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 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.

scholarly journals Evaluation of PWR pressure vessel fast neutron fluence benchmarks from NUREG/CR-6115 with ares transport code

2017 ◽  
Vol 32 (3) ◽  
pp. 204-210 ◽  
Author(s):  
Liang Zhang ◽  
Bin Zhang ◽  
Cong Liu ◽  
Yixue Chen
Keyword(s):  
Fast Neutron ◽  
Pressure Vessel ◽  
Neutron Fluence ◽  
Neutron Transport ◽  
Source Distribution ◽  
Discrete Ordinates ◽  
Source Modeling ◽  
Transport Code ◽  
Vessel Integrity ◽  
Angular Discretization

An accurate evaluation of PWR pressure vessel fast neutron fluence is essential to ensure pressure vessel integrity over the design lifetime. The discrete ordinates method is one of the main methods to treat such problems. In this paper, evaluations have been performed for three PWR benchmarks described in NUREG/CR-6115 using ARES transport code. The calculated results were compared to the reference values and a satisfactory agreement was obtained. In addition, the effects of SN numeric and source distribution modeling for pressure vessel fast neutron fluence calculation are investigated. Based on the fine enough grids adopted, the different spatial and angular discretization introduces derivations less than 3 %, and fix-up for negative scattering source causes no noticeable effects when calculating pressure vessel fast neutron fluence. However, the discrepancy of assembly-wise and pin-wise source modeling for peripheral assemblies reaches ~20 %, which indicates that pin-wise modeling for peripheral assemblies is essential. These results provide guidelines for pressure vessel fast neutron fluence calculation and demonstrate that the ARES transport code is capable of performing neutron transport calculations for evaluating PWR pressure vessel fast neutron fluence.

Fast neutron fluence calculations as support for a BWR pressure vessel and internals surveillance program

2010 ◽  
Vol 240 (6) ◽  
pp. 1271-1280 ◽  
Author(s):  
Marco A. Lucatero ◽  
Javier C. Palacios-Hernández ◽  
Javier Ortiz-Villafuerte ◽  
J. Vicente Xolocostli-Munguía ◽  
Armando M. Gómez-Torres
Keyword(s):  
Fast Neutron ◽  
Pressure Vessel ◽  
Neutron Fluence ◽  
Surveillance Program

scholarly journals A Fast Neutron Fluence Reduction Strategy for Pressure Vessel Lifetime Extension

2000 ◽  
Vol 37 (sup1) ◽  
pp. 120-124 ◽  
Author(s):  
Jong Kyung Kim ◽  
Chang Ho Shin ◽  
Bo Kyun Seo ◽  
Myung Hyun Kim ◽  
Goung Jin Lee
Keyword(s):  
Fast Neutron ◽  
Pressure Vessel ◽  
Neutron Fluence ◽  
Reduction Strategy ◽  
Lifetime Extension

scholarly journals Predicting Ex-core Detector Response in a PWR with Monte Carlo Neutron Transport Methods

2020 ◽  
Vol 225 ◽  
pp. 03007
Author(s):  
Tanja Goričanec ◽  
Domen Kotnik ◽  
Žiga Štancar ◽  
Luka Snoj ◽  
Marjan Kromar
Keyword(s):  
Monte Carlo ◽  
Neutron Transport ◽  
Reactor Core ◽  
Response Prediction ◽  
Detector Response ◽  
Design Calculation ◽  
Monte Carlo Code ◽  
Control Rod ◽  
The Core ◽  
Fuel Assemblies

An approach for calculating ex-core detector response using Monte Carlo code MCNP was developed. As a first step towards ex-core detector response prediction a detailed MCNP model of the reactor core was made. A script called McCord was developed as a link between deterministic program package CORD-2 and Monte Carlo code MCNP. It automatically generates an MCNP input from the CORD-2 data. A detailed MCNP core model was used to calculate 3D power distributions inside the core. Calculated power distributions were verified by comparison to the CORD-2 calculations, which is currently used for core design calculation verification of the Krško nuclea power plant. For the hot zero power configuration, the deviations are within 3 % for majority of fuel assemblies and slightly higher for fuel assemblies located at the core periphery. The computational model was further verified by comparing the calculated control rod worth to the CORD-2 results. The deviations were within 50 pcm and considered acceptable. The research will in future be supplemented with the in-core and ex-core detector signal calculations and neutron transport outside the reactor core.

Pressure Vessel Investigations of the Former Greifswald NPP: Fluence Calculations and Niobium Based Fluence Measurements

2006 ◽  
Author(s):  
J. Konheiser ◽  
U. Rindelhardt ◽  
H.-W. Viehrig ◽  
B. Boehmer ◽  
B. Gleisberg
Keyword(s):  
Fast Neutron ◽  
Pressure Vessel ◽  
Half Life ◽  
Branching Ratio ◽  
Assessment Tools ◽  
Surveillance Program ◽  
Integrity Assessment ◽  
Icp Ms ◽  
Nb Content

Pressure vessel integrity assessment after long-term service irradiation is commonly based on surveillance program results. Nevertheless, only the investigation of RPV material from decommissioned NPPs enables the evaluation of the real toughness response. Such a chance is given now through the investigation of material from the former Greifswald NPP (VVER-440/230) to evaluate the material state of a standard RPV design and to assess the quality of prediction rules and assessment tools. The operation of the four Greifswald units was finished in 1991 after 12–15 years of operation. In autumn 2005 the first trepans (diameter 120 mm) were gained from the unit 1 of this NPP. Some details of the trepanning procedure will be given. The paper mainly deals with the retrospective dosimetry based on Niobium, which is a trace element of the RPV material. The reaction 93Nb(n,n′)93mNb with an energy dependence highly correlated to radiation damage and a half life of the reaction product of 16.13 years is well suited for retrospective fast neutron dosimetry. Fluence calculations using the code TRAMO were based on pin-wise time dependent neutron sources and an updated nuclear data base (ENDF/B-VI release 8). The neutron spectra were determined at the trepan positions. The different loading schemes of unit 1 (standard and with 4 or 6 dummy assemblies) were taken into account. The calculated specific 93mNb activities for February, 2006 at the sample positions were determined to 16.3 Bq/μg Nb for sample 1, (0.1cm distance from inner wall), and 4.0 Bq/μg Nb for sample 2 (11.5 cm distance from inner wall). Unfortunately, a second neutron reaction besides 93Nb(n,n′) leading to 93mNb-activity is the reaction 92Mo(n,γ)93Mo. 93Mo decays by electron capture to 93mNb with a half life of 4000 years and a branching ratio br = 0.88. As (n,γ)-reactions are produced mainly by low energy neutrons, being less important for material damage, the 93mNb-activity generated through the Mo-path should be determined separately and subtracted from the measured activity. For the sample 1 in the maximum flux azimuthal position of weld SN4 with a Nb-content of 8 ppm and an Mo-content of 4000 ppm for February 3, 2006 was obtained a Mo-induced 93mNb-activity of 80 Bq/g steel, amounting to 37.7% of the total 93mNb-activity. It turns out that the 93mNb generation on the second path is nearly of the same order as the fast neutron induced generation from Niobium. For the experimental determination of the 93mNb-activity the Nb-content was determined by ICP-MS (inductive coupled plasma mass spectrometry) after dissolution of the material sample. The radiochemical isolation of Nb was done by anion exchange separation. The radiochemical separation was accompanied by determination of the chemical yield of Nb using again the ICP-MS method. The measurement of the 93mNb activity was realized by Liquid Scintillation Spectrometry (LSC). The first comparison between the calculated and the measured 93mNb activities resulted in deviations between 15 and 50%. Possible reasons for the observed differences are discussed.

scholarly journals Collisions of Nucleons with Atoms: Calculated Cross Sections and Monte Carlo Simulation

2021 ◽  
Vol 9 ◽  
Author(s):  
Francesc Salvat ◽  
José Manuel Quesada
Keyword(s):  
Monte Carlo ◽  
Cross Sections ◽  
Nuclear Reactions ◽  
Impact Ionization ◽  
Neutron Transport ◽  
Emission Spectra ◽  
Monte Carlo Code ◽  
Dose Distributions ◽  
Elastic Collisions ◽  
Proton Impact

After a summary description of the theory of elastic collisions of nucleons with atoms, we present the calculation of a generic database of differential and integrated cross sections for the simulation of multiple elastic collisions of protons and neutrons with kinetic energies larger than 100 keV. The relativistic plane-wave Born approximation, with binding and Coulomb-deflection corrections, has been used to calculate a database of proton-impact ionization of K-shell and L-, M-, and N-subshells of neutral atoms These databases cover the whole energy range of interest for all the elements in the periodic system, from hydrogen to einsteinium (Z = 1–99); they are provided as part of the penh distribution package. The Monte Carlo code system penh for the simulation of coupled electron-photon-proton transport is extended to account for the effect of the transport of neutrons (released in proton-induced nuclear reactions) in calculations of dose distributions from proton beams. A simplified description of neutron transport, in which neutron-induced nuclear reactions are described as a fractionally absorbing process, is shown to give simulated depth-dose distributions in good agreement with those generated by the Geant4 code. The proton-impact ionization database, combined with the description of atomic relaxation data and electron transport in penelope, allows the simulation of proton-induced x-ray emission spectra from targets with complex geometries.

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