Impact of Using ASME Section XI Code Case N-830 on Plant Heatup and Cooldown Pressure-Temperature Limit Curves for Pressurized Water Reactors

2019 ◽  
Author(s):  
Alexandria Carolan ◽  
Benjamin Mays ◽  
Anees Udyawar ◽  
J. Brian Hall
Keyword(s):  
Lower Bound ◽  
Pressure Vessel ◽  
Master Curve ◽  
Temperature Limit ◽  
Reactor Pressure Vessel ◽  
Pressurized Water Reactors ◽  
Pressurized Water ◽  
Water Reactors ◽  
The Impact ◽  
Reactor Pressure

Abstract Nuclear plant reactor pressure vessel heat-up and cool-down pressure temperature (P-T) limit curves are determined using ASME Section XI, Appendix G. ASME has adopted into ASME Section XI Appendix G the allowable use of the ASTM E1921 master curve fracture toughness based reference temperature (T0) to index the KIc curve. ASME Section XI Code Case N-830 allows the use of the KJc 95% lower bound master curve indexed using T0 directly. In ASME Section XI Appendix G, the equation RTT0 = T0 + 19.4°C, as an alternate RTNDT, shifts the KIc curve to approximate the KJc 95% lower bound master curve, however, the KIc exponential curve parameters are different. Thus, this paper evaluates the impact on plant heatup and cooldown pressure-temperature limit curves between the two ASME approved methods for typical pressurized water reactors (PWR). Different degrees of embrittlement are assessed to determine differences in the two approaches on reactor pressure vessel (RPV) beltline operating curves. Furthermore, in the proposed Revision 19 of Regulatory Guide 1.147, the US NRC has included a condition on the use of Code Case N-830 that prohibits the use of the current KIc equation in ASME Section XI Appendix G when these values are above the KJc lower bound 95% curve at temperatures below T0 − 64°C. This paper briefly discusses this NRC condition on the P-T limit curves.

scholarly journals Analysis of the Technological Evolution of Materials Requirements Included in Reactor Pressure Vessel Manufacturing Codes

2021 ◽  
Vol 13 (10) ◽  
pp. 5498
Author(s):  
Alvaro Rodríguez-Prieto ◽  
Mariaenrica Frigione ◽  
John Kickhofel ◽  
Ana M. Camacho
Keyword(s):  
Sustainable Development ◽  
Pressure Vessel ◽  
Nuclear Regulatory Commission ◽  
Green Energy ◽  
Reactor Pressure Vessel ◽  
Technical Literature ◽  
Pressurized Water ◽  
Water Reactors ◽  
Regulatory Commission ◽  
Reactor Pressure

The growth of green energy technologies within the frame of the 7th Sustainable Development Goal (SDG) along with the concern about climatic changes make nuclear energy an attractive choice for many countries to ensure energy security and sustainable development as well as to actively address environmental issues. Unlike nuclear equipment (immovable goods), which are often well-catalogued and analyzed, the design and manufacturing codes and their standardized materials specifications can be considered movable and intangible goods that have not been thoroughly studied based on a detailed evaluation of the scientific and technical literature on the reactor pressure vessel (RPV) materials behavior. The aim of this work is the analysis of historical advances in materials properties research and associated standardized design codes requirements. The analysis, based on the consolidated U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.99 Rev.2 model, enables determination of the best materials options, corresponding to some of the most widely used material specifications such as WWER 15Kh2MFAA (used from the 1970s and 1980s; already in operation), ASME SA-533 Grade B Cl.1 (used in pressurized water reactor-PWR 2nd–4th; already in operation), DIN 20MnMoNi55 and DIN 22NiMoCr37 (used in PWR 2nd–4th) as well as ASTM A-336 Grade F22V (current designs). Consequently, in view of the results obtained, it can be concluded that the best options correspond to recently developed or well-established specifications used in the design of pressurized water reactors. These assessments endorse the fact that nuclear technology is continually improving, with safety being its fundamental pillar. In the future, further research related to the technical heritage from the evolution of materials requirements for other clean and sustainable power generation technologies will be performed.

The Development of Radiation Embrittlement Models for U.S. Power Reactor Pressure Vessel Steels

2006 ◽  
Author(s):  
J. A. Wang ◽  
N. S. V. Rao ◽  
S. Konduri
Keyword(s):  
Pressure Vessel ◽  
Prediction Models ◽  
Irradiation Temperature ◽  
Reactor Pressure Vessel ◽  
Boiling Water ◽  
Radiation Embrittlement ◽  
Pressurized Water ◽  
Water Reactor ◽  
Water Reactors ◽  
Reactor Pressure

The information fusion technique is used to develop radiation embrittlement prediction models for reactor pressure vessel (RPV) steels from U.S. power reactors, including boiling water reactors and pressurized water reactors. The Charpy transition temperature-shift data is used as the primary index of RPV radiation embrittlement in this study. Six parameters—Cu, Ni, P, neutron fluence, irradiation time, and irradiation temperature—are used in the embrittlement prediction models. The results indicate that this new embrittlement predictor achieved reductions of about 49.5% and 52% in the uncertainties for plate and weld data, respectively, for pressurized water reactor and boiling water reactor data, compared with the Nuclear Regulatory Commission Regulatory Guide 1.99, Rev. 2. The implications of dose-rate effect and irradiation temperature effects for the development of radiation embrittlement models are also discussed.

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.

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.

Measurement Challenges Associated With Irradiated Reactor Components

1997 ◽  
Vol 503 ◽  
Author(s):  
Steven R. Doctor
Keyword(s):  
Pressure Vessel ◽  
Nuclear Reactor ◽  
Reactor Pressure Vessel ◽  
Boiling Water Reactors ◽  
Background Information ◽  
Nondestructive Measurement ◽  
Pressurized Water ◽  
Radiation Fields ◽  
Water Reactors ◽  
Reactor Pressure

ABSTRACTNuclear reactor components are known to degrade as reactors age. The reactor pressure vessel and the reactor internals are subjected to the highest radiation fields and thus, age faster. Some boiling water reactors are experiencing extensive cracking of the core shroud and some pressurized water reactors are experiencing a reduction in their fracture toughness. This paper provides background information on reactor pressure vessel materials, fabrication, and construction practices. It identifies some of the challenges that are associated with the nondestructive measurement of changes in material properties associated with these components. The inspection environment, surface conditions and designs are also discussed.

Investigation of the Beltline Welding Seam and Base Metal of the Greifswald WWER-440 Unit 1 Reactor Pressure Vessel

2009 ◽  
Author(s):  
Jan Schuhknecht ◽  
Hans-Werner Viehrig ◽  
Udo Rindelhardt
Keyword(s):  
Weld Metal ◽  
Base Metal ◽  
Pressure Vessel ◽  
Master Curve ◽  
Reactor Pressure Vessel ◽  
Reference Temperature ◽  
Metal Ring ◽  
Welding Seam ◽  
Root Region ◽  
Reactor Pressure

The investigation of reactor pressure vessel (RPV) materials from decommissioned NPPs offers the unique opportunity to scrutinize the irradiation behaviour under real conditions. Material samples taken from the RPV wall enable a comprehensive material characterisation. The paper describes the investigation of trepans taken from the decommissioned WWER-440 first generation RPVs of the Greifswald NPP. Those RPVs represent different material conditions such as irradiated (I), irradiated and recovery annealed (IA) and irradiated, recovery annealed and re-irradiated (IAI). The working program is focussed on the characterisation of the RPV steels (base and weld metal) through the RPV wall. The key part of the testing is aimed at the determination of the reference temperature T0 following the ASTM Test Standard E1921-05 to determine the fracture toughness of the RPV steel in different thickness locations. In a first step the trepans taken from the RPV Greifswald Unit 1 containing the X-butt multilayer submerged welding seam and from base metal ring 0.3.1 both located in the beltline region were investigated. Unit 1 represents the IAI condition. It is shown that the Master Curve approach as adopted in ASTM E1921 is applicable to the investigated original WWER-440 weld metal. The evaluated T0 varies through the thickness of the welding seam. The lowest T0 value was measured in the root region of the welding seam representing a uniform fine grain ferritic structure. Beyond the welding root T0 shows a wavelike behaviour. The highest T0 of the weld seam was not measured at the inner wall surface. This is important for the assessment of ductile-to-brittle temperatures measured on sub size Charpy specimens made of weld metal compact samples removed from the inner RPV wall. Our findings imply that these samples do not represent the most conservative condition. Nevertheless, the Charpy transition temperature TT41J estimated with results of sub size specimens after the recovery annealing was confirmed by the testing of standard Charpy V-notch specimens. The evaluated Charpy-V TT41J shows a better accordance with the irradiation fluence along the wall thickness than the Master Curve reference temperature T0. The evaluated T0 from the trepan of base metal ring 0.3.1 varies through the thickness of the RPV wall. T0 increases from −120°C at the inner surface to −104°C at a distance of 33 mm from it and again to −115°C at the outer RPV wall. The KJc values generally follow the course of the MC, although the scatter is large. The re-embrittlement during 2 campaigns operation can be assumed to be low for the weld and base metal.

Inadvertent Blowdown Event in a Nuclear Boiling Water Reactor

2011 ◽  
Author(s):  
Robert Engel
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Service Level ◽  
Nuclear Safety ◽  
Reactor Pressure Vessel ◽  
Point Of View ◽  
Primary Circuit ◽  
Event Scale ◽  
The Impact ◽  
Reactor Pressure

On March 6th 2007, the Leibstadt Nuclear Power Plant in Switzerland experienced an automatic blowdown of eight safety/relief valves installed on the main steam lines caused by a faulty electrical manipulation while performing planned maintenance during full power operation. Due to the temperature measurements inside the reactor recirculation system and the reactor pressure vessel this event, at a first glance, appeared to be Event No. 23 (Automatic Blowdown event) as an Emergency (Service Level C) Condition in accordance with the relevant reactor pressure vessel Thermal Cycle Diagram. According to the ASME Code Section III, Service Level C limits permit large deformations in areas of structural discontinuity which may necessitate the removal of a component from service for inspection or repair. This paper presents a summary of thermal-hydraulic, stress, fatigue, and fracture mechanical evaluations as well as plant inspections performed to demonstrate the impact of the event on the reactor pressure vessel and associated components and to fulfill the requirements of the Swiss Federal Nuclear Safety Inspectorate. It is shown that the primary circuit of the plant was not inadmissibly stressed by the event and that it was acceptable from a safety-related point of view to return the plant to service. Corresponding to the 7-level International Nuclear and Radiological Event Scale this event was rated afterwards as level 1 (anomaly) by the Swiss Federal Nuclear Safety Inspectorate.

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