The biserm cross-section library for analyzing radiation damage to materials caused by nucleons with energies up to 800 MeV

One-dimensional calculations of neutron and gamma group fluxes and radiation damage in terms of DPA have been carried out for VVER-1000 mock-up realized at LR-0 critical assembly, Rez, Czech Republic in order to evaluate the influence of energy dependence multigroup approximation on VVER-1000 RPV Neutron/Gamma flux calculation. Two calculational methods have been applied: Monte Carlo method (MC) with continuous energy dependence description and discrete ordinates (DO) method with multigroup energy dependence approximation. Two different multigroup neutron/gamma cross section libraries have been used: BGL1000—created especially for VVER-1000 application and BUGLE96—created for PWR application. A good consistency between the results for both methods and different cross section libraries is obtained for neutrons with energy above 0.1 MeV. The worst consistency is obtained for the thermal neutrons. For gammas a better consistency between DO and MC results is obtained when BGL1000 library is applied.

Evaluation of neutron radiation damage in the VVER-1200 reactor pressure vessel

Description Presents the most up to date research available from the world's leading experts in neutron and gammaray physics and dosimetry. This state-of-the-art review features over 85 peer-reviewed papers on such wide-ranging subjects as: Pressure Vessel Surveillance Dosimetry -- 17papers, including review papers on the current status of surveillance standards, several papers on the status of surveillance programs in central and eastern European countries, and a number of papers on handling uncertainties in surveillance data. Passive and Active Neutron Dosimetry Techniques -- two sections containing papers that describe some of the newest techniques in neutron metrology. Benchmarks -- includes papers describing the analysis of benchmark neutron environments for verification and validation of measurement and calculational results. Gamma-Ray Dosimetry Techniques -- 5 papers describing some of the latest techniques for determining gamma-ray fluences in reactor environments. Radiation Damage Evaluation and Measurements-- contains papers correlating various aspects of radiation damage with fluence. Radiation Field Characterization -- 8 papers describe analyses in a wide variety of neutron environments, ranging from neutron fields in light-water moderated research reactors to photoneutron fields produced by an electron accelerator. Nuclear Data -- papers on evaluations of specific cross sections important for dosimetry applications, as well as evaluation and testing of updated dosimetry cross-section libraries. In addition, this section includes 2 papers on the evaluation and testing of the most recent multigroup cross-section libraries used for neutron and gamma-ray transport calculations. High-Energy Neutron Dosimetry -- 5 papers that support research in fusion energy. Three keynote presentations and summaries of nine workshop meetings are also included in this comprehensive edition.

Presents the most up to date research available from the world's leading experts in neutron and gammaray physics and dosimetry. This state-of-the-art review features over 85 peer-reviewed papers on such wide-ranging subjects as: Pressure Vessel Surveillance Dosimetry -- 17 papers, including review papers on the current status of surveillance standards, several papers on the status of surveillance programs in central and eastern European countries, and a number of papers on handling uncertainties in surveillance data. Passive and Active Neutron Dosimetry Techniques -- two sections containing papers that describe some of the newest techniques in neutron metrology. Benchmarks -- includes papers describing the analysis of benchmark neutron environments for verification and validation of measurement and calculational results. Gamma-Ray Dosimetry Techniques -- 5 papers describing some of the latest techniques for determining gamma-ray fluences in reactor environments. Radiation Damage Evaluation and Measurements-- contains papers correlating various aspects of radiation damage with fluence. Radiation Field Characterization -- 8 papers describe analyses in a wide variety of neutron environments, ranging from neutron fields in light-water moderated research reactors to photoneutron fields produced by an electron accelerator. Nuclear Data -- papers on evaluations of specific cross sections important for dosimetry applications, as well as evaluation and testing of updated dosimetry cross-section libraries. In addition, this section includes 2 papers on the evaluation and testing of the most recent multigroup cross-section libraries used for neutron and gamma-ray transport calculations. High-Energy Neutron Dosimetry -- 5 papers that support research in fusion energy. Three keynote presentations and summaries of nine workshop meetings are also included in this comprehensive edition.

This paper presents a synthesis of the ENEA-Bologna Nuclear Data Group programme dedicated to generate and validate group-wise cross section libraries for shielding and radiation damage deterministic calculations in nuclear fission reactors, following the data processing methodology recommended in the ANSI/ANS-6.1.2-1999 (R2009) American Standard. The VITJEFF311.BOLIB and VITENDF70.BOLIB finegroup coupled n-γ (199 n + 42 γ – VITAMIN-B6 structure) multi-purpose cross section libraries, based on the Bondarenko method for neutron resonance self-shielding and respectively on JEFF-3.1.1 and ENDF/B-VII.0 evaluated nuclear data, were produced in AMPX format using the NJOY-99.259 and the ENEA-Bologna 2007 Revision of the SCAMPI nuclear data processing systems. Two derived broad-group coupled n-γ (47 n + 20 γ – BUGLE-96 structure) working cross section libraries in FIDO-ANISN format for LWR shielding and pressure vessel dosimetry calculations, named BUGJEFF311.BOLIB and BUGENDF70.BOLIB, were generated by the revised version of SCAMPI, through problem-dependent cross section collapsing and self-shielding from the cited fine-group libraries. The validation results on the criticality safety benchmark experiments for the fine-group libraries and the preliminary validation results for the broad-group working libraries on the PCA-Replica and VENUS-3 engineering neutron shielding benchmark experiments are reported in synthesis.

Neutron spectrum measurements have been performed by the multiple-foil activation method at the spallation neutron source SINQ (PSI, Switzerland). The energy region of interest ranges from thermal energies to 590 MeV. A new high-energy neutron cross-section library has been prepared for use at spallation neutron sources. First a master library, called SPALLDOS, was produced, containing cross-sections for 239 neutron activation and radiation damage characteristic reactions. From the master library, problem dependent cross-section libraries — corresponding to the interesting measurement positions at SINQ — with 138 reactions were derived. Part of the cross-section data was tested by experiments.

It is known that the blanket module in the fusion reactor is designed for slowing and absorbing neutrons, breeding tritium, and limiting radiation damage to other structures. These blankets are installed remotely at individual positions. In this process, gaps between the blanket modules can occur and the streaming effect through gaps has an effect on the performance of the fusion reactor. The streaming effect affects two aspects in the neutron transport problem: (1) activation of the fusion reactor structures and (2) radiation shielding. It is noted that the shape of the gap is locally different due to the specific shape of the fusion reactor. Also, the mechanical difference of the blanket modules changes the gap size. In this study, the streaming effect caused by different shape and size of the gap are evaluated for a K-DEMO fusion reactor. The blanket modules of the fusion reactor, which are the blanket, vacuum vessel, and water channel, were simply simulated. In each module, the neutron streaming effects were assessed by changing the gap size and shapes. The MCNP5 code was used for the evaluations with ENDF/B-VI cross-section library and SAB2002 thermal cross section library. It was noticed that the neutron flux was significantly changed due to the streaming effect. This study can give guidelines for the simplification of blanket modeling in radiation transport problems.

Adjoint Monte Carlo calculations have been carried out using the three-dimensional radiation transport code, MORSE, to estimate the nuclear heating and radiation damage in the toroidal field (TF) coils adjacent to a 28- × 68-cm2 rectangular neutral beam injector duct that passes through the blanket and shield of a deuterium-tritium (D-T) burning tokamak reactor. The plasma region, blanket, shield, and TF coils were represented in cylindrical geometry using the same dimensions and compositions as those of the Experimental Power Reactor. The radiation transport was accomplished using coupled 35-group neutron, 21-group gamma-ray cross sections obtained by collapsing the DLC-37 cross-section library. Nuclear heating rates were obtained using fluence-to-kerma factors generated by the computer codes MACK and SMUG, and radiation damage rates were calculated using damage response functions generated by the computer code RECOIL. The presence of the neutral beam injector duct leads to increases in the nuclear heating rates in the TF coils ranging from a factor of 3 to a factor of 196, depending on the location. Increases in the radiation damage also result in the TF coils. The atomic displacement rates increase by factors of 2 to 138 and the hydrogen and helium gas production rates increase from factors of 11 to 7600 and from 15 to 9700, respectively.

The bulk shielding benchmark experiment at the Frascati Neutron Generator (FNG)

High-temperature and high-density plasmas are achieved by means of real-time control, fast diagnostic, and high-power heating systems. Those systems are precisely controlled via highly integrated electronic components, but can be seriously affected by radiation damage. Therefore, the effects of irradiation on currently used electronic components should be investigated for the control and measurement of Large Helical Device (LHD) deuterium plasmas. For the precise estimation of the radiation field in the LHD torus hall, the MCNP6 code is used with the cross-section library ENDF B-VI. The geometry is modeled on the computer-aided design. The dose on silicon, which is a major ingredient of electronic components, over nine years of LHD deuterium operation shows that the gamma-ray contribution is dominant. Neutron irradiation tests were performed in the OKTAVIAN at Osaka University and the Fast Neutron Laboratory at Tohoku University. Gamma-ray irradiation tests were performed at the Nagoya University Cobalt-60 irradiation facility. We found that there are ethernet connection failures of programmable logic controller (PLC) modules due to neutron irradiation with a neutron flux of 3 × 106 cm−2 s−1. This neutron flux is equivalent to that expected at basement level in the LHD torus hall without a neutron shield. Most modules of the PLC are broken around a gamma-ray dose of 100 Gy. This is comparable with the dose in the LHD torus hall over nine years. If we consider the dose only, these components may survive more than nine years. For the safety of the LHD operation, the electronic components in the torus hall have been rearranged.

Two broad-group coupled neutron/photon working cross section libraries in FIDO-ANISN format, dedicated to LWR shielding and pressure vessel dosimetry applications, were generated following the methodology recommended by the US ANSI/ANS-6.1.2-1999 (R2009) standard. These libraries, named BUGJEFF311.BOLIB and BUGENDF70.BOLIB, are respectively based on JEFF-3.1.1 and ENDF/B-VII.0 nuclear data and adopt the same broad-group energy structure (47 n + 20 γ) of the ORNL BUGLE-96 similar library. They were respectively obtained from the ENEA-Bologna VITJEFF311.BOLIB and VITENDF70.BOLIB libraries in AMPX format for nuclear fission applications through problem-dependent cross section collapsing with the ENEA-Bologna 2007 revision of the ORNL SCAMPI nuclear data processing system. Both previous libraries are based on the Bondarenko self-shielding factor method and have the same AMPX format and fine-group energy structure (199 n + 42 γ) as the ORNL VITAMIN-B6 similar library from which BUGLE-96 was obtained at ORNL. A synthesis of a preliminary validation of the cited BUGLE-type libraries, performed through 3D fixed source transport calculations with the ORNL TORT-3.2 SN code, is included. The calculations were dedicated to the PCA-Replica 12/13 and VENUS-3 engineering neutron shielding benchmark experiments, specifically conceived to test the accuracy of nuclear data and transport codes in LWR shielding and radiation damage analyses.

Preliminary evaluation of the expected radiation damage of the bayonet IFMIF back-plate

The PCA-Replica 12/13 (H 2 O/Fe) neutron shielding benchmark experiment was analysed using the ORNL TORT-3.2 3D S N code. PCA-Replica, specifically conceived to test the accuracy of nuclear data and transport codes employed in LWR shielding and radiation damage calculations, reproduces a PWR ex-core radial geometry with alternate layers of water and steel including a PWR pressure vessel simulator. Three broad-group coupled neutron/photon working cross section libraries in FIDO-ANISN format with the same energy group structure (47 n + 20 γ) and based on different nuclear data were alternatively used: the ENEA BUGJEFF311.BOLIB (JEFF-3.1.1) and BUGENDF70.BOLIB (ENDF/B-VII.0) libraries and the ORNL BUGLE-96 (ENDF/B-VI.3) library. Dosimeter cross sections derived from the IAEA IRDF-2002 dosimetry file were employed. The calculated reaction rates for the Rh-103(n,n′)Rh-103 m, In-115(n,n′)In-115m and S-32(n,p)P-32 threshold activation dosimeters and the calculated neutron spectra are compared with the corresponding experimental results.

DPA cross section library FermiDPA 1.0 based on the industry standard NRT model calculations is described. The library contains DPA cross sections for neutrons in the energy range 10$$^{-5}$$ eV 20 (150) MeV. Calculations used neutron-induced reaction cross sections from ENDFB-VII database of evaluated nuclear data. The NJOY99 nuclear data processing system's module HEATR was applied to calculate NRT model radiation damage cross sections. The FermiDPA 1.0 library is a database of 395 text files (for 395 known isotopes) with DPA cross sections. It is code-independent and can be implemented in any transport code.

Cross-section libraries for study of radiation damage, activation, and transmutation of materials irradiated by high-energy particles