff's request is not received by the above date, the application will not be accepted for review pursuant to 10 CFR 2.101, and the NRC will cease its review activities associated with the application. If the application is subsequently accepted for review, you will be advised of any further information needed to support the staff's detailed technical review by separate correspondence.
S. Sharp The information requested and associated time frame in this letter were discussed with Jeffery Kivi of your staff on November 15, 2018.
If you have any questions, please contact the Senior Project Manager, Robert F. Kuntz, at (301) 415-3733.
Sincerely, Robert F. Kun. r. Project Manager Plant Licensing ranch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No(s). 50-282 and 50-306
Enclosure:
As stated cc: Listserv
SUPPLEMENTAL INFORMATION NEEDED LICENSE AMENDMENT REQUEST NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT NOS. 1 AND 2 DOCKET NOS. 50-282 AND 50-306 By letter dated October 2, 2018 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML18275A370), Northern States Power Company (the licensee) submitted a license amendment request (LAR) for the Prairie Island Nuclear Generating Plant (PINGP). The proposed amendment would allow certain fuel handling equipment to be reclassified from Quality Assurance (QA) Type I ("safety-related") to QA Type Ill ("non-safety related").
In accordance with the Office Instruction LIC-109, Revision 2, "Acceptance Review Procedures" (ADAMS Accession No. ML16144A521), the NRC staff reviewed the LAR dated October 2, 2018, for administrative and technical sufficiency. The NRC staff found that the LAR contains information insufficiencies contrary to LIC-109, Section 3.1.2, "Technical Staff Criteria." Contrary to Section 3.1.2, the LAR contains insufficient information in the areas of "Completeness of Scope," and "Sufficiency of Information."
The Introduction to Appendix 8, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," of 10 CFR, Part 50, "Domestic Licensing of Production and Utilization Facilities," states, in part:
Every applicant for a construction permit is required by the provisions of§ 50.34 to include in its preliminary safety analysis report a description of the quality assurance program to be applied to the design, fabrication, construction, and testing of the structures, systems, and components of the facility. Every applicant for an operating license is required to include, in its final safety analysis report, information pertaining to the managerial and administrative controls to be used to assure safe operation . . . . Nuclear power plants and fuel reprocessing plants include structures, systems, and components that prevent or mitigate the consequences of postulated accidents that could cause undue risk to the health and safety of the public. This appendix establishes quality assurance requirements for the design, manufacture, construction, and operation of those structures, systems, and components. The pertinent requirements of this appendix apply to all activities affecting the safety-related functions of those structures, systems, and components; these activities include designing, purchasing, fabricating, handling, shipping, storing, cleaning, erecting, installing, inspecting, testing, operating, maintaining, repairing, refueling, and modifying.
The October 2, 2018, LAR does not properly apply the applicable regulations in that the LAR does not demonstrate that, after the change, pertinent requirements of Appendix B to 10 CFR, Part 50, would continue to apply to all activities affecting the safety-related functions of Enclosure
certain structures, systems, and components. Specifically, the NRC staff identified the following information insufficiencies:
- Sufficiency of information: The equipment proposed for quality downgrade on the basis of the low dose consequences of the design-basis fuel handling accident includes large structures that operate over and could affect the safety-related functions of the reactor vessel, spent fuel pool, and irradiated fuel. However, the consequences of collapse or failure of these large structures, such as the manipulator crane, the spent fuel pool bridge crane, and the auxiliary building crane, was not demonstrated to be bounded by the fuel handling accident.
- Completeness of scope: The auxiliary building crane was licensed to the.guidance in NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants," for single-failure-proof cranes, and, on that basis, the cask handling accident was removed from the PINGP Updated Final Safety Analysis Report (USAR). Guidance in NUREG-0554, "Single Failure Proof Cranes for Nuclear Power Plants," which is referenced by NUREG-0612 for single failure proof handling systems specifies application of quality assurance measures for the design, fabrication, installation, testing, and operation of crane handling systems for safe handling of critical loads. The PINGP USAR discusses the movement of dry fuel storage casks as a critical load and also notes spent fuel pool barrier movement is within the maximum critical load rating of the single failure proof hoist on the spent fuel pool bridge crane. However, the amendment did not evaluate and justify the potential adverse effect on the safety-related functions of nearby structures and components during critical load handling evolutions that could result from the proposed change in Quality Type of these cranes.
ML18318A045 *via e-mail OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DSS/SCPB NAME RKuntz SRohrer SAnderson*
DATE 11/14/18 11/14/18 11/8/18 OFFICE NRR/DORL/LPL3/BC NRR/DORL/LPL3/PM NAME DWrona RKuntz DATE 11/19/18 11/19/18