Determination of difficult to measure radionuclides in primary circuit facilities of NPP V1 Jaslovske Bohunice

2013 ◽  
Vol 298 (3) ◽  
pp. 1879-1884 ◽  
Author(s):  
Boris Remenec ◽  
Silvia Dulanska ◽  
L’ubomír Mátel
Keyword(s):  
Primary Circuit ◽  
Difficult To Measure Radionuclides

Consideration of Environmentally Assisted Fatigue in Austenitic Stainless Steel: Calculation and Practical Application

2012 ◽  
Author(s):  
Sven H. Reese ◽  
Johannes Seichter ◽  
Dietmar Klucke
Keyword(s):  
Primary Circuit ◽  
Stress And Strain ◽  
Loading Conditions ◽  
Practical Application ◽  
Circuit Component ◽  
Detailed Assessment ◽  
The Mean ◽  
Fatigue Evaluation ◽  
Assessment Procedures

The influence of LWR coolant environment to the lifetime of materials has been discussed recent years. Nowadays the consideration of environmentally assisted fatigue is under consideration in Codes and Standards like ASME and the German KTA Rules (e.g. Standard No. 3201.2 and Standard No. 3201.4) by means of so called attention thresholds. Basic calculation procedures in terms of quantifying the influence of LWR coolant environment by the Fen correction factor were proposed by Higuchi and others and are given in NUREG/CR-6909. This paper deals with the application of the proposed assessment procedures of ANL and the application to plant conditions. Therefore conservative assessment procedures are introduced without assuming the knowledge of detailed stress and strain calculations or temperature transients. Additionally, detailed assessment procedures based on Finite-Element calculations, respecting in-service temperature measurements including thermal reference transients and complex operational loading conditions are carried out. Fatigue evaluation of a PWR primary circuit component is used in order to evaluate the influence of plant like conditions numerically. Conclusions regarding the practical application are drawn by means of comparing the ANL approach considering laboratory conditions, conservative assessment procedures for the determination of cumulative fatigue usage factors of plant components and detailed assessment procedures. Plant like loading conditions, complex component geometries, loading scenarios and reference temperature transients shall be taken into account. Practical issues like the determination of the mean temperature or the strain rate have to be considered adequately.

Implementation of an Approach for Characterisation of Radiological Inventory for Decommissioning at Chinon A2 NPP

2014 ◽  
Author(s):  
A. Traichel ◽  
F. Tardy ◽  
M. Mummert
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Statistical Processing ◽  
Analysis Data ◽  
Measured Data ◽  
Primary Circuit ◽  
Measurement Systems ◽  
Definition Of ◽  
Local Sampling

A general overview of the existing radioactive inventory in the plant is necessary for the decommissioning of nuclear power plants. Based on the knowledge about radiological inventory, appropriate decommissioning techniques and procedures can be specifically used. In order to derive the existing radiological activity in the facility a study was carried out to obtain a representative overview of the total radiological situation at the NPP. Within a study a generic methodology for the radiological characterization was developed. This methodology has been applied on the CO2-circuit of the gas-cooled, graphite-moderated reactor Chinon A2 (MAGNOX type). This paper covers the implementation of an approach for characterisation of radiological inventory for decommissioning. The approach aims at the definition of the number and distribution of local sampling, required measurements as well as suitable measurement systems leading to a confident result with minimized effort in sampling. The paper covers two main objectives: 1. Methodology at and 2. Determination of radiological inventory based on measured data. The proposed methodology is a stepwise procedure which offers the possibility for minimizing the number of required measurements/sample analyses. At the first step the underlying system is an “as-simple-as-possible”-example with homogeneous contamination. In a second step the methodology is expanded to a more realistic and complex system, for which additional investigations have to be performed. The determination of the radiological inventory using the methodology has to consider a given confidence level and maximum allowed error. Therefore statistical assessment is widely used in estimations. The result of this first part of study generates the basis for further investigation. This comprises application of methodology to the mentioned technical system. Therefore corresponding measurement and analysis data have to be delivered and proven regarding adequacy for the proposed methodology. From the dataset various measurement systematic and retained parameters could be derived. The accuracy of given measured data was checked by further examination. The result of the performed analysis leads to a statement about the activity in the primary circuit. The result of this study is an comprehensive estimation of the activity by defined statistical processing of analysed data. The result consists moreover of the analysis of the measurement plan and of distribution and deviation within the measured data. Suggestions for further measurement campaigns are provided based on the deviations and inconsistencies of the data. With the help of these suggestions it should be possible to decrease the number of samples and measuring data as well as improve the comparability of separate measurement processes. Particular potential for improvement of the result for inventory can be seen in a deeper analysis of uncertainties, this was realised and will be explained in the paper.

Numerical Investigation of Thermal Stratification in the Primary Circuit of VVER-440 Type Reactors

2006 ◽  
Author(s):  
Ildiko´ Boros ◽  
Attila Aszo´di ◽  
Ga´bor Le´gra´di
Keyword(s):  
Thermal Stratification ◽  
Cooling System ◽  
Maximum Temperature ◽  
Primary Circuit ◽  
Operational Experience ◽  
Extensive Evaluation ◽  
Transient Simulations ◽  
Surge Line ◽  
Core Cooling

Thermal stratification in the primary loops and in the connected pipes can limit the lifetime of the piping, or lead to penetrating cracks due to the stresses caused by the temperature differences and the cyclic temperature changes. Therefore it is essential to determine the thermal hydraulic parameters of the stratified flow. The determination of the affected pipes can be based on the international operational experience and on engineering consideration. The most affected pipes in PWRs are the pressurizer surge line, the injection pipe of the emergency core cooling systems and the feedwater injection pipe of the steam generators. CFD codes can provide an appropriate tool for the examination of the development and the breaking up of the stratification and the determination of the temperature distribution. However, the challenge of the uncertainty of the boundary conditions has to be faced because of the unknown flow circumstances. According to an extensive evaluation, performed in 1998 by the VEIKI, in the VVER-440/213 units of Paks NPP the most affected pipe is the pressurizer surge line [1]. To find out the possible thermal stratification in the surge line, a temperature monitoring system was installed on the YP20 leg of the surge line of the Unit 1 of the Paks NPP in 2000. The measurements showed that during the heat-up period there is a thermal stratification almost all time in the surge line [2]. The maximum temperature differences reach 140 K (140 °C). The surge line has been modeled with the CFD code CFX-5.7. The performed transient simulations confirmed the existence of a thermal stratification in the surge line, but showed permanent recirculation of colder coolant in the lower layer, caused by the asymmetric arrangement of the surge line legs and the asymmetric connection of the two legs to the main loop. In this paper, the surge line model and the results of the transient simulations are presented. The CFD model of the injection pipe of the high pressure Emergency Core Cooling System and the performed simulations for the analysis of occurrence of thermal stratification are presented as well.

Residual Stresses in Austenitic Cladding

2007 ◽  
Author(s):  
Milan Brumovsky ◽  
Vladislav Pistora ◽  
Ivan Kupka
Keyword(s):  
Residual Stresses ◽  
Water Environment ◽  
Strength Properties ◽  
Pressure Vessels ◽  
Primary Circuit ◽  
Adjacent Area ◽  
Pressurized Thermal Shock ◽  
Welding Direction ◽  
Reactor Pressure

Reactor pressure vessels (RPV) are usually manufactured with austenitic cladding on their inner surface as a protection against corrosion from the primary circuit water environment. Thus, they are not included into the strength calculations of pressure vessels due to their lower strength properties and much smaller thickness in comparison with those of vessels as they are taken only as a corrosion layer. In the same time, due to different thermal coefficients and Young moduli, welding of austenitic cladding results in a high residual stresses in the cladding and also in the adjacent area in the base ferritic metal. These residual stresses as well as stresses resulted from the temperature field in the vessels represent necessary inputs into pressurized thermal shock calculations. WWER (Water-Water Energy Reactor = PWR type) reactor pressure vessels have relatively thick cladding — nominally 8 mm — made from two layers: first layer of 25/10 type welded by one pass while the second layer of 18/10/Ti typed is usually welded by three passes. The main part of the vessels was performed by strip welding with strips of 60 mm wide. Results of residual stresses measurements are given in the paper. Method with incremental milling of beams was used for the measurements and determination of residual stresses. Tests were performed on specimens in as-welded state and also after final heat treatment of the vessels, i.e. after several stress relieves including first hydrotest in shop. As residual stresses depends strongly also on direction of welding, beams were oriented in both directions — parallel and perpendicular to the welding direction. Results of these measurements are shown and discussed in the paper.

scholarly journals Intercomparison exercise on difficult to measure radionuclides in activated concrete—statistical analysis and comparison with activation calculations

2021 ◽  
Author(s):  
Anumaija Leskinen ◽  
Celine Gautier ◽  
Antti Räty ◽  
Tommi Kekki ◽  
Elodie Laporte ◽  
...  
Keyword(s):  
Statistical Analysis ◽  
Calculation Result ◽  
Nuclear Safety ◽  
Performance Assessments ◽  
Radiochemical Analysis ◽  
Intercomparison Exercise ◽  
Safety Research ◽  
Z Scores ◽  
Difficult To Measure Radionuclides

AbstractThis paper reports the results obtained in a Nordic Nuclear Safety Research project during the second intercomparison exercise for the determination of difficult to measure radionuclides in decommissioning waste. Eight laboratories participated by carrying out radiochemical analysis of 3H, 14C, 36Cl, 41Ca, 55Fe and 63Ni in an activated concrete. In addition, gamma emitters, namely 152Eu and 60Co, were analysed. The assigned values were derived from the submitted results according to ISO 13,528 standard and the performance assessments were determined using z scores. The measured results were compared with activation calculation result showing varying degree of comparability.

scholarly journals New methodological approaches in solving certification problem of historical solid radioactive waste sent for burial from Chornobyl Nuclear Power Plant

2020 ◽  
Vol 19 (4) ◽  
pp. 39-49
Author(s):  
O. V. Mykhailov ◽  
◽  
V. M. Bezmylov ◽  
Keyword(s):  
Power Plant ◽  
Radioactive Waste ◽  
Nuclear Power ◽  
Treatment Plant ◽  
Management Systems ◽  
Standard Methods ◽  
Solid Radioactive Waste ◽  
Methodological Approaches ◽  
Difficult To Measure Radionuclides

Two methodological approaches for radioactive waste (RAW) certification used in RAW management systems in Italy and France, are addressed. Their applicability was assessed in solving certification problem of historical waste accumulated at the Chornobyl NPP in comparison with the standard methods recommended by the IAEA. Testing new methodological approaches was carried out on the example of solid RAW (SRW) of operational origin, which were previously studied for the content of 24 radionuclides within their composition. The procedures for testing researchable methods have used the criteria for SRW acceptance for burial valid in SRW Treatment Plant, which met their current provisions and those ones planned for approval. It was established that the use of quantitative criteria applied in the algorithms of studied methodological approaches for radwaste certification can significantly reduce overestimation degree of summary activity of waste packages by way of removing from the list of difficult-to-measure radionuclides, whose presence can be neglected in view of negligible risk of exceeding the activity limits established for them. The methodological approaches addressed in this work allow optimizing radionuclide contents subject to mandatory measurement, or calculated determination of their activity in waste packages, and can be recommended to solve the problems when characterizing ChNPP’s historical waste transported for their final disposal.

Non-Destructive Techniques in the Tacis and Phare Nuclear Safety Programmes

2002 ◽  
Author(s):  
Michel Bie`th
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Slovak Republic ◽  
Nuclear Safety ◽  
Primary Circuit ◽  
Non Destructive ◽  
Service Inspection ◽  
Non Destructive Techniques

Decisions regarding the verification of design plant lifetime and potential license renewal periods involve a determination of the component and circuit condition. In Service Inspection of key reactor components becomes a crucial consideration for continued safe plant operation. The determination of the equipment properties by Non Destructive Techniques during periodic intervals is an important aspect of the assessment of fitness-for-service and safe operation of nuclear power plants The Tacis and Phare were established since 1991 by the European Union as support mechanisms through which projects could be identified and addressed satisfactorily. In Nuclear Safety, the countries mainly concerned are Russia, Ukraine, Armenia, and Kazakhstan for the Tacis programme, and Bulgaria, Czech Republic, Hungary, Slovak Republic, Lithuania, Romania and Slovenia for the Phare programme. The Tacis and Phare programs concerning the Nuclear Power Plants consist of: • On Site Assistance and Operational Safety, • Design Safety, • Regulatory Authorities, • Waste management, and are focused on reactor safety issues, contributing to the improvement in the safety of East European reactors and providing technology and safety culture transfer. The main parts of these programmes are related to the On-Site Assistance and to the Design Safety of VVVER and RBMK Nuclear power plants where Non Destructive Techniques for In Service Inspection of the primary circuit components are addressed.

Sign in / Sign up

Export Citation Format

Share Document