Safety Evaluation of the Hydrogen Flaking Damage in the Doel 3 and Tihange 2 Reactor Pressure Vessels

2016 ◽  
Author(s):  
Guy Roussel
Keyword(s):  
Safety Evaluation ◽  
Pressure Vessels ◽  
Reactor Vessel ◽  
Pressurized Water Reactors ◽  
Pressurized Water ◽  
Significant Safety ◽  
Nuclear Control ◽  
Safety Case ◽  
Water Reactors ◽  
Reactor Pressure

In the summer of 2012, the detection of a large number of quasi-laminar flaw indications in the reactor vessel beltline ring forgings of two Belgian pressurized water reactors (Doel Unit 3 and Tihange Unit 2) posed a significant safety threat that led the licensee to shutdown both plants. Those indications were identified by the licensee as hydrogen flakes that developed during the fabrication of the forgings. As a prerequisite for a potential restart of the units, the Belgian Nuclear Safety Regulator, the Federal Agency for Nuclear Control (FANC), requested the licensee to provide, for each unit, a safety case demonstrating the acceptability of the reactor pressure vessel for continued operation. As the technical subsidiary of the FANC, Bel V performed a safety evaluation of the condition of the reactor pressure vessels. The paper documents the approach Bel V used in his safety evaluation and the criteria he defined to evaluate the acceptability of the hydrogen flaking damage in the reactor pressure vessels.

Background on Low Temperature Overpressure Protection System Setpoints for Pressure-Temperature Curves

2020 ◽  
Author(s):  
Jeffrey C. Poehler ◽  
Gary L. Stevens ◽  
Anees A. Udyawar ◽  
Amy Freed
Keyword(s):  
Low Temperature ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Pressure Vessels ◽  
Pressurized Water Reactors ◽  
Documentation System ◽  
Pressurized Water ◽  
System Pressure ◽  
Water Reactors

Abstract ASME Code, Section XI, Nonmandatory Appendix G (ASME-G) provides a methodology for determining pressure and temperature (P-T) limits to prevent non-ductile failure of nuclear reactor pressure vessels (RPVs). Low-Temperature Overpressure Protection (LTOP) refers to systems in nuclear power plants that are designed to prevent inadvertent challenges to the established P-T limits due to operational events such as unexpected mass or temperature additions to the reactor coolant system (RCS). These systems were generally added to commercial nuclear power plants in the 1970s and 1980s to address regulatory concerns related to LTOP events. LTOP systems typically limit the allowable system pressure to below a certain value during plant operation below the LTOP system enabling temperature. Major overpressurization of the RCS, if combined with a critical size crack, could result in a brittle failure of the RPV. Failure of the RPV could make it impossible to provide adequate coolant to the reactor core and result in a major core damage or core melt accident. This issue affected the design and operation of all pressurized water reactors (PWRs). This paper provides a description of an investigation and technical evaluation regarding LTOP setpoints that was performed to review the basis of ASME-G, Paragraph G-2215, “Allowable Pressure,” which includes provisions to address pressure and temperature limitations in the development of P-T curves that incorporate LTOP limits. First, high-level summaries of the LTOP issue and its resolution are provided. LTOP was a significant issue for pressurized water reactors (PWRs) starting in the 1970s, and there are many reports available within the U.S. Nuclear Regulatory Commission’s (NRC’s) documentation system for this topic, including Information Notices, Generic Letters, and NUREGs. Second, a particular aspect of LTOP as related to ASME-G requirements for LTOP is discussed. Lastly, a basis is provided to update Appendix G-2215 to state that LTOP setpoints are based on isothermal (steady-state) conditions. This paper was developed as part of a larger effort to document the technical bases behind ASME-G.

Evaluation of Laser Peening for Mitigation of Primary Water Stress Corrosion Cracking in Pressurized Water Reactors

2019 ◽  
Author(s):  
Stephen Marlette ◽  
Stan Bovid
Keyword(s):  
Surface Stress ◽  
Reactor Vessel ◽  
Pressurized Water Reactors ◽  
Laser Peening ◽  
Alloy 600 ◽  
Test Results ◽  
Pressurized Water ◽  
Performance Requirements ◽  
Primary Water ◽  
Water Reactors

Abstract For several decades pressurized water reactors have experienced Primary Water Stress Corrosion Cracking (PWSCC) within Alloy 600 components and welds. The nuclear industry has developed several methods for mitigation of PWSCC to prevent costly repairs to pressurized water reactor (PWR) components including surface stress improvement by peening. Laser shock peening (LSP) is one method to effectively place the surface of a PWSCC susceptible component into compression and significantly reduce the potential for crack initiation during future operation. The Material Reliability Program (MRP) has issued MRP-335, which provides guidelines for effective mitigation of reactor vessel heads and nozzles constructed of Alloy 600 material. In addition, ASME Code Case N-729-6 provides performance requirements for peening processes applied to reactor vessel head penetrations in order to prevent degradation and take advantage of inspection relief, which will reduce operating costs for nuclear plants. LSP Technologies, Inc. (LSPT) has developed and utilized a proprietary LSP system called the Procudo® 200 Laser Peening System. System specifications are laser energy of 10 J, pulse width of 20 ns, and repetition rate of 20 Hz. Scalable processing intensity is provided through automated focusing optics control. For the presented work, power densities of 4 to 9.5 GW/cm2 and spot sizes of nominally 2 mm were selected. This system has been used effectively in many non-nuclear industries including aerospace, power generation, automotive, and oil and gas. The Procudo® 200 Laser Peening System will be used to process reactor vessel heads in the United States for mitigation of PWSCC. The Procudo® 200 Laser Peening System is a versatile and portable system that can be deployed in many variations. This paper presents test results used to evaluate the effectiveness of the Procudo® 200 Laser Peening System on Alloy 600 material and welds. As a part of the qualification process, testing was performed to demonstrate compliance with industry requirements. The test results include surface stress measurements on laser peened Alloy 600, and Alloy 182 coupons using x-ray diffraction (XRD) and crack compliance (slitting) stress measurement techniques. The test results are compared to stress criteria developed based on the performance requirements documented in MRP-335 and Code Case N-729-6. Other test results include surface roughness measurements and percent of cold work induced by the peening process. The test results demonstrate the ability of the LSP process to induce the level and depth of compression required for mitigation of PWSCC and that the process does not result in adverse conditions within the material.

Deterministic and Probabilistic Fracture Mechanics Analysis for Structural Integrity Assessment of Pressurized Water Reactor Pressure Vessel

2016 ◽  
Vol 138 (3) ◽  
Author(s):  
Kuan-Rong Huang ◽  
Chin-Cheng Huang ◽  
Hsiung-Wei Chou
Keyword(s):  
Structural Reliability ◽  
Pressure Vessels ◽  
Reactor Vessel ◽  
Benchmark Problems ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Oak Ridge ◽  
Integrity Assessment ◽  
Reactor Pressure

Cumulative radiation embrittlement is one of the main causes for the degradation of pressurized water reactor (PWR) reactor pressure vessels (RPVs) over their long-term operations. To assess structural reliability of degraded reactor vessels, the FAVOR code from the Oak Ridge National Laboratories of the U.S. is employed to perform probabilistic fracture analysis for existing Taiwan domestic PWR reactor vessels with consideration of irradiation embrittlement effects. The plant specific parameters of the analyzed reactor vessel associated with assumed design transients are both considered as the load conditions in this work. Furthermore, two overcooling transients of steam generator tube rupture (SGTR) and pressurized thermal shock (PTS) addressed in the USNRC/EPRI benchmark problems are also taken into account. The computed low failure probabilities can demonstrate the structural reliability of the analyzed reactor vessel for its both license base and extended operations. This work is helpful for the risk assessment and aging management of operating PWR RPVs and can also be referred as its regulatory basis in Taiwan.

Crack Initiation and Arrest of Irradiated RPV-Steels: A National Project on German RPV Weld and Base Materials

2004 ◽  
Author(s):  
Elisabeth Keim ◽  
Reinhard Langer ◽  
Hilmar Schnabel ◽  
Hieronymus Hein
Keyword(s):  
Crack Initiation ◽  
Safety Assessment ◽  
Master Curve ◽  
Charpy Impact ◽  
Pressurized Water Reactors ◽  
Charpy Impact Energy ◽  
Pressurized Water ◽  
Base Materials ◽  
Water Reactors ◽  
Reactor Pressure

In Germany the procedure which has to be applied for the safety assessment of the reactor pressure vessel is based on the RTNDT concept. The Master Curve concept (based on T0) has the advantage compared to the RTNDT concept that the basic tests are fracture toughness tests instead of Charpy impact energy or Pellini tests. By means of the recently initiated German project CARISMA (Crack Initiation and Arrest of Irradiated Steel Materials), a data base will be created on pre-irradiated original materials of the four generations of German nuclear pressurized water reactors, which allows to examine the consequences if the Master Curve instead of the RTNDT concept will be applied.

Situation of the Baffle-Former Bolts in Belgian Units

2009 ◽  
Author(s):  
Robert Ge´rard ◽  
Fre´de´ric Somville
Keyword(s):  
Stress Corrosion ◽  
Water Flow ◽  
Stress Corrosion Cracking ◽  
Structural Integrity ◽  
Operating Time ◽  
Reactor Vessel ◽  
Pressurized Water Reactors ◽  
Pressurized Water ◽  
Fuel Assemblies ◽  
Water Reactors

The baffle to former bolts are used in Pressurized Water Reactors to attach the baffle plates to the former plates in the reactor vessel lower internals. The resulting structure forms a boundary for the flow of coolant and provides lateral support to the fuel assemblies. Some edge bolts are also present, assembling together the baffle plates. After an operating time of the order of 120 000 hours, some bolts exhibit cracking at the junction of the head and the shaft of the bolt. Examinations of failed bolts have made it possible to identify the cause of cracking as irradiation assisted stress corrosion cracking (IASCC). Up to now, baffle bolt cracking has been detected in units older than 15 years, where the baffle bolts are not cooled (no holes in the former to allow a water flow on the bolt shaft). In Belgium the concerned unit are Tihange 1 and Doel 1–2. The paper summarizes the experience with baffle bolts cracking in Belgian units and the strategy implemented to mitigate this problem, consisting of structural integrity analyses, baffle bolts inspections and replacement, and research programs in the field of IASCC, including examinations of highly irradiated replaced bolts.

Probabilistic Fracture Mechanics Analysis for Degraded Reactor Pressure Vessel in Pressurized Water Reactor Nuclear Power Plant

2014 ◽  
Author(s):  
Kuan-Rong Huang ◽  
Chin-Cheng Huang ◽  
Hsoung-Wei Chou
Keyword(s):  
Nuclear Power ◽  
Structural Reliability ◽  
The United States ◽  
Pressure Vessels ◽  
Reactor Vessel ◽  
Benchmark Problems ◽  
Irradiation Embrittlement ◽  
Pressurized Water ◽  
Oak Ridge ◽  
Reactor Pressure

Cumulative radiation embrittlement is one of the main causes for the degradation of PWR reactor pressure vessels over their long term operations. To assess structural reliability of degraded reactor vessels, the FAVOR code from the Oak Ridge National Laboratories of the United States is employed to perform probabilistic fracture analysis for existing Taiwan domestic PWR reactor vessels with consideration of irradiation embrittlement effects. The plant specific parameters of the analyzed reactor vessel associated with assumed design transients are both considered as the load conditions in this work. Further, two overcooling transients of steam generator tube rupture and pressurized thermal shock addressed in the USNRC/EPRI benchmark problems are also taken into account. The computed low failure probabilities can demonstrate the structural reliability of the analyzed reactor vessel for its both license base and extended operations. This work is helpful for the risk assessment and aging management of operating PWR reactor pressure vessels and can be also referred as its regulatory basis in Taiwan.

Aging Management of Reactor Coolant System Mechanical Components in Pressurized Water Reactors for License Renewal

2002 ◽  
Author(s):  
M. Subudhi ◽  
R. Morante ◽  
A. D. Lee
Keyword(s):  
Reactor Vessel ◽  
Pressurized Water Reactors ◽  
Pressurized Water ◽  
Reactor Coolant ◽  
Management Review ◽  
Mechanical Components ◽  
Water Reactors ◽  
Effects Of Aging ◽  
License Renewal ◽  
Coolant System

The reactor coolant system (RCS) mechanical components in pressurized water reactors (PWRs) that require an aging management review for license renewal include the primary loop piping and associated connections to other support systems, reactor vessel, reactor vessel internals, pressurizer, steam generators, reactor coolant pumps, and all other inter-connected piping, pipe fittings, valves, and bolting. All major RCS components are located inside the reactor building. Based on the evaluation findings of recently submitted license renewal applications for pressurized water reactors, this paper presents the plant programs and/or activities proposed by the applicants to manage the effects of aging. These programs and/or activities provide reasonable assurance that the intended function(s) of these mechanical components will be maintained for the period of extended operation. The license renewal application includes identification of RCS subcomponents that are within the scope of license renewal and are vulnerable to age-related degradation when exposed to environmental and operational conditions, determination of the effects of aging on their intended safety functions, and implementation of the aging management programs and/or activities including both current and new programs. Industry-wide operating experience, including generic communication by the NRC, is part of the aging management review for the RCS components. This paper presents a number of generic issues, including the time-limited aging analyses, associated with RCS components that require further review by the staff.

Failure Frequency Estimates for Boiling Water Reactor Pressure Vessels

2010 ◽  
Author(s):  
F. A. Simonen ◽  
S. R. Gosselin ◽  
J. E. Rhoads ◽  
A. T. Chiang
Keyword(s):  
Pressure Vessels ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Thermal Transients ◽  
Core Damage ◽  
Water Reactors ◽  
Vessel Rupture ◽  
Reactor Pressure

This paper reviews estimates of rupture frequencies for reactor pressure vessels (RPV) at boiling water reactor (BWR) nuclear power plants as reported in the literature. Results permit improved probabilistic risk assessments (PRA) for severe accidents that could cause core damage and/or challenge the capabilities of BWR reactor containment systems. Current and historical estimates of failure frequencies are considered for light water reactors in general and more specifically for BWR plants. The focus is on large ruptures that could give flow rates exceeding the rates associated with double ended breaks of large diameter recirculation piping. Rupture frequencies for BWR vessels as used for PRA evaluations have historically been assigned low values (e.g. 10−7 to 10−6 per vessel per year). The objective of the present work was to establish possible technical bases for more realistic values of rupture frequency (i.e. 10−8). Historical estimates from the early WASH-1400 reactor safety study were first reviewed and used as a point-of-reference. More recent estimates came from various sources such as a U.S. Nuclear Regulatory Commission expert elicitation process that estimated Loss-of-Coolant Accident (LOCA) frequencies. Other studies both by industry and by the USNRC have addressed rupture frequencies for BWR vessels subject to low-temperature-over-pressure (LTOP) events. On the other hand, recent comprehensive evaluations have focused mostly on RPV failure frequencies for pressurized water reactors (PWRs) caused by pressurized thermal shock events. An important consideration was that rupture frequencies for BWR vessels are believed to be lower than those for PWR vessels, because BWR vessels are less embrittled than PWR vessels and are subject to less severe thermal transients. The review concludes that prior studies support an estimate of 10−8 or less for BWR vessel rupture frequencies. Probabilistic fracture mechanics calculations for individual vessels accounting for plant specific conditions are recommended to support even lower estimated frequencies. Use of more realistic vessel rupture frequencies in a plant’s PRA provides an improvement in not only the perceived plants risk of core damage, but also provides better decision making for plant operation and maintenance activities in that a conservative initiating event treatment within a PRA can mask other initiating events of higher importance.

Supplemental Stress and Fracture Mechanics Analyses of Pressurized Water Reactor Pressure Vessel Nozzles

2012 ◽  
Author(s):  
Matthew Walter ◽  
Shengjun Yin ◽  
Gary L. Stevens ◽  
Daniel Sommerville ◽  
Nathan Palm ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Pressure Vessel ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Additional Stress ◽  
Corner Crack ◽  
Pressurized Water ◽  
The Core ◽  
Water Reactors ◽  
Reactor Pressure

In past years, the authors have undertaken various studies of nozzles in both boiling water reactors (BWRs) and pressurized water reactors (PWRs) located in the reactor pressure vessel (RPV) adjacent to the core beltline region. Those studies described stress and fracture mechanics analyses performed to assess various RPV nozzle geometries, which were selected based on their proximity to the core beltline region, i.e., those nozzle configurations that are located close enough to the core region such that they may receive sufficient fluence prior to end-of-life (EOL) to require evaluation of embrittlement as part of the RPV analyses associated with pressure-temperature (P-T) limits. In this paper, additional stress and fracture analyses are summarized that were performed for additional PWR nozzles with the following objectives: • To expand the population of PWR nozzle configurations evaluated, which was limited in the previous work to just two nozzles (one inlet and one outlet nozzle). • To model and understand differences in stress results obtained for an internal pressure load case using a two-dimensional (2-D) axi-symmetric finite element model (FEM) vs. a three-dimensional (3-D) FEM for these PWR nozzles. In particular, the ovalization (stress concentration) effect of two intersecting cylinders, which is typical of RPV nozzle configurations, was investigated. • To investigate the applicability of previously recommended linear elastic fracture mechanics (LEFM) hand solutions for calculating the Mode I stress intensity factor for a postulated nozzle corner crack for pressure loading for these PWR nozzles. These analyses were performed to further expand earlier work completed to support potential revision and refinement of Title 10 to the U.S. Code of Federal Regulations (CFR), Part 50, Appendix G, “Fracture Toughness Requirements,” and are intended to supplement similar evaluation of nozzles presented at the 2008, 2009, and 2011 Pressure Vessels and Piping (PVP) Conferences. This work is also relevant to the ongoing efforts of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section XI, Working Group on Operating Plant Criteria (WGOPC) efforts to incorporate nozzle fracture mechanics solutions into a revision to ASME B&PV Code, Section XI, Nonmandatory Appendix G.

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