scholarly journals Uncertainty determination of fast neutron fluence onto the WWER pressure vessel metal surveillance specimens

2021 ◽  
Vol 22 (1) ◽  
pp. 42-47
Author(s):  
O.M. Pugach ◽  
◽  
S.M. Pugach ◽  
V.L. Diemokhin ◽  
V.N. Bukanov ◽  
...  
Keyword(s):  
Fast Neutron ◽  
Pressure Vessel ◽  
Neutron Fluence ◽  
Neutron Transport ◽  
Monte Carlo Code ◽  
Neutron Fluences ◽  
Surveillance Programs ◽  
Surveillance Specimens ◽  
Reactor Pressure

The standard surveillance programs of WWER reactors do not allow to measure the surveillance specimens irradiation conditions with the required accuracy. Therefore, the special methodology for the determination of the surveillance specimens irradiation conditions of the reactor pressure vessel metal has been developed by the specialists of the INR of NASU and is successfully applied. The developed methodology bases on the use of the Monte-Carlo code for neutron transport calculations to the surveillance specimens locations. The methodology improvement is described. The fundamentals of the calculation-experimental determination of the fast neutron fluences onto surveillance specimens and their uncertainties are presented.

3D Calculations of PWR Vessels Neutron Fluence With EFLUVE 3D Code

2013 ◽  
Author(s):  
Cécile-Aline Gosmain ◽  
Sylvain Rollet ◽  
Damien Schmitt
Keyword(s):  
Monte Carlo ◽  
Fast Neutron ◽  
Pressure Vessel ◽  
Reaction Rate ◽  
Neutron Fluence ◽  
Neutron Transport ◽  
Reaction Rates ◽  
Surveillance Program ◽  
Monte Carlo Code ◽  
Neutron Sources

In the framework of surveillance program dosimetry, the main parameter in the determination of the fracture toughness and the integrity of the reactor pressure vessel (RPV) is the fast neutron fluence on pressure vessel. Its calculated value is extrapolated using neutron transport codes from measured reaction rate value on dosimeters located on the core barrel. EDF R&D has developed a new 3D tool called EFLUVE3D based on the adjoint flux theory. This tool is able to reproduce on a given configuration the neutron flux, fast neutron fluence and reaction rate or dpa results of an exact Monte Carlo calculation with nearly the same accuracy. These EFLUVE3D calculations does the Source*Importance product which allows the calculation of the flux, the neutronic fluence (flux over 1MeV integrated on time) received at any point of the interface between the skin and the pressure vessel but also at the capsules of the pressurized water reactor vessels surveillance program and the dpa and reaction rates at different axial positions and different azimuthal positions of the vessel as well as at the surveillance capsules. Moreover, these calculations can be carried out monthly for each of the 58 reactors of the French current fleet in challenging time (less than 10mn for the total fluence and reaction rates calculations considering 14 different neutron sources of a classical power plant unit compared to more than 2 days for a classic Monte Carlo flux calculation at a given neutron source). The code needs as input: - for each reaction rate, the geometric importance matrix produced for a 3D pin by pin mesh on the basis of Green’s functions calculated by the Monte Carlo code TRIPOLI; - the neutron sources calculated on assemblies data (enrichment, position, fission fraction as a function of evolution), pin by pin power and irradiation. These last terms are based on local in-core activities measurements extrapolated to the whole core by use of the EDF core calculation scheme and a pin by pin power reconstruction methodology. This paper presents the fundamental principles of the code and its validation comparing its results to the direct Monte Carlo TRIPOLI results. Theses comparisons show a discrepancy of less than 0,5% between the two codes equivalent to the order of magnitude of the stochastic convergence of Monte Carlo results.

scholarly journals Safety Factors for Neutron Fluences in NPP Safety Assessment

2017 ◽  
Author(s):  
V. Diemiokhin ◽  
V. Bukanov ◽  
V. Ilkovych ◽  
A. Pugach
Keyword(s):  
Safety Factor ◽  
Safety Assessment ◽  
Neutron Fluence ◽  
Neutron Transport ◽  
Safety Factors ◽  
Neutron Fluences ◽  
Maximum Value ◽  
Transport Calculation ◽  
Calculation Code

In accordance with global practice and a number of existing regulations, the use of conservative approach is required for the calculations related to nuclear safety assessment of NPP. It implies the need to consider the determination of neutron fluence errors that is rather complicated. It is proposed to carry out the consideration by the way of multiplying the neutron fluences obtained with transport calculations by safety factors. The safety factor values are calculated by the developed technique based on the theory of errors, features of the neutron transport calculation code and the results obtained with the code. It is shown that the safety factor value is equal 1.18 with the confidence level of not less than 0.95 for the majority of VVER-1000 reactor places where neutron fluences are determined by MCPV code, and its maximum value is 1.25.

Verifying Calculation of Reactor Pressure Vessel Fast Neutron Fluence

2014 ◽  
Vol 986-987 ◽  
pp. 985-989
Author(s):  
Qiao Feng Liu ◽  
Jing Ru Han ◽  
Hai Ying Chen ◽  
Chun Ming Zhang
Keyword(s):  
Standard Deviation ◽  
Fast Neutron ◽  
Radiation Protection ◽  
Pressure Vessel ◽  
Neutron Fluence ◽  
Reactor Pressure Vessel ◽  
Light Water Reactor ◽  
Light Water ◽  
Water Reactor ◽  
Reactor Pressure

The reactor pressure vessel is an unchangeable component of the light water reactor. To some extent, the life of the pressure vessel depends on the fast neutron fluence. In addition, the fast neutron fluence is an important parameter for radiation protection. So, the fast neutron fluence is one of the main parameters which should be verifying calculated by the reviewers. The verifying calculation of the fast neutron fluence of one reactor pressure vessel is presented in this paper, and the standard deviation between the verifying and designing calculations is lower than 10%. The reasons for the deviation are discussed.

On the effects of oxygen cross-sections in the fast neutron fluence analysis for a reactor pressure vessel scraping test

2008 ◽  
Vol 35 (4) ◽  
pp. 565-569 ◽  
Author(s):  
Alexander Vasiliev ◽  
Hakim Ferroukhi ◽  
Martin A. Zimmermann ◽  
Rakesh Chawla
Keyword(s):  
Fast Neutron ◽  
Cross Sections ◽  
Pressure Vessel ◽  
Neutron Fluence ◽  
Reactor Pressure Vessel ◽  
Reactor Pressure

scholarly journals To the issue of accuracy of determination of irradiation embrittlement coefficient

2020 ◽  
Vol 21 (3) ◽  
pp. 245-248
Author(s):  
L.I. Chyrko ◽  
◽  
V.M. Revka ◽  
Yu.V. Chaikovskyi ◽  
M.G. Goliak ◽  
...  
Keyword(s):  
Fracture Toughness ◽  
Statistical Analysis ◽  
Fast Neutron ◽  
Neutron Fluence ◽  
Reactor Vessel ◽  
Irradiation Embrittlement ◽  
Radiation Induced ◽  
The Impact ◽  
Surveillance Specimens

The paper presents the statistical analysis of experimental results of radiation-induced critical brittle temperature ΔTF shifts and reference temperatures ΔT0 obtained respectively from the impact bend and fracture toughness tests of the reactor vessel metal surveillance specimens to define the possibility of their mutual application for the irradiation embrittlement coefficient to be determined more accurately. The correlation between these parameters is shown to remain up to the accumulation of over-design fast neutron fluence.

scholarly journals Analysis for the Effect of Spatial Discretization Method on AP1000 Reactor Pressure Vessel Fluence Calculation

2016 ◽  
Vol 2016 ◽  
pp. 1-10 ◽  
Author(s):  
Junxiao Zheng ◽  
Bin Zhang ◽  
Shengchun Shi ◽  
Yixue Chen
Keyword(s):  
Fast Neutron ◽  
Pressure Vessel ◽  
Power Plants ◽  
Structural Integrity ◽  
Neutron Fluence ◽  
Mesh Size ◽  
Reactor Pressure Vessel ◽  
Life Extension ◽  
Discrete Ordinates ◽  
Reactor Pressure

Maintaining the structural integrity of the reactor pressure vessel (RPV) is a critical concern related to the safe operation of nuclear power plants. To estimate the structural integrity over the designed lifetime and to support analyses for a potential plant life extension, an accurate calculation of the fast neutron fluence (E>1.0 MeV orE>0.1 MeV) at the RPV is significant. The discrete ordinates method is one of the main methods to solve such problems. During the calculation process, many factors will affect the results. In this paper, the deviations introduced by different differencing schemes and mesh sizes on the AP1000 RPV fast neutron fluence have been studied, which are based on new discrete ordinates code ARES. The analysis shows that the differencing scheme (diamond difference with or without linear zero fix-up, theta weighted, directional theta weighted, and exponential directional weighted) introduces a deviation within 4%. The coarse mesh (4 × 4 cm meshes inXYplane) leads to approximately 23.7% calculation deviation compared to those of refined mesh (1 × 1 cm meshes inXYplane). Comprehensive study on the deviation introduced by differencing scheme and mesh size has great significance for reasoned evaluation of RPV fast neutron fluence calculation results.

Ex-Vessel Neutron Dosimetry Results in the Vicinity of RPV Supports

2007 ◽  
Author(s):  
Daniel N. Hopkins ◽  
Eugene T. Hayes ◽  
Arnold H. Ferro
Keyword(s):  
Fast Neutron ◽  
Cross Section ◽  
Pressure Vessel ◽  
Transport Theory ◽  
Neutron Fluence ◽  
Reactor Pressure Vessel ◽  
Life Extension ◽  
Neutron Dosimetry ◽  
Pressurized Water ◽  
Reactor Pressure

Neutron-induced embrittlement of the reactor pressure vessel has been a long standing concern for pressurized water reactors (PWR). To date, the beltline region of the pressure vessel, defined as the portion of the pressure vessel experiencing fast neutron fluence (E > 1.0 MeV) equal to or greater than 1017 n/cm2, has been the primary focus of evaluations assessing this embrittlement. These evaluations typically include a calculation of the neutron flux incident on the reactor pressure vessel beltline region, which is in part validated by direct comparison with dosimetry measurements. Two general types of measurements are commonly used, those being dosimetry sets that are included as part of the in-vessel surveillance capsules, and at some plants, those that are included in supplemental surveillance programs such as Ex-Vessel Neutron Dosimetry. In the context of life extension, the beltline region as defined above is getting larger. Present fluence calculations for a number of plants indicate that beltline region at the end of the 60 years of operation will extend to the bottom of the reactor pressure vessel nozzle welds. This extended beltline creates a new problem in terms of validating the neutron fluence calculations in this region well above the top of the active fuel, in that there are no measurements available to confirm calculated results in this new region of interest. Prior to the start-up of Cycle 11 at Comanche Peak Unit 1, an Ex-Vessel Neutron Dosimetry Program was initiated. This program included placement of neutron dosimetry sensor sets in the vicinity of the reactor pressure vessel supports. At the conclusion of Cycle 11, the first set of dosimetry was replaced and the irradiated set analyzed. The Ex-Vessel Neutron Dosimetry set from Cycle 11 was analyzed using a 2D/1D flux synthesis technique using the two dimensional discrete ordinates transport theory calculations (DORT) along with the BUGLE 96 cross-section library and the SNLRML neutron dosimetry cross-section library. The measurements in the vicinity of the vessel supports compare well with the transport calculations, thus confirming that the expected fast neutron fluence (E > 1.0 MeV) in the vicinity of the reactor vessel supports is below the 1018 n/cm2.

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