6. Reactor pressure vessel structural integrity research in the US Nuclear Regulatory Commission HSST and HSSI programs

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.

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Deterministic and Probabilistic Assessments of the Reactor Pressure Vessel Structural Integrity: Benchmark Comparisons

Design and Analysis of Pressure Vessels and Piping: Implementation of ASME B31, Fatigue, ASME Section VIII, and Buckling Analyses ◽

10.1115/pvp2003-2190 ◽

2003 ◽

Cited By ~ 2

Author(s):

Silvia Turato ◽

Vincent Venturini ◽

Eric Meister ◽

B. Richard Bass ◽

Terry L. Dickson ◽

...

Keyword(s):

Fracture Mechanics ◽

Pressure Vessel ◽

Structural Integrity ◽

Safety Concern ◽

Nuclear Regulatory Commission ◽

Reactor Pressure Vessel ◽

Oak Ridge ◽

Integrity Assessment ◽

Probabilistic Fracture Mechanics ◽

Reactor Pressure

The structural integrity assessment of a nuclear Reactor Pressure Vessel (RPV) during accidental conditions, such as loss-of-coolant accident (LOCA), is a major safety concern. Besides Conventional deterministic calculations to justify the RPV integrity, Electricite´ de France (EDF) carries out probabilistic analyses. Since in the USA the probabilistic fracture mechanics analyses are accepted by the Nuclear Regulatory Commission (NRC), a benchmark has been realized between EDF and Oak Ridge Structural Assessments, Inc. (ORSA) to compare the models and the computational methodologies used in respective deterministic and probabilistic fracture mechanics analyses. Six cases involving two distinct transients imposed on RPVs containing specific flaw configurations (two axial subclad, two circumferential surface-breaking, and two axial surface-braking flaw configurations) were defined for a French vessel. In two separate phases, deterministic and probabilistic, fracture mechanics analyses were performed for these six cases.

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Irradiation Effects in Reactor Pressure Vessel Steels: A Nuclear Regulatory Commission Perspective

Radiation Embrittlement of Nuclear Reactor Pressure Vessel Steels: An International Review (Second Volume) ◽

10.1520/stp23025s ◽

2008 ◽

pp. 5-5-8 ◽

Cited By ~ 2

Author(s):

CZ Serpan ◽

PN Randall

Keyword(s):

Pressure Vessel ◽

Nuclear Regulatory Commission ◽

Reactor Pressure Vessel ◽

Irradiation Effects ◽

Pressure Vessel Steels ◽

Regulatory Commission ◽

Reactor Pressure

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Structural Integrity Research for Reactor Pressure Vessel Under In-Vessel Retention of a Core Melt

Volume 1: Operations and Maintenance, Aging Management and Plant Upgrades; Nuclear Fuel, Fuel Cycle, Reactor Physics and Transport Theory; Plant Systems, Structures, Components and Materials; I&C, Digital Controls, and Influence of Human Factors ◽

10.1115/icone24-60092 ◽

2016 ◽

Author(s):

Yongjian Gao ◽

Yinbiao He ◽

Ming Cao ◽

Yuebing Li ◽

Shiyi Bao ◽

...

Keyword(s):

Pressure Vessel ◽

Material Properties ◽

Power Plants ◽

Structural Integrity ◽

Plastic Material ◽

Plastic Region ◽

Reactor Pressure Vessel ◽

Elastic Plastic ◽

Elastic Plastic Material ◽

Reactor Pressure

In-Vessel Retention (IVR) is one of the most important severe accident mitigation strategies of the third generation passive Nuclear Power Plants (NPP). It is intended to demonstrate that in the case of a core melt, the structural integrity of the Reactor Pressure Vessel (RPV) is assured such that there is no leakage of radioactive debris from the RPV. This paper studied the IVR issue using Finite Element Analyses (FEA). Firstly, the tension and creep testing for the SA-508 Gr.3 Cl.1 material in the temperature range of 25°C to 1000°C were performed. Secondly, a FEA model of the RPV lower head was built. Based on the assumption of ideally elastic-plastic material properties derived from the tension testing data, limit analyses were performed under both the thermal and the thermal plus pressure loading conditions where the load bearing capacity was investigated by tracking the propagation of plastic region as a function of pressure increment. Finally, the ideal elastic-plastic material properties incorporating the creep effect are developed from the 100hr isochronous stress-strain curves, limit analyses are carried out as the second step above. The allowable pressures at 0 hr and 100 hr are obtained. This research provides an alternative approach for the structural integrity evaluation for RPV under IVR condition.

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Development of an Integrity Evaluation System on the Basis of Cooperative Virtual Reality Environment for Nuclear Power Plant

Experience With Creep-Strength Enhanced Ferritic Steels and New and Emerging Computational Methods ◽

10.1115/pvp2004-2585 ◽

2004 ◽

Author(s):

J. C. Kim ◽

J. B. Choi ◽

Y. H. Choi

Keyword(s):

Virtual Reality ◽

Power Plant ◽

Nuclear Power Plant ◽

Pressure Vessel ◽

Evaluation System ◽

Nuclear Power ◽

Structural Integrity ◽

Reactor Pressure Vessel ◽

Virtual Reality Environment ◽

Reactor Pressure

Since early 1950’s fracture mechanics has brought significant impact on structural integrity assessment in a wide range of industries such as power, transportation, civil and petrochemical industries, especially in nuclear power plant industries. For the last two decades, significant efforts have been devoted in developing defect assessment procedures, from which various fitness-for-purpose or fitness-for-service codes have been developed. From another aspect, recent advances in IT (Information Technologies) bring rapid changes in various engineering fields. IT enables people to share information through network and thus provides concurrent working environment without limitations of working places. For this reason, a network system based on internet or intranet has been appeared in various fields of business. Evaluating the integrity of structures is one of the most critical issues in nuclear industry. In order to evaluate the integrity of structures, a complicated and collaborative procedure is required including regular in-service inspection, fracture mechanics analysis, etc. And thus, experts in different fields have to cooperate to resolve the integrity problem. In this paper, an integrity evaluation system on the basis of cooperative virtual reality environment for reactor pressure vessel which adapts IT into a structural integrity evaluation procedure for reactor pressure vessel is introduced. The proposed system uses Virtual Reality (VR) technique, Virtual Network Computing (VNC) and knowledge based programs. This system is able to support 3-dimensional virtual reality environment and to provide experts to cooperate by accessing related data through internet. The proposed system is expected to provide a more efficient integrity evaluation for reactor pressure vessel.

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Manufacturing and Properties of Closure Head Forging Integrated With Flange for PWR Reactor Pressure Vessel

Storage Tank Integrity and Materials Evaluation ◽

10.1115/pvp2004-3056 ◽

2004 ◽

Cited By ~ 2

Author(s):

Komei Suzuki ◽

Etsuo Murai ◽

Yasuhiko Tanaka ◽

Iku Kurihara ◽

Tomoharu Sasaki ◽

...

Keyword(s):

Weld Joint ◽

Pressure Vessel ◽

Material Properties ◽

Impact Test ◽

Structural Integrity ◽

Reactor Pressure Vessel ◽

Conventional Design ◽

Manufacturing Technologies ◽

Reactor Pressure

Closure head forging (SA508, Gr.3 Cl.1) integrated with flange for PWR reactor pressure vessel has been developed. This is intended to enhance structural integrity of closure head resulted in elimination of ISI, by eliminating weld joint between closure head and flange in the conventional design. Manufacturing procedures have been established so that homogeneity and isotropy of the material properties can be assured in the closure head forging integrated with flange. Acceptance tensile and impact test specimens are taken and tested regarding the closure head forging integrated with flange as very thick and complex forgings. This paper describes the manufacturing technologies and material properties of the closure head forging integrated with flange.

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Probabilistic Assessments of the Reactor Pressure Vessel Structural Integrity: Direct Coupling Between Probabilistic and Finite-Element Codes to Model Sensitivity to Key Thermo-Hydraulic Variability

12th International Conference on Nuclear Engineering, Volume 2 ◽

10.1115/icone12-49301 ◽

2004 ◽

Cited By ~ 2

Author(s):

Etienne de Rocquigny ◽

Yoan Chevalier ◽

Silvia Turato ◽

Eric Meister

Keyword(s):

Finite Element ◽

Pressure Vessel ◽

Mechanical Model ◽

Analytical Approach ◽

Structural Integrity ◽

Reactor Pressure Vessel ◽

Direct Coupling ◽

Integrity Assessment ◽

Key Variables ◽

Reactor Pressure

The structural integrity assessment of a nuclear Reactor Pressure Vessel (RPV) during accidental conditions such as loss-of-coolant accident (LOCA) is a major safety concern. Besides conventional deterministic calculations to justify as a nuclear operator the RPV integrity, Electricite´ de France (EDF) carries out probabilistic analyses. Probabilistic analyses become most interesting when some key variables, albeit conventionally taken at conservative values, can be modelled more accurately through statistical variability. In the context of low failure probabilities, this requires however a specific coupling effort between a specific probabilistic analysis method (e.g. Form-Sorm method) and the thermo-mechanical model to be reasonable in computing time. In this paper, the variability of a key variable — the mid-transient cooling temperature, tied to a climate-dependent tank — has been modelled, in some flaw configurations (axial sub-clad) for a French vessel. In a first step, a simplified analytical approach was carried out to assess its sensitivity upon the thermo-mechanical phenomena; hence, a direct coupling had to be implemented to allow a probabilistic calculation on the finite-element mechanical model, taking also into account a failure event properly defined through minimisation of the instantaneous failure margin during the transient. Comparison with the previous (indirectly-coupled) studies and the simplified analytical approach is drawn, demonstrating the interest of this new modelling effort to understand and order the sensitivity of the probability of crack initiation to the key variables. While being noticeable in the cases studied, sensitivity to the safety injection temperature variability proves to be less than the choice of the toughness model. Finally, regularity of the thermo-mechanical model is evidenced by the coupling exercise, suggesting that a modified response-surface based method could replace direct coupling for further investigation.

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Structural Integrity of Penetration Nozzle of Reactor Pressure Vessel Head of PWR Plant

Volume 1: Codes and Standards ◽

10.1115/pvp2008-61178 ◽

2008 ◽

Author(s):

Kiminobu Hojo ◽

Naoki Ogawa ◽

Yoichi Iwamoto ◽

Kazutoshi Ohoto ◽

Seiji Asada ◽

...

Keyword(s):

Residual Stress ◽

Stress Intensity Factor ◽

Stress Intensity ◽

Intensity Factor ◽

Stress Field ◽

Pressure Vessel ◽

Structural Integrity ◽

Complex Structure ◽

Reactor Pressure Vessel ◽

Reactor Pressure

A reactor pressure vessel (RPV) head of PWR has penetration holes for the CRDM nozzles, which are connected with the vessel head by J-shaped welds. It is well-known that there is high residual stress field in vicinity of the J-shaped weld and this has potentiality of PWSCC degradation. For assuring stress integrity of welding part of the penetration nozzle of the RPV, it is necessary to evaluate precise residual stress and stress intensity factor based on the stress field. To calculate stress intensity factor K, the most acceptable procedure is numerical analysis, but the penetration nozzle is very complex structure and such a direct procedure takes a lot of time. This paper describes applicability of simplified K calculation method from handbooks by comparing with K values from finite element analysis, especially mentioning crack modeling. According to the verified K values in this paper, fatigue crack extension analysis and brittle fracture evaluation by operation load were performed for initial crack due to PWSCC and finally structural integrity of the penetration nozzle of RPV head was confirmed.

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Assessment of Structural Integrity on Irradiated Steel Structure: Focusing on Long Column Type Reactor Pressure Vessel Supports

Volume 6B: Materials and Fabrication ◽

10.1115/pvp2019-93640 ◽

2019 ◽

Author(s):

Goeun Han ◽

Sukru Guzey

Keyword(s):

Neutron Irradiation ◽

Pressure Vessel ◽

Structural Integrity ◽

Element Model ◽

Reactor Pressure Vessel ◽

Primary Coolant ◽

Type Reactor ◽

Column Type ◽

Unfavorable Condition ◽

Reactor Pressure

Abstract The structural steel in a nuclear facility experiences significant degradation due to the accumulated neutron irradiation. Particularly, the long-column type reactor pressure vessel supports have been focused since they resist considerable loading to maintain the primary coolant system in their position and experience high neutron irradiation in low temperature, which is an unfavorable condition for the fracture toughness. This study implemented the API 579-1/ASME FFS-1, fitness-for-service (FFS) method to consider both irradiated mechanical properties and multiple loading cases. A three-dimensional (3D) finite element model of long column type reactor pressure vessel support was built for the linear analysis. The metallurgical properties of reactor pressure vessel support for assessment were estimated by empirical equations. This study provides the structural margin of long-column type reactor pressure vessel support by levels of the loads and levels of the neutron fluence.

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Fracture Mechanics Assessment of Reactor Pressure Vessel Irradiated Structural Steel for Short Column Type Supports and Neutron Shield Tank

Volume 6: Materials and Fabrication ◽

10.1115/pvp2020-21142 ◽

2020 ◽

Author(s):

Alexandria M. Carolan ◽

J. Brian Hall ◽

Stephen K. Longwell ◽

F. Arzu Alpan ◽

Gregory M. Imbrogno ◽

...

Keyword(s):

Fracture Mechanics ◽

Structural Steel ◽

Pressure Vessel ◽

Radiation Effects ◽

Structural Integrity ◽

Reactor Pressure Vessel ◽

Radiation Embrittlement ◽

Short Column ◽

License Renewal ◽

Reactor Pressure

Abstract As plants apply for 80 year licensure (subsequent license renewal), the United States Nuclear Regulatory Commission (U.S. NRC) has queried the nuclear power plant industry to investigate the impact of neutron embrittlement (radiation effects) on the reactor pressure vessel (RPV) structural steel supports due to extended plant operation past 60 years. The radiation effects on RPV supports were previously investigated and resolved as part of Generic Safety Issue No. 15 (GSI-15) in NUREG-0933 Revision 3 [1], NUREG-1509 [2] (published in May 1996), and NUREG/CR-5320 [3] (published in January 1989) for design life (40 years) and for first license renewal (20 additional years). The conclusions in NUREG-0933, Revision 3 stated that there were no structural integrity concerns for the RPV support structural steels; even if all the supports were totally removed (i.e. broken), the piping has acceptable margin to carry the load of the vessel. Nevertheless, for plants applying for 80 year life licensure, the U.S. NRC has requested an evaluation to show structural integrity of the RPV supports by accounting for radiation embrittlement (radiation damage) for continued operation into the second license renewal period (i.e. 80 years). The RPV support designs in light water reactors are grouped into one of five categories or types of supports: (1) skirt; (2) long-column; (3) shield-tank; (4) short column; and (5) suspension. In this paper, two of these RPV support configurations (short column supports and neutron shield tank) will be investigated using fracture mechanics to evaluate the effect of radiation embrittlement of the structural steel supports for long term operations (i.e. 80 years). The technical evaluation of other support configurations will be provided in a separate technical publication at a future date.

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