Relating ASME Flaw Evaluation to ASME Task Group on Carbon Fiber Repairs

Author(s):  
Tomas Jimenez ◽  
Eric Houston ◽  
Nico Meyer

As most nuclear power stations in the US have surpassed their initial 40 years of operability, the industry is now challenged with maintaining safe operations and extending the operating life of structures, systems and components. The US Nuclear Regulatory Commission (NRC), Nuclear Energy Institute (NEI), and Electric Power Research Institute (EPRI) have identified safety related buried piping systems as particularly susceptible to degradation. These systems are required to maintain the structural factors of the ASME Construction Codes under pressure and piping loads, which includes seismic wave passage. This paper focuses on evaluation approaches for metallic buried piping that can be used to demonstrate that localized thinning meets the requirements of the Construction Code. The paper then addresses a non-metallic repair option using carbon fiber reinforced polymers (CFRP) as the new pressure boundary.

Author(s):  
Tao Zhang ◽  
Frederick W. Brust ◽  
Gery Wilkowski ◽  
Heqin Xu ◽  
Alfredo A. Betervide ◽  
...  

The Atucha II nuclear power plant is a unique pressurized heavy water reactor (PHWR) being constructed in Argentina. The original plant design was by Kraftwerk Union (KWU) in the 1970’s using the German methodology of break preclusion. The plant construction was halted for several decades, but a recent need for power was the driver for restarting the construction. The US NRC developed leak-before-break (LBB) procedures in draft Standard Review Plan (SRP) 3.6.3 for the purpose of eliminating the need to design for dynamic effects that allowed the elimination of pipe whip restraints and jet impingement shields. This SRP was originally written in 1987 with a modest revision in 2005. The United States Nuclear Regulatory Commission (US NRC) is currently developing a draft Regulatory Guide on what is called the Transition Break Size (TBS). However, modeling crack pipe response in large complex primary piping systems under seismic loading is a difficult analysis challenge due to many factors. The initial published work on the seismic evaluations for the Atucha II plant showed that even with a seismic event with the amplitudes corresponding to the amplitudes for an event with a probability of 1e−6 per year, that a Double-Ended Guillotine Break (DEGB) was pragmatically impossible due to the incredibly high leakage rates and total loss of make-up water inventory. The critical circumferential through-wall flaw size in that case was 94-percent of the circumference. This paper discusses further efforts to show how much higher the applied accelerations would have to be to cause a DEGB for an initial circumferential through-wall crack that was 33 percent around the circumference. This flaw length would also be easily detected by leakage and loss of make-up water inventory. These analyses showed that the applied seismic peak-ground accelerations had to exceed 25 g’s for the case of this through-wall-crack to become a DEGB during a single seismic loading event. This is a factor of 80 times higher than the 1e−6 seismic event accelerations, or 240 times higher than the safe shutdown earthquake (SSE) accelerations.


Author(s):  
Joseph S. Miller

The United States utilities started preparing for external events that could lead to a loss of all ac power in the 1980’s, when the Station Blackout (SBO) rulemaking was first introduced by the United States Nuclear Regulatory Commission (USNRC). Following the events at the Fukushima Dai-ichi nuclear power plant on March 11, 2011, the USNRC established a senior-level agency task force referred to as the Near-Term Task Force (NTTF). The NTTF was tasked with conducting a systematic, methodical review of Nuclear Regulatory Commission (NRC) regulations and processes to determine if the agency should make additional improvements to these programs in light of the events at Fukushima Dai-ichi. As a result of this review, the NTTF developed a comprehensive set of recommendations, documented in SECY-11-0093, “Near-Term Report and Recommendations for Agency Actions Following the Events in Japan,” dated July 12, 2011. Documentation of the staff’s efforts is contained in SECY-11-0124, “Recommended Actions to be Taken without Delay from the Near-Term Task Force Report,” dated September 9, 2011, and SECY-11-0137, “Prioritization of Recommended Actions to be Taken in Response to Fukushima Lessons Learned,” dated October 3, 2011. To satisfy some of the NRC’s recommendations, the industry described its proposal for a Diverse and Flexible Mitigation Capability (FLEX), as documented in Nuclear Energy Institute’s (NEI) letter, dated December 16, 2011 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML11353A008). FLEX was proposed as a strategy to fulfill the key safety functions of core cooling, containment integrity, and spent fuel cooling. The events at Fukushima Dai-ichi highlight the possibility that extreme natural phenomena could challenge the prevention, mitigation and emergency preparedness defense-in-depth layers. At Fukushima, limitations in time and unpredictable conditions associated with the accident significantly challenged attempts by the responders to preclude core damage and containment failure. During the events in Fukushima, the challenges faced by the operators were beyond any faced previously at a commercial nuclear reactor. NRC Order 12-049 (Ref. 1) and NRC Interim Staff Guidance JLD-ISG-2012-01 (Ref. 6) provided additional requirements to mitigate beyond-design-basis external events. These additional requirements impose guidance and strategies to be available if the loss of power, motive force and normal access to the ultimate heat sink to prevent fuel damage in the reactor and spent fuel pool affected all units at a site simultaneously. The NEI submitted document NEI 12-06, “Diverse and Flexible Coping Strategies (FLEX) Implementation Guide” in August 2012 (ADAMS Accession No. ML12242A378) to provide specifications for the nuclear power industry in the development, implementation, and maintenance of guidance and strategies in response to NRC Order EA-12-049. The US utilities are currently proposing modifications to their plants that will follow specifications provided in NEI 12-06. This paper presents some of the NEI 12-06 requirements and some of the proposed modifications proposed by the US utilities.


Author(s):  
J. G. Merkle ◽  
K. K. Yoon ◽  
W. A. VanDerSluys ◽  
W. Server

ASME Code Cases N-629/N-631, published in 1999, provided an important new approach to allow material specific, measured fracture toughness curves for ferritic steels in the code applications. This has enabled some of the nuclear power plants whose reactor pressure vessel materials reached a certain threshold level based on overly conservative rules to use an alternative RTNDT to justify continued operation of their plants. These code cases have been approved by the US Nuclear Regulatory Commission and these have been proposed to be codified in Appendix A and Appendix G of the ASME Boiler and Pressure Vessel Code. This paper summarizes the basis of this approach for the record.


Author(s):  
Christopher T. Kupper ◽  
Mark A. Gray

In NUREG-1801 (GALL) Revision 0 and Revision 1, the US Nuclear Regulatory Commission (NRC) defined the locations evaluated in NUREG/CR-6260 as a minimum acceptable set for evaluation of environmentally assisted fatigue (EAF), in addressing license renewal for nuclear plant components. Within GALL Revision 2, the NRC revised the expectation, so that plants also investigate the possibility of other locations being more limiting. To address GALL Revision 2 and NUREG-1800 Revision 2, an EAF screening methodology was developed that considers all Safety Class 1 reactor coolant pressure boundary components in major equipment and piping systems that are susceptible to EAF, including those locations listed in NUREG/CR-6260. While the overall screening process steps are similar to those published by EPRI, elements of the detailed application of some steps were performed using alternative techniques. The screening process utilized the comprehensive database of plant component fatigue qualifications available in NSSS vendor documentation, and yielded a comprehensive list of lead indicator locations for EAF consideration. This paper describes the overall process and alternate methods in the context of a specific plant license renewal application.


2016 ◽  
Vol 2016 ◽  
pp. 1-9 ◽  
Author(s):  
Andrija Volkanovski ◽  
Antonio Ballesteros Avila ◽  
Miguel Peinador Veira

This paper presents the results of the statistical analysis of the loss of offsite power events (LOOP) registered in four reviewed databases. The reviewed databases include the IRSN (Institut de Radioprotection et de Sûreté Nucléaire) SAPIDE database and the GRS (Gesellschaft für Anlagen- und Reaktorsicherheit mbH) VERA database reviewed over the period from 1992 to 2011. The US NRC (Nuclear Regulatory Commission) Licensee Event Reports (LERs) database and the IAEA International Reporting System (IRS) database were screened for relevant events registered over the period from 1990 to 2013. The number of LOOP events in each year in the analysed period and mode of operation are assessed during the screening. The LOOP frequencies obtained for the French and German nuclear power plants (NPPs) during critical operation are of the same order of magnitude with the plant related events as a dominant contributor. A frequency of one LOOP event per shutdown year is obtained for German NPPs in shutdown mode of operation. For the US NPPs, the obtained LOOP frequency for critical and shutdown mode is comparable to the one assessed in NUREG/CR-6890. Decreasing trend is obtained for the LOOP events registered in three databases (IRSN, GRS, and NRC).


Author(s):  
Thomas S. LaGuardia

The US Nuclear Regulatory Commission (NRC) established criteria for acceptable residual radioactivity related to decommissioning nuclear power plants in the US [1]. A level of 25 mRem per year to the maximum exposed individual by site-specific pathways analysis, with ALARA is acceptable to the NRC. Systems and structures containing very low levels of radioactivity that meet this criteria are deemed acceptable to abandon in place as part of the decommissioning process and termination of the license. Upon license termination by the NRC, the owner may then demolish and remove remaining structures. In practice, site-specific criteria imposed by local state mandates, company management decisions, real estate value, and long-term liability potential have driven nuclear plant licensees to adopt an alternative disposition for these materials. Although the reasons are different at each site, the NRC’s criteria of 25 mRem per year are not the controlling factor. This paper will describe the regulatory process for termination of the license, and the other factors that drive the decision to remove radioactive and non-radioactive material for decommissioning. Several case histories are presented to illustrate that the NRC’s criteria for license termination are not the only consideration.


2000 ◽  
Vol 122 (3) ◽  
pp. 234-241 ◽  
Author(s):  
Owen F. Hedden

This article will describe the development of Section XI from a pamphlet-sized document to the lengthy and complex set of requirements, interpretations, and Code Cases that it has become by the year 2000. Section XI began as a set of rules for inservice inspection of the primary pressure boundary system of nuclear power plants. It has evolved to include other aspects of maintaining the structural integrity of safety class pressure boundaries. These include procedures for component repair/replacement activities, analysis of revised and new plant operating conditions, and specialized provisions for nondestructive examination of components and piping. It has also increased in scope to cover other Section III construction: Class 2, Class 3 and containment structures. First, to provide a context for the discussions to follow, the differences in administration and enforcement between Section XI and the other Code Sections will be explained, including its dependence on the US Nuclear Regulatory Commission. The importance of interpretations and Code Cases then will be discussed. The development of general requirements and requirements for each class of structure will be traced. The movement of Section XI toward a new philosophy, risk-informed inspection, will also be discussed. Finally, an annotated bibliography of papers describing the philosophy and technical basis behind Section XI will be provided. [S0094-9930(00)01703-0]


Author(s):  
Taunia Wilde ◽  
Shannan Baker ◽  
Gary M. Sandquist

The design, construction, operation, maintenance, and decommissioning and decontamination of nuclear infrastructure particularly nuclear power plants licensed in the US by the US Nuclear Regulatory Commission (NRC) or operated by the US Department of Energy (DOE) or the US Department of Defense (DOD) must be executed under a rigorous and documented quality assurance program that provides adequate quality control and oversight. Those codes, standards, and orders regulate, document and prescribe the essentials for quality assurance (QA) and quality control (QC) that frequently impact nuclear facilities operated in the US are reviewed and compared.


Author(s):  
Ma Chao ◽  
Deng Wei ◽  
An Jin

Maintenance effectiveness is important for the safety and power production of Nuclear Power Plants (NPP). U.S. Nuclear Regulatory Commission (NRC) Maintenance Rule (10CFR50.65, MR) became effective in 1996, and it is mandatory for all the US plants to use Maintenance Rule in their daily maintenance activities. With the development and wide usage of Probabilistic Risk Assessment (PRA) technique in China, China regulator and utilities are trying to adopt MR in maintenance activities. Brief study on application of MR in some VVER-typed China nuclear plant is carried out and some main results are shown. All the application process and results will be useful for later official application of MR in China.


2013 ◽  
Vol 135 (5) ◽  
Author(s):  
Tao Zhang ◽  
Frederick W. Brust ◽  
Gery Wilkowski ◽  
Heqin Xu ◽  
Alfredo A. Betervide ◽  
...  

The Atucha II nuclear power plant is a unique pressurized heavy water reactor (PHWR) being constructed in Argentina. The original plant design was by Kraftwerk Union (KWU) in the 1970's using the German methodology of break preclusion. The plant construction was halted for several decades, but a recent need for power was the driver for restarting the construction. The United States Nuclear Regulatory Commission (US NRC) developed leak-before-break (LBB) procedures in standard review plan (SRP) 3.6.3 Revision 1 for the purpose of eliminating the need to design for dynamic effects that allowed the elimination of pipe whip restraints and jet impingement shields. This SRP was originally written in 1987. The US NRC is currently developing a draft Regulatory Guide on what is called the transition break size (TBS). However, modeling crack pipe response in large complex primary piping systems under seismic loading is a difficult analysis challenge due to many factors. The initial published work (Wilkowski et al., “Robust LBB Analysis for Atucha II Nuclear Plant,” 2011 ASME PVP Conference, July 17–21, Baltimore, MD) on the seismic evaluations for the Atucha II plant showed that even with a seismic event with the amplitudes corresponding to the amplitudes for an event with a probability of 1 × 10−6 per year, that a double-ended guillotine break (DEGB) was pragmatically impossible due to the high leakage rates and total loss of make-up water inventory. The critical circumferential through-wall flaw size in that case was 94% of the circumference. This paper discusses further efforts to show how much higher the applied accelerations would have to be to cause a DEGB for an initial circumferential through-wall crack that was 33% around the circumference. This flaw length would also be easily detected by leakage and loss of make-up water inventory. These analyses showed that the applied seismic peak-ground accelerations had to exceed 25 g's for the case of this through-wall-crack to become a DEGB during a single seismic loading event. This is a factor of 80 times higher than the 1 × 10−6 seismic event accelerations, or 240 times higher than the safe shutdown earthquake (SSE) accelerations.


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