Decommissioning and Dismantling of the French Brennilis NPP

2011 ◽  
Author(s):  
Thierry Andre´ ◽  
Werner Botzem
Keyword(s):  
Dose Rate ◽  
Heavy Water ◽  
Pressure Vessel ◽  
Central Element ◽  
Reactor Pressure Vessel ◽  
Core Area ◽  
Pipe System ◽  
The Core ◽  
Reactor Block ◽  
Reactor Pressure

This paper deals with the dismantling of the Brennilis NPP plant located in the west of France (Finiste`re). This prototype NPP of Brennilis was the unique reactor of the heavy water developed in France during the 50’ and the 60’. The reactor diverged in December 1966 and the NPP was operated during 9 years from 1972 to 1981, then the permanent shutdown occurred in July 1985. In 2008, the operator and owner of the plant Electricite´ de France (EDF) commissioned the consortium Onet Technologies Grands Projets (France) and Nukem Technologies (Germany) with the dismantling of the reactor block of the NPP. The reactor block essentially contains the reactor vessel including built-in units and biological shields, the peripheral piping as well as systems for controlling the nuclear-related process. In addition to the complete dismantling, the scope of the contractual services also includes their proper handling in accordance with the applicable regulation: safety requirements, waste management, radioprotection optimization and management. The central element of the plant is the reactor pressure vessel filled with heavy water. Each of the 216 horizontal fuel element channels made of zircaloy is at each side connected to a pipe which directs the heat transfer gas to a header mounted in the upper part of the reactor block. The control rods are introduced vertically into the reactor. It should be pointed out that due to this reactor design, the reactor pressure vessel is equipped with a complex pipe system to all sides which makes it difficult to freely access the core area of the reactor block and thus to dismantle the reactor. In this context, the axial and lateral neutron shields should be mentioned, which are situated in close proximity to the reactor as well as the biological shield which protects from ionizing radiation originating from the pressure vessel. The access to the core area is made difficult due to a high local dose rate and the extremely high constructive complexity of the prototype, the interior of which is virtually criss-crossed by complex piping. The elevated local dose rate in the area of the reactor pressure vessel makes manual work in this zone impossible (even after 30 years), so that remote dismantling techniques have to be used. Before starting dismantling, the remote devices are determined with the help of a test stand which representatively simulates the real conditions of the reactor block in respect to dimensions and material.

Reactor Pressure Vessel Integrity Assessment: French Simplified Analysis Method Applied to Underclad Postulated Defect in Nozzle

2016 ◽  
Author(s):  
Adolfo Arrieta-Ruiz ◽  
Eric Meister ◽  
Stéphane Vidard
Keyword(s):  
Fracture Mechanics ◽  
Fracture Risk ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Core Area ◽  
Integrity Assessment ◽  
Elastic Plastic ◽  
The Core ◽  
Simplified Method ◽  
Reactor Pressure

Structural integrity of the Reactor Pressure Vessel (RPV) is one of the main concerns regarding safety and lifetime of Nuclear Power Plants (NPP) since this component is considered as not reasonably replaceable. Fast fracture risk is the main potential damage considered in the integrity assessment of RPV. In France, deterministic integrity assessment for RPV vis-à-vis the brittle fracture risk is based on the crack initiation stage. As regards the core area in particular, the stability of an under-clad postulated flaw is currently evaluated under a Pressurized Thermal Shock (PTS) through a dedicated fracture mechanics simplified method called “beta method”. However, flaw stability analyses are also carried-out in several other areas of the RPV. Thence-forward performing uniform simplified inservice analyses of flaw stability is a major concern for EDF. In this context, 3D finite element elastic-plastic calculations with flaw modelling in the nozzle have been carried out recently and the corresponding results have been compared to those provided by the beta method, codified in the French RSE-M code for under-clad defects in the core area, in the most severe events. The purpose of this work is to validate the employment of the core area fracture mechanics simplified method as a conservative approach for the under-clad postulated flaw stability assessment in the complex geometry of the nozzle. This paper presents both simplified and 3D modelling flaw stability evaluation methods and the corresponding results obtained by running a PTS event. It shows that the employment of the “beta method” provides conservative results in comparison to those produced by elastic-plastic calculations for the cases here studied.

γ Dose Rate Estimation and Operation Management Suggestions for Decommissioning the Reactor Pressure Vessel of HTR-PM

2013 ◽  
Author(s):  
Sheng Fang ◽  
Hong Li ◽  
Jianzhu Cao ◽  
Wenqian Li ◽  
Feng Xie ◽  
...  
Keyword(s):  
Dose Rate ◽  
Pressure Vessel ◽  
Source Term ◽  
Gamma Dose ◽  
Reactor Pressure Vessel ◽  
Operation Management ◽  
Gamma Dose Rate ◽  
Calculated Dose ◽  
Chinese Standard ◽  
Reactor Pressure

China is now designing and constructing a high temperature gas cooled reactor-pebble bed module (HTR-PM). In order to investigate the future decommissioning approach and evaluate possible radiation dose, gamma dose rate near the reactor pressure vessel was calculated for different cooling durations using QAD-CGA program. The source term of this calculation was provided by KORIGEN program. Based on the calculated results, the spatial distribution and temporal changes of gamma dose rate near reactor pressure vessel was systematically analyzed. A suggestion on planning decommissioning operation of reactor pressure vessel of HTR-PM was given based on calculated dose rate and the Chinese Standard GB18871-2002.

Numerical Simulation of Irradiation-Induced Grain-Boundary Phosphorous Segregation in Reactor Pressure Vessel Steels Using Rate Theory Model With First-Principles Calculations

2009 ◽  
Author(s):  
Ken-ichi Ebihara ◽  
Masatake Yamaguchi ◽  
Yutaka Nishiyama ◽  
Kunio Onizawa ◽  
Hiroshi Matsuzawa
Keyword(s):  
Grain Boundary ◽  
Dose Rate ◽  
Pressure Vessel ◽  
First Principles ◽  
Theory Model ◽  
Irradiation Temperature ◽  
Reactor Pressure Vessel ◽  
Rate Theory ◽  
First Principles Calculations ◽  
Reactor Pressure

The experimental results on neutron-irradiated reactor pressure vessel (RPV) steels have revealed grain boundary segregation of phosphorous (P) due to neutron irradiation, which may lead to intergranular fracture. Because of the lack of experimental database, however, the dependence of the segregation on variables such as dose, dose-rate, and temperature is not clear. Here, we incorporate the parameters determined by first-principles calculations into the rate theory model which was developed for bcc lattice on the basis of the fcc lattice model proposed by Murphy and Perks [1], and apply it to the simulation of irradiation-induced P segregation in bcc iron. We evaluate the grain boundary P coverage and discuss its dependence on dose-rate and irradiation temperature by comparing our results with previously reported results and experimental data. As results, we find that dose-rate does not affect the grain boundary P coverage within the range of our simulation condition and that the dependence on irradiation temperature differs remarkably from the previous results.

Severe accident simulation for VVER-1000 reactor using ASTEC-V2.1.1.3

Kerntechnik ◽  
2021 ◽  
Vol 86 (6) ◽  
pp. 454-469
Author(s):  
S. H. Abdel-Latif
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Fission Products ◽  
Reactor Pressure Vessel ◽  
Severe Accident ◽  
Potential Core ◽  
Safety Requirements ◽  
The Core ◽  
Active Safety Systems ◽  
Reactor Pressure

Abstract The station black-out (SBO) is one of the main accident sequences to be considered in the field of severe accident research. To evaluate a nuclear power plant’s behavior in the context of this accident, the integral ASTEC-V2.1.1.3 code “Accident Source Term Evaluation Code” covers sequences of SBO accidents that may lead to a severe accident. The aim of this work is to discuss the modelling principles for the core melting and in-vessel melt relocation phenomena of the VVER-1000 reactor. The scenario of SBO is simulated by ASTEC code using its basic modules. Then, the simulation is performed again by the same code after adding and activating the modules; ISODOP, DOSE, CORIUM, and RCSMESH to simulate the ex-vessel melt. The results of the two simulations are compared. As a result of SBO, the active safety systems are not available and have not been able to perform their safety functions that maintain the safety requirements to ensure a secure operation of the nuclear power plant. As a result, the safety requirements will be violated causing the core to heat-up. Moreover potential core degradation will occur. The present study focuses on the reactor pressure vessel failure and relocation of corium into the containment. It also discusses the transfer of Fission Products (FPs) from the reactor to the containment, the time for core heat-up, hydrogen production and the amount of corium at the lower plenum reactor pressure vessel is determined.

Buoyancy Driven Coolant Mixing Studies of Natural Circulation Flows at the ROCOM Test Facility Using ANSYS CFX

2006 ◽  
Author(s):  
Thomas Ho¨hne ◽  
So¨ren Kliem ◽  
Ulrich Rohde ◽  
Frank-Peter Weiß
Keyword(s):  
Pressure Vessel ◽  
Natural Circulation ◽  
Reactor Pressure Vessel ◽  
Density Difference ◽  
Test Facility ◽  
Mean Flow ◽  
Pressurized Water ◽  
The Core ◽  
Ansys Cfx ◽  
Reactor Pressure

Coolant mixing in the cold leg, downcomer and the lower plenum of pressurized water reactors is an important phenomenon mitigating the reactivity insertion into the core. Therefore, mixing of the de-borated slugs with the ambient coolant in the reactor pressure vessel was investigated at the four loop 1:5 scaled ROCOM mixing test facility. Thermal hydraulics analyses showed, that weakly borated condensate can accumulate in particular in the pump loop seal of those loops, which do not receive safety injection. After refilling of the primary circuit, natural circulation in the stagnant loops can re-establish simultaneously and the de-borated slugs are shifted towards the reactor pressure vessel (RPV). In the ROCOM experiments, the length of the flow ramp and the initial density difference between the slugs and the ambient coolant was varied. From the test matrix experiments with 0 resp. 2% density difference between the de-borated slugs and the ambient coolant were used to validate the CFD software ANSYS CFX. To model the effects of turbulence on the mean flow a higher order Reynolds stress turbulence model was employed and a mesh consisting of 6.4 million hybrid elements was utilized. Only the experiments and CFD calculations with modeled density differences show a stratification in the downcomer. Depending on the degree of density differences the less dense slugs flow around the core barrel at the top of the downcomer. At the opposite side the lower borated coolant is entrained by the colder safety injection water and transported to the core. The validation proves that ANSYS CFX is able to simulate appropriately the flow field and mixing effects of coolant with different densities.

scholarly journals Influence of Modelling Options in RELAP5/SCDAPSIM and MAAP4 Computer Codes on Core Melt Progression and Reactor Pressure Vessel Integrity

2010 ◽  
Vol 2010 ◽  
pp. 1-11 ◽  
Author(s):  
Siniša Šadek ◽  
Srđan Špalj ◽  
Bruno Glaser
Keyword(s):  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Severe Accident ◽  
Primary System ◽  
Coolant Pump ◽  
The Core ◽  
Hydraulic Behaviour ◽  
Computer Codes ◽  
Vessel Integrity ◽  
Reactor Pressure

RELAP5/SCDAPSIM and MAAP4 are two widely used severe accident computer codes for the integral analysis of the core and the reactor pressure vessel behaviour following the core degradation. The objective of the paper is the comparison of code results obtained by application of different modelling options and the evaluation of influence of thermal hydraulic behaviour of the plant on core damage progression. The analysed transient was postulated station blackout in NPP Krško with a leakage from reactor coolant pump seals. Two groups of calculations were performed where each group had a different break area and, thus, a different leakage rate. Analyses have shown that MAAP4 results were more sensitive to varying thermal hydraulic conditions in the primary system. User-defined parameters had to be carefully selected when the MAAP4 model was developed, in contrast to the RELAP5/SCDAPSIM model where those parameters did not have any significant impact on final results.

Analysis of a Concept for the Diagnosis of the Core State During Severe Accidents in Pressurized Water Reactors

2014 ◽  
Author(s):  
Sebastian Schmidt ◽  
Daniel Fiß ◽  
Alexander Kratzsch
Keyword(s):  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Core State ◽  
Process Technology ◽  
Pressurized Water ◽  
Non Invasive ◽  
The Core ◽  
Gamma Radiations ◽  
State Diagnosis ◽  
Reactor Pressure

In line with the cooperative project “Non-invasive Condition Monitoring of Nuclear Reactors for the Detection of Level Change and Deformation of the Core” between the Technical University Dresden and the Institute of Process Technology, Process Automation and Measuring Technology (IPM) of Zittau/Goerlitz University of Applied Sciences, a measuring system for the core state diagnosis during a core melt accident in the reactor pressure vessel of a pressurized water reactor is going to be developed. The operational principle of this system is based on the non-invasive measurement of continuously changing gamma radiations (caused by the shifting of melted materials and fission products) outside of the reactor pressure vessel by means of several gamma radiation sensors. The sensors are arranged over the height of the core and the lower head. By using computer based and real-time capable methods for evaluation of the measured gamma radiations conclusions about the core state can subsequently be drawn. This paper includes a description of a concept as well as several methods for the core state diagnosis during a core melt accident. The main part is the analysis of the methods for the core state diagnosis based on different evaluation criteria.

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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