ML082660002
| ML082660002 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 09/19/2008 |
| From: | Charemagne Grimes - No Known Affiliation |
| To: | Atomic Safety and Licensing Board Panel |
| SECY RAS | |
| Shared Package | |
| ML082660029 | List: |
| References | |
| 50-282-LR, 50-306-LR, ASLBP 08-871-01-LR-BD01, RAS 1152 | |
| Download: ML082660002 (8) | |
Text
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )
)
NUCLEAR MANAGEMENT COMPANY, LLC ) Docket Nos. 50-282-LR
) 50-306-LR (Prairie Island Nuclear Generating Plant, )
Units 1 and 2) ) ASLBP No. 08-871-01-LR
)
DECLARATION OF CHRISTOPHER I. GRIMES
- 1. My name is Christopher I. Grimes. I am currently a Senior Nuclear Safety Consultant. The Prairie Island Indian Community in the State of Minnesota has retained me as a consultant with respect to the above-captioned proceeding. I hold a B.S. in Nuclear Engineering from Oregon State University.
- 2. I retired from the United States Nuclear Regulatory Commission (NRC) in June 2006 after 32 years of Federal Service. My career at the NRC included a broad range of positions of increasing responsibility covering most aspects of nuclear power regulation. At the time of my retirement, I served as the Director of the Policy and Rulemaking Division in the NRCs Office of Nuclear Reactor Regulation. In that position, I was responsible for all of the reactor-related rulemaking activities, financial assurance, regulatory analysis, generic communications, generic project management, interoffice coordination, licensing processes, and all of the licensing and inspection activities associated with research and test reactors.
- 3. During the course of my career at the NRC I managed the completion of the NRCs Systematic Evaluation Program, the licensing of the Comanche Peak nuclear
power plant, issuance of the improved Standard Technical Specifications and implementation of the Technical Specification Improvements Program, development and implementation of the NRCs license renewal process, and improved application of risk-informed decisions in the regulatory process.
- 4. I am knowledgeable of and experienced in nuclear reactor safety. As a reactor engineer and systems analyst, I am familiar with a broad variety of reactor designs. I have been responsible for performing containment response analysis, evaluating reactor system designs and preparing safety evaluation reports for construction permit and operating license applications. I have contributed to the development of a computer code for analyzing containment subcompartment pressurization using compressible fluid flow theory. I have served as the Task Manager for the Mark I containment Long Term Program to resolve pool dynamic loads in boiling water reactor designs. I have served as Emergency Officer in the NRCs Incident Response Program.
I was qualified as an Incident Investigation Team Leader. I served as Team Leader for the Oyster Creek Diagnostic Evaluation Inspection Team.
- 5. I am knowledgeable of and experienced in nuclear reactor safety management. I was appointed to the senior executive service in 1984, upon selection as the Chief of the Systematic Evaluation Program (SEP). That program evaluated ten of the oldest power reactors against current requirements and used risk insights in integrated safety assessments to develop backfitting recommendations. I was responsible for directing the safety reviews and developing proposed staff positions to resolve the safety issues. I served as Deputy Director in the Division of Engineering from 2003 to 2005. In that position, I was responsible for directing engineering-related safety evaluations of 2
licensees implementation of NRC requirements, changes to existing license requirements, and applications for new facilities or designs. I was also responsible for directing the application of engineering expertise to support special inspections, projects, programs, and policy activities in the areas of mechanical, civil-structural, materials, metallurgy, chemical, instrumentation and control systems, and electrical engineering, as well as applying that engineering expertise to conduct failure analysis, structural analysis, and represented the NRC on domestic and international codes and standards groups.
- 6. I am knowledgeable of and experienced in nuclear reactor license renewal and environmental impacts. I served as Director of the License Renewal and Environmental Impacts Program from 1997 to 2002. In that position, I was responsible for developing and implementing the license renewal review process for power reactors based on the requirements which were codified in 10 CFR Part 54 in 1995. I was responsible for establishing the plans and schedules for the first license renewal reviews, as well as developing the review standards for the associated environmental reviews.
Upon completion of the first three renewed licenses, I established a five-year schedule of license renewal reviews and implemented a process to manage changes to the license renewal review guides and related staff positions.
- 7. I am knowledgeable of and experienced in rulemaking and regulatory analysis. I was appointed as Director of the Policy and Rulemaking Division in 2005. In that position, I was responsible for all of the reactor-related rulemaking activities, financial assurance, regulatory analysis, generic communications, generic project management, interoffice coordination, licensing processes, and all of the licensing and inspection activities associated with research and test reactors.
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- 8. I am very familiar with the operation of, and safety analyses associated with, pressurized water nuclear reactors (PWRs), including the reactor design currently in operation at the Prairie Island Nuclear Generating Plant (PINGP) located near Red Wing, Minnesota.
- 9. I have reviewed the license renewal application (LRA) for the two PINGP reactors that was submitted by Nuclear Management Company, LLC (NMC). These reactors are known as Prairie Island Units 1 and 2. In my opinion, and as I explain more fully below, the NRC should grant a hearing on at least four aging management and safety issues: (1) the LRA does not describe an adequate plan to monitor and manage the effects of aging for containment coatings; (2) the LRA does not describe an adequate plan to monitor and manage the effects of aging due to embrittlement of the reactor pressure vessels internals; (3) the LRA does not describe an adequate aging management program for managing primary stress corrosion cracking for nickel-alloy components; and (4) the LRA does not describe an adequate inspection and monitoring for corrosion or leaks in all buried systems, structures, and components that may convey or contain radioactively-contaminated water or other fluids and/or may be important to plant safety.
- 10. The elements of an effective aging management program, pursuant to 10 C.F.R. §§ 54.21(a)(3), are described in the NRCs Standard Review Plan for License Renewal, NUREG-1800, Section A.1.2.2 Aging Management Program for License Renewal. Throughout the development of the Generic Aging Lessons Learned Report and the review guidance for license renewal, these ten elements were used to establish the effectiveness of aging management programs. Among the elements of an effective aging 4
management program are the ability to detect degradation occurring over time, as well as plans to prevent and/or mitigate that degradation.
- 11. The operation of the emergency core cooling systems (ECCS) depends on the ability to draw cooling water from the containment sump after a loss-of-coolant accident (LOCA) has occurred. The NRC issued Generic Letter 2004-02 (GL 04-02) to ensure that licensees carefully evaluated sources of debris in the containment and ensured that debris blockage of the sump screens would not prevent proper operation of the ECCS. A significant source of debris is the coatings on structures and components inside containment and on the interior surface of the containment. Operating experience has demonstrated that, if not managed well, these coatings can begin peeling off, which increases the potential for forming debris. In their response to GL 04-02, NMC describes how the containment inservice inspection program can provide a means to monitor the condition of coatings. However, for license renewal, NMC stated that the coatings have no intended function. The blowdown forces associated with a LOCA will cause a certain amount of debris as a result of the jet impingement on coatings, insulation and adjacent light structures. The amount of debris that is formed can be minimized by a good housekeeping program for loose materials, well designed and maintained insulation materials, and a condition monitoring and maintenance program for coatings.
- 12. I have reviewed the declaration of Dr. Richard T. Lahey, Jr., Professor of Engineering at the Rensselaer Polytechnic Institute in Troy, New York, that was submitted with the Petition to Intervene in the license renewal application for Indian Point Units 2 and 3 (IP2 and IP3). While the Indian Point plant has a different PWR design and site, there are aspects of Dr. Laheys concerns that are applicable to PINGP 5
Units 1 and 2. In particular, Dr. Lahey describes how neutron bombardment, or fluence, causes embrittlement of the reactor vessel and internals. While PINGP Units 1 and 2 do not have the same belt-line conditions Dr. Lahey describes for IP2 and IP3, the concerns about the adequacy of the monitoring and aging management program for reactor vessel internals is applicable. If the core support structure fails during the loading conditions resulting from a design basis accident or transient, the resulting core geometry could not be cooled by the Emergency Core Cooling Systems as the design intended. NMC describes a commitment to develop and implement the PWR Vessel Internals Program in Section B.2.1.32 of the LRA. However, the program description lacks sufficient detail to determine whether it can manage the effects of embrittlement for the period of extended operation.
- 13. Nickel-alloy components are susceptible to primary water stress corrosion cracking (PWSCC). As part of the efforts to improve the methods to identify and maintain a broad range of materials degradation, the industry undertook a Materials Reliability Program (MRP) which included, among other things, plans to develop augmented inspection methods to detect and correct PWSCC in nickel-alloy components.
After continued incidents of cracking in nickel-alloy welds in reactor head penetrations and significant degradation was discovered in the reactor vessel head at Davis Besse, the NRC issued an order EA-03-009, Issue of Order Establishing Interim Inspection Requirements for Reactor Pressure Vessel Heads at Pressurized Water Reactors. That Order established the interim inspection requirements until the generic MRP efforts develop augmented inspection and repair practices. The NRCs Interim Staff Guidance for License Renewal for the aging management program for PWSCC in nickel-alloy 6
components (LR-ISG-19B) states that the program is under development consistent with the interim inspection requirements in the Order. In the LRA, NMC simply explains how guidance is under development, they will comply with applicable NRC Orders and implement applicable NRC Bulletins, Generic Letters and staff-accepted industry guidelines. The LRA does not explain how the existing interim inspection requirements satisfy the requirements of an effective aging management program.
- 14. Degradation of buried and inaccessible systems, structures and components is difficult to manage because of the limited accessibility. Moreover, because of recent events involving unplanned, unmonitored releases of radioactive liquids into the environment, the NRC established a Liquid Radioactive Release Lessons Learned Task Force (LLTF) to review the industry experience and public health impacts.
The LLTF noted that leakage that enters the ground below the plant may be undetected because there are generally no NRC requirements to monitor the groundwater onsite for radioactive contamination. The LRA describes a variety of buried tanks and systems. In Section B2.1.8, NMC describes the Buried Piping and Tanks Inspection Program.
The program does not commit to conduct any inspections of buried tanks or systems to establish baseline conditions to evaluate the effectiveness of the program in the future, and only commits to conduct inspections if the opportunity arises, with at least one inspection occurring within ten years. The LRA states that the program is not applicable for some systems because there are no buried components or piping. For those components and systems for which the program is applicable, it is not clear whether the components and systems normally contain radioactive liquid or might contain radioactive liquid as a result of an accident or transient. Significant degradation could already exist 7
in those buried components and systems for which the program applies. The LRA does not explain how the proposed program satisfies the elements of an effective aging management program.
I declare under penalty of perjury that the foregoing is true and correct.
Executed this 19th day of September, 2008, at Rockville, Maryland.
/Executed by Christopher I. Grimes in Accord with 10 C.F.R. 2.304(d)/
Christopher I. Grimes Senior Nuclear Safety Consultant 27 North Shore Drive Swanton, Maryland 21561-2215 Phone: 240-678-0318 Email: CandCGrimes@aol.com State of Maryland County of Montgomery 8