Lifetime Management of WWER Reactor Pressure Vessels

2004 ◽  
Author(s):  
Milan Brumovsky
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Pressure Vessels ◽  
Point Of View ◽  
Mitigation Measures ◽  
Economic Point ◽  
Neutron Fluences ◽  
Vessel Integrity ◽  
Surveillance Specimens ◽  
Reactor Pressure

Reactor pressure vessels are the most important components of nuclear power plants from safety and economic point of view. Thus, special interest is given to their lifetime management that requires a precious and periodical evaluation of their conditions. In Czech Republic, special program has been prepared for reactor pressure vessels in both nuclear power stations — Dukovany NPP with 4 units of WWER-440 MW and Temelin with 2 units of WWER-1000 MW reactors. This program is based on the following activities: • detailed calculations of real and potential regimes, especially of pressurized thermal shock events, • determination of vessel properties on the basis of testing surveillance specimens from standard, modified and supplementary programs, • calculations of neutron fluences on vessel wall, • measurements of neutron fluences on surveillance specimens and in vessel cavity, • evaluation of all results, evaluation of fluence and material property trends, • recommendations for future reactor pressure vessel operation (potential mitigation measures). The paper describes in detail all these activities and gives examples of their results and final evaluations. Lifetime assessment is based on a “Unified Procedure for Lifetime Assessment of Components and Piping in WWER NPPs (VERLIFE)” using “Master Curve” approach for vessel integrity evaluation [1].

Ways for Extending the Reactor Pressure Vessel Lifetime

2008 ◽  
Author(s):  
Milan Brumovsky
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Integrity Assessment ◽  
Low Leakage ◽  
Vessel Integrity ◽  
Direct Use ◽  
Reactor Pressure ◽  
Lifetime Extension

Reactor pressure vessels are components that usually determine the lifetime of the whole nuclear power plant and thus also its efficiency and economy. There are several ways how to ensure conditions for reactor pressure vessel lifetime extension, mainly: - pre-operational, like: • optimal design of the vessel; • proper choice of vessel materials and manufacturing technology; - operational, like: • application of low-leakage core; • increase of water temperature in ECCS; • insertion of dummy elements; • vessel annealing; • decrease of conservatism during reactor pressure vessel integrity assessment e.g. using direct use of fracture mechanics parameters, like “Master Curve” approach. All these ways are discussed in the paper and some qualitative as well as quantitative evaluation is given.

Nuclear Reactor Pressure Vessel Integrity Assessment: Enhancement Provided by 3D Modeling Flaw Stability Calculations

2013 ◽  
Author(s):  
Adolfo Arrieta-Ruiz ◽  
Eric Meister ◽  
Henriette Churier
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Integrity Assessment ◽  
Potential Damage ◽  
Vessel Integrity ◽  
Reactor Pressure

Structural integrity of the Reactor Pressure Vessel (RPV) is one of the main considerations regarding safety and lifetime of Nuclear Power Plants (NPP) since this component is considered as not reasonably replaceable. Brittle fracture risk associated with the embrittlement of RPV steel in irradiated areas is the main potential damage. In France, deterministic integrity assessment for RPV is based on the crack initiation stage. The stability of an under-clad postulated flaw in the core area is currently evaluated under a Pressurized Thermal Shock (PTS) through a fracture mechanics simplified method. One of the axes of EDF’s implemented strategy for NPP lifetime extension is the improvement of the deterministic approach with regards to the input data and methods so as to reduce conservatisms. In this context, 3D finite element elastic-plastic calculations with flaw modelling have been carried out recently in order to quantify the enhancement provided by a more realistic approach in the most severe events. The aim of this paper is to present both simplified and 3D modelling flaw stability evaluation methods and the results obtained by running a small break LOCA event.

Structural Integrity Assessment of Doel 3 and Tihange 2 RPVs Affected by Hydrogen Flakes: Refined X-FEM Analyses

2016 ◽  
Author(s):  
Pierre Dulieu ◽  
Valéry Lacroix
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Extended Finite Element Method ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Extended Finite Element ◽  
Integrity Assessment ◽  
Safety Case ◽  
Component Manufacturing ◽  
Reactor Pressure

During the 2012 outage at Doel 3 and Tihange 2 Nuclear Power Plants, specific ultrasonic in-service inspections revealed a large number of quasi-laminar indications in the base metal of the reactor pressure vessels, mainly in the lower and upper core shells. The observed indications could subsequently be attributed to hydrogen flaking induced during the component manufacturing process. As a consequence, a Flaw Acceptability Assessment had to be performed as a part of the Safety Case demonstrating the fitness-for-service of these units. In that framework, detailed analyses using eXtended Finite Element Method were conducted to model the specific character of hydrogen flakes. Their quasi-laminar orientation as well as their high density required setting up 3D multi-flaws model accounting for flaw interaction. These calculations highlighted that even the most penalizing flaw configurations are harmless in terms of structural integrity despite the consideration of higher degradation of irradiated material toughness.

Prediction of Radiation Embrittlement for Highly Irradiated Reactor Pressure Vessel Steels

2009 ◽  
Author(s):  
Hiroshi Matsuzawa
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Pressure Vessels ◽  
Test Results ◽  
Prediction Formula ◽  
Irradiation Embrittlement ◽  
Nuclear Plants ◽  
Domestic Plant ◽  
Test Reactor ◽  
Reactor Pressure

There are 53 (fifty-three) nuclear power plants (both PWR and BWR type) are now under operating in Japan, and the oldest plant has been operating more than thirty years. These plants will be operated until sixty years for operation periods, and will be verified the integrity for assessment of nuclear plants for every ten years in Japan. Reactor Pressure Vessels (RPVs) are required to evaluate the reduction of fracture toughness and the increase of the reference temperature in the transition region. As the operating period will be longer, the prediction for these material properties will be more important. Recently the domestic prediction formula of embrittlement was revised based on the database of domestic plant surveillance test results for thirty years olds as the JEAC4201-2007 [7]. The adequacy for this prediction formula using for sixty year periods is verified by use of the results of the material test reactors (MTRs), but the effects of the accelerated irradiation on embrittlement has not been clear now. So, JNES started the national project, called as “PRE” project on 2005 in order to investigate how flux influences on the ΔRTNDT. In this project the RPV materials irradiated in the actual PWR plant have been re-irradiated in the OECD/Halden test reactor by several different fluxes up to the high fluence region, and the microstructual change for these materials will be investigated in order to make clear the cause of the irradiation embrittlement. In this paper the overall scheme of this project and the summary of the updated results will be presented.

Root Cause Analysis of the Unexpected Behaviour of a Flaked Material Under Irradiation and Transferability to the Doel 3/Tihange 2 Reactor Pressure Vessels

2016 ◽  
Author(s):  
Evy De Bruycker ◽  
Séverine De Vroey ◽  
Xavier Hallet ◽  
Jacqueline Stubbe ◽  
Steve Nardone
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Root Cause Analysis ◽  
Pressure Vessels ◽  
Operating Cycle ◽  
Cause Analysis ◽  
Root Cause ◽  
History Of ◽  
Manufacturing History ◽  
Reactor Pressure

During the 2012 outage at Doel 3 (D3) and Tihange 2 (T2) Nuclear Power Plants (NPP), a large number of nearly-laminar indications were detected mainly in the lower and upper core shells. The D3/T2 shells are made from solid casts that were pierced and forged. Restart authorization in 2013 was accompanied by a number of “mid-term” requirements, to be completed during the first operating cycle after the restart. One of these requirements was the mechanical testing of irradiated specimens containing hydrogen flakes. These tests showed unexpected results regarding the shift in the Reference Temperature for Nil Ductility Transition (RTNDT) of the flaked material VB395 (Steam Generator shell rejected because of flakes) after irradiation. This paper presents the root cause analysis of this unexpected behaviour and its transferability (or not) to the D3/T2 Reactor Pressure Vessels (RPVs). A mechanistic and a manufacturing based approach were used, aiming at identifying the microstructural mechanisms responsible for the atypical embrittlement of VB395 and evaluating the plausibility of these mechanisms in the D3/T2 RPVs. This work was based on expert’s opinions, literature data and test results. Both flaked and unflaked samples have been investigated in irradiated and non-irradiated condition. All hydrogen-related mechanisms were excluded as root cause of the unexpected behaviour of VB395. Two possible mechanisms at the basis of the atypical embrittlement of VB395 were identified, but are still open to discussion. These mechanisms could be linked to the specific manufacturing history of the rejected VB395 shell. Since the larger than predicted shift in transition temperature after irradiation of VB395 is not linked with the hydrogen flaking and since none of the specific manufacturing history features that are possible root causes are reported for the D3/T2 RPVs, the D3/T2 shells should not show the unexpected behaviour observed in VB395.

Thermal Annealing of Reactor Pressure Vessels

2015 ◽  
Author(s):  
Mikhail A. Sokolov ◽  
William L. Server ◽  
Randy K. Nanstad
Keyword(s):  
Thermal Annealing ◽  
Nuclear Power ◽  
Power Plants ◽  
Temperature Shift ◽  
Life Time ◽  
Pressure Vessels ◽  
Operating Life ◽  
Time Extension ◽  
High Doses ◽  
Reactor Pressure

Some of the current fleet of nuclear power plants is poised to reach their end of life and will require an operating life time extension. Therefore, the main structural components, including the reactor pressure vessel (RPV), will be subject to higher neutron exposures than originally planned. These longer operating times raise serious concerns regarding our ability to manage the reliability of RPV steels at such high doses. Thermal annealing is the only option that can, to some degree, recover irradiated beltline region transition temperature shift and recover upper shelf energy properties lost during radiation exposure and extend RPV service life. This paper reviews the experience accumulated internationally with development and implementation of thermal annealing to RPV and potential perspectives for carrying out thermal annealing on US nuclear power plant RPVs.

Current Issues in Operated WWER Type RPVs

2008 ◽  
Author(s):  
Milan Brumovsky´ ◽  
Milos Kytka
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessels ◽  
Point Of View ◽  
Radiation Embrittlement ◽  
Irradiation Effects ◽  
Static Fracture ◽  
High Level ◽  
Near Future ◽  
Surveillance Specimens ◽  
Reactor Pressure

RPVs of WWER type reactors are manufactured from other type of steels (15Kh2MFA of Cr-Mo-V type for WWER-440 and 15Kh2NMFA of Ni-Cr-Mo-V type for WWER-1000) and according to other Codes and standards than PWR ones, thus some specific problems are currently more important for WWER. The principal problem lies in relatively small number of manufactured and operated WWER type NPPs. Even though a high level of unification in RPVs exists — practically only two designs of RPVs exists (WWER-440 and WWER-1000) — total number is still small. All WWER-440 RPV are practically identical, either they were manufactured for V-230 or V-213 model: the only difference is in the purity of used materials and existence/non-existence of the surveillance programmes. (Fact that some V-230 type vessels were not covered by austenitic cladding is not important from irradiation effects point of view.) Regarding WWER-440/V-230 types, it is necessary to take into account, that even though most of them were successfully annealed, only some of them are still in operation but most of them will be closed in near future. Similar situation is with WWER-1000 RPVs, either they were manufactured for V-320 (most frequent), or V-338 or the newest V-428 — differences are practically only in the content of nickel in critical weldments and/or in design of surveillance specimens capsules. But, Large advantage of all WWER surveillance programmes is in loading static fracture toughness specimens in all programmes. The papers tries to summarize and analyze all current issues connected with radiation embrittlement of operated reactor pressure vessels of WWER type.

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