Safety Standard KTA 3203: Monitoring Radiation Embrittlement of Reactor Pressure Vessels of Light-Water Reactors

2008 ◽  
pp. 177-177-7 ◽  
Author(s):  
K-R Ernst ◽  
EN Klausnitzer ◽  
C Leitz
Keyword(s):  
Pressure Vessels ◽  
Radiation Embrittlement ◽  
Light Water Reactors ◽  
Safety Standard ◽  
Light Water ◽  
Water Reactors ◽  
Reactor Pressure

Ensuring the Performance of Nuclear Reactor Pressure Vessels for Long-Time Service

2010 ◽  
Author(s):  
Randy K. Nanstad ◽  
G. Robert Odette ◽  
Mikhail A. Sokolov
Keyword(s):  
Pressure Vessels ◽  
High Dose ◽  
Critical Element ◽  
Light Water Reactors ◽  
Time Service ◽  
Light Water ◽  
Time Operation ◽  
Long Time ◽  
Water Reactors ◽  
Reactor Pressure

Structural integrity of the reactor pressure vessel is a critical element in demonstrating the capability of light water reactors for operation to at least 80 y. The Light Water Reactor Sustainability Program Plan is a collaborative program between the U.S. Department of Energy and the private sector directed at extending the life of the present generation of nuclear power plants to enable such long-time operation. Given that the current generation of light water reactors were intended to operate for 40 y, there are significant issues that need to be addressed to reduce the uncertainties in regulatory application. The neutron dose to the vessel will at least double, and the database for such high dose levels under the low flux conditions in the vessel is nonexistent. Associated with this factor are uncertainties regarding flux effects, effects of relatively high nickel content, uncertainties regarding application of fracture mechanics, thermal annealing and reirradiation. The issue of high neutron fluence/long irradiation times and flux effects is the highest priority. Both data and mechanistic understanding are needed to enable accurate, reliable embrittlement predictions at high fluences. This paper discusses the major issues associated with long-time operation of existing RPVs, the LWRSP plans to address those issues, and recent relevant results.

Advanced Structural Materials and Cladding

MRS Bulletin ◽  
2009 ◽  
Vol 34 (1) ◽  
pp. 20-27 ◽  
Author(s):  
T. Allen ◽  
H. Burlet ◽  
R.K. Nanstad ◽  
M. Samaras ◽  
S. Ukai
Keyword(s):  
High Temperature ◽  
Structural Materials ◽  
Ceramic Composites ◽  
Pressure Vessels ◽  
Austenitic Steels ◽  
High Temperature Strength ◽  
Light Water Reactors ◽  
Modeling Tools ◽  
Light Water ◽  
Water Reactors

AbstractAdvanced nuclear energy systems, both fission- and fusion-based, aim to operate at higher temperatures and greater radiation exposure levels than experienced in current light water reactors. Additionally, they are envisioned to operate in coolants such as helium and sodium that allow for higher operating temperatures. Because of these unique environments, different requirements and challenges are presented for both structural materials and fuel cladding. For core and cladding applications in intermediate-temperature reactors (400–650°C), the primary candidates are 9–12Cr ferritic–martensitic steels (where the numbers represent the weight percentage of Cr in the material, i.e., 9–12 wt%) and advanced austenitic steels, adapted to maximize high-temperature strength without compromising lower temperature toughness. For very high temperature reactors (>650°C), strength and oxidation resistance are more critical. In such conditions, high-temperature metals as well as ceramics and ceramic composites are candidates. For all advanced systems operating at high pressures, performance of the pressure boundary materials (i.e., those components responsible for containing the high-pressure liquids or gases that cool the reactor) is critical to reactor safety. For some reactors, pressure vessels are anticipated to be significantly larger and thicker than those used in light water reactors. The properties through the entire thickness of these components, including the effects of radiation damage as a function of damage rate, are important. For all of these advanced systems, optimizing the microstructures of candidate materials will allow for improved radiation and high-temperature performance in nuclear applications, and advanced modeling tools provide a basis for developing optimized microstructures.

Investigations on the Melt Gate Ablation by Ex-Vessel Core Melts in the KAPOOL Experiments

2002 ◽  
Author(s):  
Beatrix Eppinger ◽  
Silke Schmidt-Stiefel ◽  
Walter Tromm
Keyword(s):  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Light Water Reactors ◽  
Light Water ◽  
Theoretical Tool ◽  
Transient Experiments ◽  
Core Meltdown ◽  
Water Reactors ◽  
Radioactive Release ◽  
Reactor Pressure

In future Light Water Reactors (LWR) containment failure should be prevented even for very unlikely core meltdown sequences with reactor pressure vessel (RPV) failure. In the case of such a postulated core meltdown accident in a future LWR the ex-vessel melt shall be retained and cooled in a special compartment inside the containment to exclude significant radioactive release to the environment. In such a case, a gate has to be designed to allow the melt release from the reactor cavity into the compartment. A series of transient experiments has been performed to investigate the melt gate ablation using iron and alumina melts as a simulant for the corium melt. The results of the KAPOOL tests are analyzed with the HEATING5 code in order to evaluate realistic cases of internally heated corium melts and melt gates with the same theoretical tool.

scholarly journals Nuclear Power after Fukushima

2011 ◽  
Vol 133 (12) ◽  
pp. 27-29 ◽  
Author(s):  
Gail H. Marcus
Keyword(s):  
Nuclear Power ◽  
Cost Savings ◽  
Nuclear Material ◽  
Light Water Reactors ◽  
Light Water ◽  
Loss Of Coolant Accident ◽  
Major Loss ◽  
Water Reactors ◽  
A Current ◽  
Reactor Pressure

This article discusses advanced reactor technologies that are now getting renewed attention after the Fukushima nuclear plant accident. Interest in smaller reactors has been growing in recent years. Some of these designs have advantages over the traditional large light water reactors (LWRs) for certain applications. The smaller designs carry less of an inventory of nuclear material, so there is less material at risk in an accident involving a release. Proponents of small modular reactors (SMRs) point to cost savings due to the factory fabrication and shorter construction times. They have significant advantages for countries with small grids, where a current 1500 MWe reactor would exceed demand and threaten grid stability. Other designs that are getting the most attention at present are small or medium LWR concepts. In addition to their smaller size, these designs differ from current large, light-water designs in that most of them use an “integral” design. Most major reactor components are inside the reactor pressure vessel, thus significantly reducing the threat of a major loss-of-coolant accident.

Repair Welding of Irradiated Reactor Pressure Vessels Steel and Current Direction of Revising JSME Fitness-for-Service Code

2017 ◽  
Author(s):  
Hiroyuki Adachi ◽  
Ryuji Kimura ◽  
Yusuke Kono
Keyword(s):  
Power Plants ◽  
Primary Repair ◽  
Pressure Vessels ◽  
Light Water Reactors ◽  
Weld Repair ◽  
Asme Code ◽  
Neutron Transmutation ◽  
Water Reactors ◽  
Reactor Pressure ◽  
Reactor Internals

A primary repair option of Light Water Reactors (LWR) components is welding. However, it is known that welding on steels that have been exposed to neutron irradiation [i.e. Reactor Pressure Vessels (RPV) in PWR and Reactor Internals in BWR] can result in Helium Induced Cracking (HeIC). Helium forms from neutron transmutation reactions of Boron (B) and Nickel (Ni) during operation of the plant. In order to address this issue and establish verified methods for weld repair of irradiated RPV and Reactor Internals materials in Japanese power plants, an investigation denominated WIM (Welding of Irradiated Materials) Project was conducted; the WIM project was carried out between the years of 1997 and 2004 in an analytical conservative manner, correlating the results of weld repaired irradiated materials with the concentration of helium and the heat input used while welding. It was concluded that, under determined conditions, the irradiated materials were able to be successfully welded in accordance with the requirements established in both the JSME and ASME Code Cases. In the light of such discovery, the necessity of establishing and a new code case and revising the standard JSME Rules on Fitness-for-Service concerning the weld repair of irradiated RPV and Reactor Internals steels is currently under investigation.

Sign in / Sign up

Export Citation Format

Share Document