ML18200A272
| ML18200A272 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 07/19/2018 |
| From: | Daniel Schroeder Reactor Projects Branch 2 |
| To: | |
| Schroeder D | |
| References | |
| Download: ML18200A272 (8) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION REGION I 2100 RENAISSANCE BOULEVARD, SUITE 100 KING OF PRUSSIA, PA 19406-2713 July 19, 2018 MEETING
SUMMARY
LICENSEE: ENTERGY NUCLEAR NORTHEAST FACILITY: INDIAN POINT UNITS 2 AND 3
SUBJECT:
SUMMARY
OF ANNUAL ASSESSMENT MEETING On June 21, 2018, at 7:00 p.m., the U.S. Nuclear Regulatory Commission (NRC) conducted a public meeting at the Doubletree by Hilton in Tarrytown, New York, to discuss its assessment of the safety performance at Indian Point Units 2 and 3 for calendar year 2017.
A notice of the meeting was issued on May 15, 2018, and was posted on the NRCs external (public) Web page. The meeting notice can be found in the NRCs Agencywide Documents Access and Management System (ADAMS) with Accession Number ML18141A218. ADAMS is accessible from the NRC Web page at: http://www.nrc.gov/reading-rm/adams.html.
The NRC discussed its assessment of the safety performance of Indian Point Units 2 and 3 for the period of January 1 through December 31, 2017, as documented in our letter dated February 28, 2018 (ADAMS Accession Number ML18058A058). Additional information relative to the NRC's Annual Assessment Process and the safety performance of Indian Point Units 2 and 3 can be found on the NRC's web site at:
https://www.nrc.gov/reactors/operating/oversight.html.
The NRC also presented slides on two topics of public interest, which included the assessment of reactor pressure vessel O-ring leakage and weld onlay repairs to the Unit 2 reactor vessel upper head penetration #3. Following the discussion of plant performance and presentations, the NRC held a question and answer session.
Members of the public, local officials, State officials, and members of the media attended the meeting and were offered the opportunity to question the NRC regarding Entergys performance and the role of the agency in ensuring safe plant operations. Some of the questions required additional research with technical experts within the NRC. The answers regarding these topics are included as an enclosure to this letter.
/RA/
Daniel L. Schroeder, Chief Projects Branch 2 Division of Reactor Projects
Enclosures:
- 1. Annual Assessment Meeting Public Topics of Interest
- 2. NRC Staff Annual Assessment Meeting Presentation
ML18200A272 Non-Sensitive Publicly Available SUNSI Review Sensitive Non-Publicly Available OFFICE RI/DRP RI/DRP NAME TSetzer DSchroeder DATE 7/12/18 7/19/18 1
Annual Assessment Meeting Public Topics of Interest Spent Fuel Storage at Indian Point and Independent Spent Fuel Storage Installations Both Indian Point Unit 2 and 3 were originally licensed for a maximum capacity of 264 fuel assemblies in the spent fuel pools. Since then, analyses and evaluations have proven that the pool can safely accommodate more than the original licensed limit. Anytime a plants owner intends to increase the capacity of its spent fuel pool beyond the licensed amount, a thorough evaluation must be conducted to ensure the continued safe storage of the material, including a review of the increased heat load and an analysis of any increased potential for safety hazards.
In the case of Indian Point, this took place each time they changed the configuration of the spent fuel pools, providing assurance that the pools remained safe. The NRC independently reviewed each of the spent fuel pool evaluations and concluded that the spent fuel pools remain safe under the licensed loading limit. Currently, the Unit 2 spent fuel pool has a capacity of 1374 fuel assemblies, and the Unit 3 spent fuel pool has a capacity of 1345 fuel assemblies.
Entergy has also been granted license amendments to allow spent fuel transfer from the Unit 3 spent fuel pool to the Unit 2 spent fuel pool using a transfer cask. The spent fuel in the Unit 1 spent fuel pool was removed to dry cask storage and the spent fuel pool was drained by the end of 2008.
An independent spent fuel storage installation, or ISFSI, is a facility that is designed and constructed for the interim storage of spent nuclear fuel. These facilities are licensed separately from a nuclear power plant and are considered independent even though they may be located on the site of another NRC-licensed facility. The NRC is responsible for inspection of dry cask storage. All casks undergo a safety review before they are certified for use by the NRC. Before casks are loaded, inspectors with specific knowledge of ISFSI operations assess the adequacy of a "dry run" by the licensee; they then observe initial cask loadings. The on-site resident inspectors or region-based inspectors may observe later cask loadings, and the regional offices also perform periodic inspections of routine ISFSI operations.
NRC regulations do not specify a maximum time for storing spent fuel in a pool or cask. The agency's waste confidence decision" expresses the Commission's confidence that the fuel can be stored safely in either a pool or cask for at least 60 years beyond the licensed life of any reactor without significant environmental effects. At current licensing terms (40 years of initial reactor operation plus 20 of extended operation), that would amount to up to 120 years of safe storage. However, it is important to note that this does not mean the NRC allows or permits storage for that period. Dry casks are licensed or certified for 20 years, with possible renewals of up to 40 years. This shorter licensing term means the casks are reviewed and inspected, and the NRC ensures the licensee has an adequate aging management program to maintain the facility.
Decommissioning Process Decommissioning involves active cleanup and decontamination, called DECON for short, or a combination of DECON and deferred dismantling, or SAFSTOR. Under DECON, equipment, structures, and portions of the facility containing radioactive contaminants are removed or decontaminated to a level that permits release of the property and termination of the NRC license. Under SAFSTOR, a nuclear power plant is maintained and monitored in a condition that allows the radioactivity to decay; afterwards, the plant shifts to DECON as the facility is dismantled and the property decontaminated. A licensee may combine these two strategies, leaving parts of the facility in SAFSTOR while actively dismantling and decontaminating other parts of the facility. The decision may be based on factors other than radioactive decay, such Enclosure 1
2 as the availability of waste disposal sites. A third strategy, known as ENTOMB, would encase contaminated facilities in concrete. No U.S. commercial nuclear power plant has chosen this option.
Once DECON is complete and the site meets the NRCs cleanup standards, the NRC will terminate the license and release the property for unrestricted use. A portion of the site may remain under NRC license to store the spent fuel. The licensee is responsible for maintaining and protecting the spent fuel storage facility. Decommissioning must be completed within 60 years of the plant ceasing operations. A licensee may choose to decommission the plant sooner than 60 years but this is not a requirement. The NRCs annual report on the status of the decommissioning program can be found in the NRCs Agencywide Documents Access and Management System (ADAMS) in SECY-17-0111 (ADAMS Accession Number ML17276B164).
ADAMS is accessible from the NRC Web page at: http://www.nrc.gov/reading-rm/adams.html.
Additional details on the decommissioning process and the NRCs roles and responsibilities are in NRC Regulatory Guide 1.184, which is available on the NRC public website at www.nrc.gov.
Decommissioning Funds Before a nuclear power plant begins operations, the licensee must establish or obtain a financial mechanism, such as a trust fund or a guarantee from its parent company, to ensure there will be sufficient money to pay for the ultimate decommissioning of the facility. Each nuclear power plant licensee must report to the NRC every two years the status of its decommissioning funding for each reactor or share of a reactor that it owns. These reports are required annually during decommissioning so the NRC can ensure the funds are being used appropriately. The report must estimate the minimum amount needed for decommissioning by using the formulas found in NRC regulations. The staff performs an independent analysis of the decommissioning reports to determine whether licensees are providing reasonable decommissioning funding assurance for radiological decommissioning of the reactor at the permanent termination of operation.
On March 31, 2017, Entergy provided its decommissioning funding status report to the NRC.
This report can be found in NRCs ADAMS with Accession Number ML17093A926. On May 8, 2018, the NRC published its analysis of licensees 2017 decommissioning funding status reports (ADAMS Accession Number ML18122A001). Based on the NRC staff's review of this information, the NRC determined that Indian Point satisfied the decommissioning funding assurance reporting requirements and had provided decommissioning funding assurance as required in the regulations under Part 50.75 of Title 10 of the Code of Federal Regulations.
Post-Shutdown Decommissioning Activities Report and Decommissioning Schedule Within 2 years after submitting the certification of permanent closure, the licensee must submit a post-shutdown decommissioning activities report (PSDAR) to the NRC. This report provides a description of the planned decommissioning activities, a schedule for accomplishing them, and an estimate of the expected costs. The report must discuss the reasons for concluding that environmental impacts associated with the site-specific decommissioning activities have already been addressed in previous environmental analyses. Ninety days after submitting the PSDAR, the licensee may begin major decommissioning activities.
After receiving the PSDAR, the NRC will publish a notice of receipt in the Federal Register, make the report available for public review and comment, and hold a public meeting in the vicinity of the plant to discuss the licensees intentions. At this meeting, members of the public are invited to communicate their needs, concerns, and interests in the decision making process during decommissioning. The regulations regarding the PSDAR are found under Part 50.82 of Title 10 of the Code of Federal Regulations.
3 Release of Part of a Power Reactor Facility or Site for Unrestricted Use Title 10 of the Code of Federal Regulations, Part 50.83, describes the regulations regarding partial release of a nuclear plant site for unrestricted use. If a licensee chooses to pursue this, prior written NRC approval is required to release part of a facility or site for unrestricted use at any time before receiving approval of a license termination plan.
License Transfer During Decommissioning and Public Involvement The NRC maintains regulatory authority and oversight of the licensee (currently, Entergy) both during operation and throughout decommissioning. Should Entergy pursue transferring the license to another company to perform decommissioning, that transfer would require NRC review and approval. Any new licensee, if approved, would be subject to all NRC requirements and regulations. Within 2 years after submitting the certification of permanent closure, the licensee must submit a PSDAR to the NRC. This report provides a description of the planned decommissioning activities, a schedule for accomplishing them, and an estimate of the expected costs. After receiving the report, the NRC will publish a notice of receipt in the Federal Register, make the report available for public review and comment, and hold a public meeting in the vicinity of the plant to discuss the licensees intentions. Specific regulations regarding the PSDAR are found under Part 50.82 of Title 10 of the Code of Federal Regulations. As with all our processes, the NRC encourages the public to attend these meetings and provide input and insights into the process. The NRC encourages licensees to involve the public in their decommissioning process through citizen engagement panels or oversight boards.
Performance Indicators, NRC Findings and Violations Documented at Indian Point Publically available inspection reports, performance indicators, and NRC documented findings are available at the NRCs public webpage:
https://www.nrc.gov/reactors/operating/oversight.html Natural Gas Pipelines Near Indian Point The Algonquin Gas Transmission Company has a 26 inch gas main line and a 30 inch gas main line on a right-of-way running east to west through Entergys property. One 30 inch main and two 24 inch mains pass under the river to a pipeline facilities station on the easement near the river. In the summer of 2013, Spectra (a gas line company) informed Entergy of their intention to pursue surveys for building a new 42-inch gas line on the right-of-way through the southern part of the Indian Point property. This pipeline was completed and placed in service in late December 2016. The new pipeline was routed significantly further away from safety-related systems, structures, and components than the existing gas pipelines at the Indian Point site.
Multiple blast analyses performed by the licensee and the confirmatory analysis performed by the NRC concluded that resultant pressure waves and critical heat flux from a pipeline rupture would not adversely impact the safe shutdown of the Indian Point facility.
High Burn-up Fuel Burn-up is a way to measure the uranium burned in the reactor, and is expressed in gigawatt-days per metric ton of uranium (GWd/MTU). Anything above 45 GWd/MTU is considered high burn-up fuel. High burn-up fuel allows utilities to get more power out of their fuel before replacing it.
Due to security concerns, the NRC does not confirm whether or not a licensee currently uses or has used high-burnup fuel. The NRC has conducted extensive efforts to review the safety of high burn-up fuel. The staff performed an evaluation of data collected and confirmed that the use of fuel up to the existing limits did not pose safety problems. The NRC published Generic Safety Issue 170, Fuel Damage for High Burn-up Fuel, and documented its resolution in
4 NUREG-0933 (Main Report with Supplements 1-34). These documents are publically available on the NRCs website at www.nrc.gov.
Reactor Pressure Vessel O-rings Indian Point Units 2 and 3 use a double, redundant O-ring seal between the reactor pressure vessel (RPV) and the reactor head to maintain reactor coolant system (RCS) integrity. Two metallic O-rings are installed in concentric machined grooves to maintain the boundary integrity of the RPV head. The seal design specifies the inner O-ring to seal RCS pressure and the outer O-ring is a redundant feature. If the inner O-ring leaks, the leak will be detected in a leak off line and an isolation valve will be closed to stop the leakage. Valve closure transfers the RCS pressure seal from the inner O-ring seating surface to the redundant, outer O-ring seal.
This design ensures that only one O-ring seal is in service at a time, restraining RCS pressure.
If the inner O-ring leaks, the RCS water that is trapped between the inner and outer O-ring seal becomes pressurized to RCS pressure and this pressure ensures that the water will be forced through the leak off detection line into a drain tank in containment. Any RCS water that leaks past both O-rings is captured in a sump within the containment building.
O-ring failures have occurred at both Indian Point Units 2 and 3. The most recent failure of the inner and outer RPV O-rings occurred on Unit 2 in December 2017. The recent trend in O-ring failures was previously documented as an annual problem identification and resolution sample in an NRC quarterly inspection report (Indian Point Integrated Inspection Report 05000247/2017003 and 05000286/2017003 (ADAMS Accession No. ML17303A977)). During the Spring 2018 Unit 2 refueling outage, Entergy made repairs to the RPV mating surfaces and replaced the failed O-rings with newly designed O-rings that have greater malleability to better conform to any minor defects in the RPV flange seating surfaces. After restarting the reactor, the O-rings were monitored for leakage and no leakage was identified.
Reactor Pressure Vessel Head Penetration #3 Leak In March 2018, while completing NRC required visual examinations of the Indian Point Unit 2 RPV head, Entergy staff identified very small boron deposits and a leak on penetration #3 (spare penetration). A liquid penetrant examination was performed of the J-groove weld face from under the reactor head. This revealed three rounded flaws with linear aspects.
Entergy proposed to repair the J-groove weld of RPV penetration #3 using an embedded flaw repair process as an alternative to the defect removal and weld repair provisions of American Society of Mechanical Engineers (ASME) Code Section XI, IWA-4000 and ASME Section Ill, NB-4450. The proposed embedded flaw repair process has been previously approved by the NRC, and has been implemented at other sites. This proposed alternative was requested for one-cycle of operation as Entergy intends to cease power operations of Unit 2 upon completion of Cycle 24 (Spring 2020). Verbal approval of the alternate repair method was given on April 9, 2018. Entergy has completed the repair in accordance with the strategy outlined in the relief request. An NRC inspector specializing in RPV head examinations was on site and independently assessed Indian Point staff performance in evaluating and correcting the issue.
1 NRC Staff Annual Assessment Meeting Presentation Enclosure 2
2