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The Micro Modular Reactor (MMR) represents a new paradigm of Ultra Safe Nuclear power with intrinsic safety. The safety starts from its revolutionary ceramic fuel, that cannot melt and makes it impossible to release radioactive elements in the environment. The elimination of water as a heat transport fluid eliminates the possibility of chemical explosions. Uranium 238, that absorbs more neutrons at higher temperatures shuts down the chain reaction without damage even in case of coolant loss at full power. These characteristics make MMR ideal to provide heat process directly inside factories, to replace natural gas, and to provide power to small communities. [ABSTRACT FROM AUTHOR]

The physics of nuclear fission will be briefly illustrated, from the basic mechanism behind this phenomenon to the relevant physical quantities like nuclear cross sections, neutron flux and reaction products, together with the accompanying phenomenon of neutron capture and its role in determining how the fuel transforms in a nuclear reactor. The basic concepts underlying the operation of different types of nuclear reactors will be illustrated, along with the concept of fuel cycle. The aspects of radioactive waste, fuel resources and safety will also be briefly illustrated. [ABSTRACT FROM AUTHOR]

The Micro Modular Reactor (MMR) represents a new paradigm of Ultra Safe Nuclear power with intrinsic safety. The safety starts from its revolutionary ceramic fuel, that cannot melt and makes it impossible to release radioactive elements in the environment. The elimination of water as a heat transport fluid eliminates the possibility of chemical explosions. Uranium 238, that absorbs more neutrons at higher temperatures shuts down the chain reaction without damage even in case of coolant loss at full power. These characteristics make MMR ideal to provide heat process directly inside factories, to replace natural gas, and to provide power to small communities. [ABSTRACT FROM AUTHOR]

The characterization of the neutron, prompt gamma-ray, and delayed gamma-ray radiation fields for the White Sands Missile Range (WSMR) Fast Burst Reactor, also known as molybdenum-alloy Godiva (Molly-G) has been assessed at the 6-inch irradiation location. The neutron energy spectra, uncertainties, and common radiation metrics are presented. Code-dependent recommended constants are given to facilitate the conversion of various dosimetry readings into radiation metrics desired by experimenters. The Molly-G core was designed and configured similarly to Godiva II, as an unreflected, unmoderated, cylindrical annulus of uranium-molybdenum-alloy fuel with molybdenum loading of 10%. At the 6-inch position, the axial fluence maximum is about 2.4×1013 n/cm2 per MJ of reactor energy; about 0.1% of the neutron fluence is below 1 keV and 96% is above 100 keV. The 1-MeV Damage-Equivalent Silicon (DES) fluence is estimated at 2.2×1013 n/cm2 per MJ of reactor energy. The prompt gamma-ray dose is roughly 2.5E+03 rad(Si) per MJ and the delayed gamma-ray dose is about 1.3E+03 rad(Si) per MJ. [ABSTRACT FROM AUTHOR]

The characterization of the neutron, prompt gamma-ray, and delayed gamma-ray radiation fields in the University of Texas at Austin Nuclear Engineering Teaching Laboratory (NETL) TRIGA reactor for the beam port (BP) 1/5 free-field environment at the 128-inch location adjacent to the core centerline has been accomplished. NETL is being explored as an auxiliary neutron test facility for the Sandia National Laboratories radiation effects sciences research and development campaigns. The NETL reactor is a TRIGA Mark-II pulse and steady-state, above-ground pool-type reactor. NETL is intended as a university research reactor typically used to perform irradiation experiments for students and customers, radioisotope production, as well as a training reactor. Initial criticality of the NETL TRIGA reactor was achieved on March 12, 1992, making it one of the newest test reactor facilities in the US. The neutron energy spectra, uncertainties, and covariance matrices are presented as well as a neutron fluence map of the experiment area of the cavity. For an unmoderated condition, the neutron fluence at the center of BP 1/5, at the adjacent core axial centerline, is about 8.2×1012 n/cm2 per MJ of reactor energy. About 67% of the neutron fluence is below 1 keV and 22% above 100 keV. The 1-MeV Damage-Equivalent Silicon (DES) fluence is roughly 1.6×1012 n/cm2 per MJ of reactor energy. [ABSTRACT FROM AUTHOR]

The Advanced Test Reactor (ATR) has a wide variety of irradiation positions that have had experiments that were developed by users from around the world. Most experiment irradiations rely on thoroughly benchmarked numerical models. However, some irradiation positions in ATR are not as well- characterized and are complicated by spectral perturbations from control cylinder orientation. The "small I" and "large-B" irradiation positions are located nearby control cylinders and suffer from these flux perturbations from control cylinder orientations that change during an irradiation cycle to maintain the desired core power distribution. Models predicted that these types of position would exhibit both spectral shifts and amplitude changes in neutron flux, but few measurements have been conducted to benchmark these predictions. Recently, requalification testing was performed to confirm the operational readiness of the ATR following the completion of the Core Internals Change-out (CIC). These tests provided a unique opportunity to validate the analytical methods used to simulate the ATR because nearly all components in the reactor were in a clean as-built state, significantly reducing modelling uncertainties. One subset of the testing included characterization of the fast and thermal neutron flux in the "small-I" and "large-B" positions using silver, cobalt, and nickel neutron dosimetry. In contrast to typical irradiation cycles, the control cylinders were held in position during the post-CIC nuclear testing. The specific activity of these dosimeter wires was measured following two low-power tests with different control cylinder positions. [ABSTRACT FROM AUTHOR]

This work compares the results of radiation transport calculations performed with VITAMIN-B7 cross section data to calculations performed with a similar library prepared with the TRANSX code. The bases for the comparisons include a simplified, one-dimensional PWR model and the industry standard H.B. Robinson Cycle 9 reactor dosimetry benchmark. Comparisons are also made with a modified TRANSX-generated library with enhanced background cross section data. The results show significant differences between VITAMIN and TRANSX-generated libraries through iron media. [ABSTRACT FROM AUTHOR]

Activation data from seven different reactor-based reference neutron fields are examined to provide enhanced validation evidence for the newly released IRDFF-II library. A least-squares based spectrum adjustment methodology is used and rigorous statistical metrics demonstrate the consistency of the set of IRDFF-II dosimetry cross sections. The use of these reactor-based neutron fields provides validation evidence for nine more reactions than have been addressed in the 252Cf(sf) and 235U(th) benchmark validation testing. The use of covers to shift their energy response increases the power of the response validation but makes it challenging to properly capture response correlations. [ABSTRACT FROM AUTHOR]

This paper deals with the measurement of Spectrum Averaged Cross Sections in two different neutron fields formed in zero power reactors. The first was Benchmark Neutron Reference Field in the LR-0 reactor, and the second field was in the center of the vertical channel touching the fuel in the VR-1 reactor. The spectrum averaged cross section differs for both cases as the spectra differ, but after normalization to 235U PFNS using calculated correction, both results are in good agreement, thus confirming the spectra in both cases are similar in the 1 – 14 MeV region. A good agreement between lower threshold reactions averaged in actual reactor spectra and prompt fission neutron spectrum of 235U is reported as well. [ABSTRACT FROM AUTHOR]

This paper presents an overview of the Reactor Pressure Vessel surveillance program as it is applied to the seven Belgian Pressurised Water Reactors (PWR). The first part of the paper recalls the original objectives of the surveillance program, detailing the number of capsules that were installed per reactor, the number that have been retrieved so far and discusses the type of dosimeters installed. The second part is a description of the experimental techniques applied to determine the neutron fluences, a work which is performed in the reactor dosimetry laboratory at SCK CEN and is based on qualified activity measurements. Finally, the paper ends by briefly presenting the calculation scheme developed at Tractebel and by providing a statistical analysis of the C/E values over the ~ 400 in-core dosimeters and 800 ex-core dosimeters that have been analysed to date. The method offers satisfactory results with an average C/E on all dosimeters close to 1.0 and a standard deviation comprised between 8 % and 10 % depending on the dosimeter considered. [ABSTRACT FROM AUTHOR]

The operational lifespan of all four units at the Hungarian Paks Nuclear Power Plant (NPP) has been extended by 20 years beyond the original 30-year term. This extension has been implemented gradually, in one- to two-year increments, since 2013. The Hungarian Atomic Energy Authority, responsible for granting approval for these life extensions, has prescribed the continuous and rigorous monitoring of the reactor pressure vessel (RPV) throughout the plant's extended service life. Among the key requirements is the validation of the neutron transport calculations, with specific attention to the large Lead Factor (LF), which reaches around 12 at the reactor core's center. To meet these regulatory requirements, neutron transport calculations and reactor dosimetry measurements using neutron activation detectors (neutron monitors) have been conducted both inside and outside the RPV. Inside measurements were taken in the RPV surveillance channel, while outside measurements were conducted in the cavity adjacent to the outer wall of the RPV. These efforts aim to determine the key parameters of the neutron irradiation and validation is achieved if the experimental results agree with the neutron transport calculations at both the surveillance position and the RPV's outer wall (in the cavity), ensuring that the calculations are also valid for the inner wall of the RPV. This paper presents the validation of experimental data used to confirm the accuracy of the neutron transport calculations. The results of irradiations conducted over one- and four-year periods are compared and analyzed to demonstrate the authenticity of the experimental data for validation of the calculation results. [ABSTRACT FROM AUTHOR]

The United States Nuclear Regulatory Commission (USNRC) has expressed concerns about the calculated fast neutron fluence exposure uncertainties for the reactor vessel internals (RVI) of light water reactors for extended operation [1, 2]. The current guidance for calculating neutron fluence, Regulatory Guide (RG) 1.190 [3], was developed with a focus on the reactor pressure vessel (RPV) region surrounding the active core, which may not be adequate for the RVI above and below the active core during the license renewal periods, especially the subsequent license renewal (SLR) period. An updated version of ANSI/ANS 19-10 [4] may fill the regulatory guidance gap on neutron fluence determination methodology for RVI but still faces the challenges of lacking RVI fluence measurement benchmark data and analytical uncertainty analysis, which are two cornerstones for qualifying a neutron fluence determination methodology. The Pressurized Water Reactor Owners Group (PWROG) has responded to this regulatory concern and have performed analytical uncertainty analysis and measurement benchmark work to address these two technical challenges for the nuclear power industry. [ABSTRACT FROM AUTHOR]

The present paper summarizes the experience accumulated at the Laboratory for Reactor Physics and Thermal-Hydraulics (LRT) of the Paul Scherrer Institute in the field of reactor dosimetry, as well as outlines the recent progress achieved in relation to the associated nuclear data uncertainty propagation methodologies. For dosimetry simulations, the CASMO/SIMULATE/MCNP/FISPACT-II system of codes is in operation at LRT/Nuclear Energy and Safety (NES) Research Division, which is based on the use of validated CASMO/SIMULATE cycle-specific core-follow models of Swiss LWRs. Coupling with FISPACT-II provides the capability for detailed isotopic inventory tracking under irradiation, assessment of materials activation and dpa values, etc. The use of the seamless calculation scheme with translation of the core-follow simulation results into the detailed neutron source specifications for consequent Monte Carlo simulations, accomplished with nuclear data uncertainty propagation capabilities and integrated with appropriate dosimetry validation database, makes the PSI methodology well aligned with the generic best estimate plus uncertainty (BEPU) approach principles. For specific illustrations, this paper presents some results on evaluation of: 1) a few well-known OECD/NEA reactor shielding experimental benchmarks (SINBAD benchmarks H.B. Robinson-2, ASPIS-PCA REPLICA), 2) some dosimetry data obtained from a Swiss PWR and 3) simulation of the dosimetry measurements for the 'PETALE' experimental program at the EPFL research reactor 'CROCUS', as they were foreseen at the time of the experimental planning. Finally, ways to further enhance the simulation methodology and models are discussed. [ABSTRACT FROM AUTHOR]

Reactor pressure vessel (RPV) samples of units 1 and 4 of the Greifswald NPP were investigated with focus on retrospective dosimetry. Specific activities of long-lived radionuclides 63Ni, 93mNb, 94Nb and 99Tc as well as the concentrations of the producing elements were measured. Investigated samples comprise base metal, welding metal and cladding of the RPV. Neutron fluences obtained with the M onte-Carlo codes TRAM O and M CNP were used to calculate specific activities. The gamma-emitter 94Nb appears as a promising candidate due to the large Nb concentration in the cladding of the VVER RPVs. For 93mNb, a very good agreement of measured und calculated activities was found for the RPV cladding samples where 93mNb is primarily produced by the threshold reaction 93Nb(n,n')93mNb. A strong scatter of the ratios of calculated and experimental activities is observed for 63Ni which is primarily produced by slow neutrons. For 99Tc, a good agreement of calculated and measured activities was found for the majority of the base metal and welding metal samples but not for the cladding samples. Updated reaction cross section data for 62Ni(n,γ) and 92M o(n,γ) lead to a better agreement of calculated and measured activities for 63Ni and 93mNb. [ABSTRACT FROM AUTHOR]

This paper presents a method dedicated to the calculation of DPA (displacements per atom) in a stainless steel element irradiated inside Chooz-A reactor core. This method is based on experimental results obtained in samples taken from the analyzed stainless steel element, and calculation results. Experimental results consist in the mass of 59Co and the activity of 60Co, both measured in 2018, i.e. 27 years after the final shutdown of Chooz-A reactor. Calculations are carried out with the TRIPOLI-4® 3D Monte Carlo particle transport code, and the DARWIN/PEPIN2 depletion code, with a simplified modelling of Chooz-A reactor. As explained in this paper, these simplifications are corrected by the use of measurement results. Due to a lack of information about the irradiated steel element (initial 59Co quantity in particular), an iterative process was developed and used. It allows to evaluate missing information and to calculate corrected DPA results in samples. Besides, stress tests were carried out to check the robustness of the process and the uniqueness of DPA result for each sample. This work was carried out by CEA/SERMA with the financial support of EDF. [ABSTRACT FROM AUTHOR]

On 15 April 2023, the last German NPPs were shut down. The final shutdown is followed by a post-operational phase in which measures can be carried out to prepare for the NPPs dismantling and decommissioning. One of the tasks in preparation for the dismantling is to acquire accurate knowledge of the activation distribution within the reactor components. The activation levels depend on the neutron fluence within the reactor. Therefore, to achieve this task, a 3D detailed Monte Carlo model of a German PWR has been developed to calculate the neutron fluence characteristics (spectrum and distribution) within the components of such nuclear reactors type. The reliability of the calculations was validated via metal foil-activation measurements carried out in two active PWRs. This paper presents the Monte Carlo model and describes the calculations and measurement procedures. [ABSTRACT FROM AUTHOR]

This paper presents a quantitative approach to characterize the effectiveness of neutron fluence monitoring programs using adjoint neutron transport calculations. The results of adjoint calculations are useful for this purpose because the adjoint solution is the expected contribution from one region of phase space to a response of interest, i.e., the importance of that region of phase space to the response. Thus, by identifying a response of interest (for this work, the fast (E > 1.0 MeV) neutron fluence rate at the peak fluence rate location in the reactor pressure vessel (RPV)), one can identify the fuel regions most important to that response. Once identified, the ability of various surveillance programs to adequately monitor the most-important fuel region(s) can be assessed. The approach described is applied to a three-loop Westinghouse pressurized water reactor to compare the effectiveness of surveillance-capsule neutron dosimetry versus ex-vessel neutron dosimetry to monitor the fuel assemblies that contribute to the peak fast (E > 1.0 MeV) neutron fluence rate in the RPV. [ABSTRACT FROM AUTHOR]

The special methodology for the determination of the neutron fluences onto the surveillance specimens of the WWER-1000 reactor pressure vessel metal has developed by INR NASU specialists and is successfully applied. The methodology is based on the Monte-Carlo code that is used for neutron transport calculations to the locations of the surveillance specimens. The methodology improvement is described. The fundamentals of the technique of the calculation-experimental determination of the fast neutron fluences onto surveillance specimens and their errors are presented. It is shown a dosimetry experiment in a particular reactor is needed to obtain experimental data about irradiation conditions in the locations of the surveillance specimens. [ABSTRACT FROM AUTHOR]

It is necessary to verify (and validate when experimental data are available) shielding codes and methodologies for Molten Salt Reactors (MSRs). This paper examines the impact of using a fine and coarse multigroup cross section library vs. the reference continuous energy data. Molten Salt Research Reactor (MSRR) being developed by the NEXT Research Alliance (NEXTRA) is used as a testbed. Shielding analyses are performed in support of MSRR development using the MAVRIC sequence of the SCALE 6.2.4 code package. The sequence uses deterministic forward and adjoint radiation transport calculations with FW-CADIS methodology to generate variance reduction parameters that are used by MONACO, a fixed source Monte Carlo code. Originally, MONACO was developed as a multigroup code, and a 200-group and 27-group neutron shielding-focused libraries were provided as part of the package. While the 200-group library provides adequate accuracy for many applications, it is highly desirable to confirm this for any specific analysis. In SCALE 6.2, continuous energy capability was implemented in MONACO, thus enabling direct and consistent evaluation and quantification of the impact of using a multigroup rather than the reference continuous energy library. This paper examines the impact of multigroup approximation on main radiation environment parameters in MSRR shielding analysis, including the fast flux and dpa. [ABSTRACT FROM AUTHOR]

The fast neutron fluence and the activation monitor activities are routinely calculated with TORT deterministic code and BUGLE-B7 nuclear data library with 47 broad energy groups. The objective of the paper is to analyse options to improve reactor dosimetry transport calculations. There are two paths to improve reactor dosimetry calculations. Increasing geometry, angular and energy mesh size is applicable for TORT code while using newer nuclear data libraries is relevant for both deterministic and Monte Carlo codes. Two new calculation options (improved TORT and Monte Carlo MCNP6) were compared with the standard TORT calculation for VVER-440 Dukovany Unit 3 Cycle 31. The fast neutron fluence with 0.5 MeV threshold as well as activity of Fe. Ni, Ti, Cu. Mn and Nb monitors were evaluated. Standard TORT calculations were improved from S16P3 to S30P3 with three times finer axial mesh size. 120° core symmetry r-ϑ mesh size with 0.5° step and fine multigroup libraries VUAMIN-B7 with 199 neutron energy groups and ENDF/B-VH.1 with 200 neutron energy groups. Both ENDFB and IRDFF activation cross sections were used. The drawback of expanded mesh size is raised calculation runtime since TORT deterministic code is not parallelized and one calculation can require multiple weeks of CPU time. An alternative option of using MCNP6 Monte Carlo code with continuous ENDF/B-VH.1 nuclear data with detailed 3-D geometry and pin-wise effective neutron source prepared by MOBY-DICK diffusion code reactor analysis wras explored. It was found that using finer mesh size affects reactor dosimetry' tallies less than the choice of nuclear data library. BUGLE-B7 and VTTAMIN-B7 produce results typically within 1% difference. ENDF/B-VH.1 calculations with 200 neutron energy' groups with TORT code are even in better agreement with MCNP6 calculations with continuous nuclear data libraries. The largest differences of around 2% were observed between VTTAMIN-B7 library based onENDFB-VH.O nuclear data and ENDF/B-VH.1 library. Nuclear data library' has larger impact on the results with up to 7 % difference between all 0.5 MeV fast neutron fluence calculations. The largest intact of nuclear data was observed for Mn(n.2n) monitor. [ABSTRACT FROM AUTHOR]

The Advanced Gas-cooled Reactors (AGR) operating in the UK are, like the Magnox reactors that preceded them, graphite moderated and cooled by pressurised CO2 gas. As such, the nuclear data for carbon are of primary importance in neutron dosimetry and nuclear heating calculations. There is evidence from benchmark data and plant measurements of underprediction of neutron fluxes at the lower end of the fast neutron range for thicknesses of graphite greater than ~50 cm. Furthermore, the range of validation is limited to ~70 cm of graphite, with components beyond this being allocated bounding damage rates. Following the previous ISRD, opportunities were taken to: Look again at the carbon scattering cross-sections by incorporating the ENDF/B-VIII.0 Beta 4 carbon data to assess the impact on the analysis of benchmark data. Reassess validation data from the earlier Magnox reactors to put them on a more appropriate equivalent graphite thickness scale to the AGRs. The investigation into graphite nuclear data libraries concluded that the incorporation of ENDF/B-VIII Beta 4 carbon data into the JEF2.2 and ENDF/B-VII.1 data libraries with BINGO collision processing had only a limited effect on the calculated responses. The UKNDL data, with DICE collision processing, currently used in dosimetry calculations for EDF Energy's AGRs produces the closest agreement with measurement for all calculated responses for graphite thicknesses greater than 10 cm. For reactor plant validation, applying the sensitivity of the 93mNb(n,n') response to graphite density resulted in a 15 cm increase to the validation database when compared to a simple "straight-line" distance. However, caution is advised in applying the benchmark equivalent distance derived from locations within a graphite column, to ex-core plant measurements. [ABSTRACT FROM AUTHOR]

TRAMO is a Monte Carlo code which is specially designed for geometries of nuclear power plants. In recent years, TRAMO has undergone a number of technical developments. In addition to the conversion to modern Fortran, standardized models for VVER-440 and VVER-1000 for the evaluation of monitor experiments on the surface of the reactor pressure vessel were developed, and the corresponding pre- and post-processing was automated. The new TRAMO version is continuously validated and verified. Here we present comparisons with experimental data from unit 1 of the Greifswald nuclear power plant and with MCNP calculations. Earlier calculations with TRAMO showed good agreements in the middle region of the reactor core, but significant deviations outside. A possible explanation is the wrong documentation of the experimental data – especially the heights of the monitors. Assuming a downward shift of the monitor positions by about 24cm, experimental and calculated data show very good agreement. [ABSTRACT FROM AUTHOR]

MCNP code provides numerous variance reduction techniques that are mainly used to optimize the analog Monte Carlo simulations. These techniques can reduce the statistical uncertainty of the simulation results while minimizing computational costs. Weight window is one such technique that is commonly used to optimize shielding calculations. However, for complex shielding problems such as deep penetration, the existing MCNP weight window generator is not efficient enough to generate the appropriate weight parameters for the weight window in a single run. To address this issue, an automated algorithm was developed using the code TRAWEI, which is based on recursive Monte Carlo methodology. The TRAWEI code can generate suitable weight values in a single run with less calculation time. This paper focuses on the verification of TRAWEI-generated weights using a simplified cask model. Further, the results of MCNP simulation using TRAWEI-generated weights are compared with those using weights generated by an existing MCNP weight window generator. [ABSTRACT FROM AUTHOR]

For decommissioning purposes, stacks of foils are used in order to determine the neutron fluence rates in reactor buildings. Due to the relatively great thickness of these foils, self-shielding, as well as mutual shielding become significant for these stacks. In this paper, the results of Monte Carlo calculations are presented for the self-shielding of single foils, and also for the mutual shielding of foils in a stack. Because the irradiation geometry in a reactor building at a large distance from the reactor core is probably not isotropic, especially not so for neutrons of higher energy, the shielding effects were calculated for both isotropic and mono-directional irradiation geometries. The results obtained show a number of general trends which can be understood with a few rules of thumb. [ABSTRACT FROM AUTHOR]

Hybrid halide perovskites have demonstrated significant efficiency in detecting a broad spectrum of high-energy radiation, including X-rays, gamma rays (γ-rays), and neutrons. Given the common occurrence of mixed radiation fields, we investigated the performance of a perovskite-based detector in a neutron-gamma mixed field. A large methylammonium lead tribromide (MAPbBr3) single crystal (SC) was synthesized via the oriented crystal-crystal intergrowth method. This SC was used to fabricate a gamma detector with carbon electrodes, which was tested in the CROCUS zero-power reactor cavity. The detector's photocurrent response exhibited a strong correlation with known gamma dose rates, as measured by an ambient Berthold LB 112 gamma probe, facilitating the accurate conversion of photocurrent to dose rate. Notably, the device did not exhibit degradation under neutron radiation exposure. To further assess the impact of neutrons, X-ray diffraction and electron paramagnetic resonance analyses were performed on small MAPbBr3 SCs grown by inverse temperature crystallization. These SCs were irradiated within the CROCUS reactor core and by a Pu-Be neutron source at liquid nitrogen temperature. Our findings indicate that the perovskite material can withstand the nominal in-core operation conditions of the CROCUS reactor. Additionally, it endures irradiation at liquid nitrogen temperature, corresponding to a fast neutron fluence of approximately 1010 cm-2 and a gamma radiation dose of about 50 Gy, confirming only the temporary creation of defects. No signs of long-term deterioration were observed, suggesting a potential self-healing mechanism. This resilience positions perovskite SCs as viable candidates for in-core radiation detection, supporting the further development of miniaturized MAPbBr3 SC devices for such applications. [ABSTRACT FROM AUTHOR]

The epithermal to fast neutron domain lacks reactions to enhance the knowledge of spectrum in this energy range (1 keV to 1 MeV). Previous studies on this topic focused on 94Zr(n, γ) capture and 117Sn(n, n′) inelastic reactions. In the framework of a research collaboration between the French Atomic and Alternative Energies Commission (CEA) and the Jožef Stefan Institute (JSI), a new project consists in finding new inelastic reactions suitable for dosimetry. This paper describes the selection process of new reactions to be tested in an experimental campaign at JSI. [ABSTRACT FROM AUTHOR]

The assessment of the neutron flux distribution in nuclear power plant components, like reactor internals and the pressure vessel, is one of the most important parts of their residual lifetime evaluation process because the irradiation damage of these components is strongly dependent on it. The fast neutron fluences in power reactors are generally determined using calculations and verified by measurements. Discrepancies between them sometimes occur, which can be caused by be the inaccuracies in the power distribution used in neutron transport calculations. This paper quantifies the effect of uncertainty in power density on the neutron fluences behind the reactor pressure vessel (RPV). An increase in power density was experimentally simulated by implementing a higher uranium enrichment of certain selected pins. The experiment was also simulated using the MCNP code with the ENDF/B-VII.1 library. Both the experimental and calculational data show a significant local increase in neutron flux. Even behind the RPV, the increase is as high as 25%, while the local power increase in the perturbed pins is about 70–80%, which is approximately equal to the 2 σ power density uncertainties. A good agreement between measurement and calculation was found. [ABSTRACT FROM AUTHOR]

The paper builds on the results so far within the research project focused on conceptual design of an innovative safety system for a Gas-Cooled Nuclear Reactors. The paper focuses on a description of the design methodology of the gas turbine flow path, which is part of the secondary circuit of the proposed safety system. The aim is to propose the design of the last stage turbine blading and then to extract a special bucket blade profile that will match the expected flow direction during deep off-design modes. The extracted blade profile will then be subsequently assembled into a blade row which will be tested in the suction-type high-speed wind tunnel in the Aerodynamic Laboratory of the Institute of Thermomechanics Czech Academy of Science in Nový Knín. [ABSTRACT FROM AUTHOR]

Small-Angle Neutron Scattering (SANS) is a ubiquitous technique at neutron scattering facilities throughout the world, and they are particularly attractive instruments to consider for a CANS facility as they are relatively simple in design, do not require high resolution, and can facilitate research across a wide swath of science. On the other hand, CANS facilities employ a fundamentally different primary neutron source than earlier generations of neutron facilities, and such CANS facilities must often be built within a different set of constraints than is typically seen at a major reactor or spallation source. We discuss briefly how these considerations may lead to minor variations in the design optimization for a SANS instrument at a CANS facility from the typical descriptions available from other facilities using the design of the instrument at LENS as an example. [ABSTRACT FROM AUTHOR]