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Abstract: Under the ITER/ITA task, we have conducted the neutron streaming experiment simulating narrow and deep gaps at boundaries between ITER vacuum vessel and equatorial port plugs. Micro-fission chambers and some activation foils were used to measure fission rates and reaction rates to evaluate the relative fast and slow neutron fluences along the gap in the experimental assembly. The MCNP4C, TORT and Attila codes were used for the experimental analysis. From comparing our measurements and calculations, the following facts were found: (1) in case of a such narrow and deep gap structure, the calculation with MCNP, TORT and Attila codes and FENDL-2.1 is sufficient to predict fast neutron field inside the gap; (2) by scattering neutrons in the experimental room, experimental error considerably increased at the deeper region than 100cm; (3) angular quadrature set of upward biased U315 and last collided source calculation on TORT and Attila were very important technique for accurate estimation of neutron transport. [Copyright &y& Elsevier]

Abstract: A series of breeding blanket neutronics experiments were carried out with DT neutrons in the Japanese ITER test blanket module at JAEA Fusion Neutronics Source (FNS), and tritium production rates (TPRs) inside different blanket mock-ups were measured in detail. Most of calculations made using FENDL-2.1 and MCNP agreed well with the measured TPRs. On the other hand, there were overestimations of the calculation for the TPR in the experiment with a reflector and the TPR around the boundary between the rear part of the breeder layer and the beryllium in the pebble bed mock-up experiment. In this study, we investigated the possibility of improvement of the calculation by modifying the part of angular distributions in backward directions in FENDL-2.1. By uniformly decreasing the backward elastic scattering cross-sections of 56Fe by 20% below the original data for incoming neutrons in the energy range from 0.5234MeV up to 15.0MeV, the overestimation of the calculations for the TPR in the enriched breeding blanket experiment with a reflector was reduced. Also, by uniformly decreasing the backward elastic scattering cross-sections of 9Be by 20% and 30% below the original data for the incoming neutrons in the energy range from 0.93MeV up to 14.94MeV, and by 20% for those from 0.62MeV up to 14.94MeV, the overestimation of the calculations for the TPR around the boundary between the rear part of the breeder layer and the beryllium was lessened. [Copyright &y& Elsevier]

Abstract: There are very thin and long slits at boundaries between vacuum vessel and port plugs in ITER. The slits are 0.5–2cm in width and their length amount about 200cm in depth. Therefore, it is a key issue to evaluate neutron-shielding effects of assembly with such thin and long slits. In order to verify them, our group has started to conduct the D-T neutronics experiment with an experimental iron assembly at the Fusion Neutronics Source (FNS) in Japan Atomic Energy Agency (JAEA) under the ITER/ITA Task 73–10. Fission rates among a slit in the assembly were measured with U-238 and U-235 micro-fission chambers. From our experiment, the following facts were found: (1) the C/E value of U-238 fission rate is within 0.9–1.1 at the most measured points and (2) the C/Es of the U-235 fission rate were almost closed to the unity within 0.9–1.2. However, they tended to decrease for the depth slightly. [Copyright &y& Elsevier]

Abstract: In order to investigate the activation performance of prospective materials of the IFMIF facility, we have carried out irradiation experiments with the AVF cyclotron in the TIARA facility of the Japan Atomic Energy Research Institute (JAERI). We measured the production cross sections of the nuclides Cu-61, Cu-64, Zn-62 and Zn-63 by deuterons irradiation of copper with natural isotope abundance (natCu) in the energy range of 22–40MeV. The obtained cross sections were compared with data from the ACSELAM library and with previously measured cross section data. The comparisons revealed that some of the ACSELAM cross section data are different from the experimental data and have to be improved. [Copyright &y& Elsevier]

Abstract: Two nuclear microprobe techniques; deuteron-induced nuclear reaction analysis (d-NRA) and neutron elastic recoil detection analysis (NERDA), are presented to obtain hydrogen isotope depth profiles of fusion reactor materials in a shallow and deep region, respectively. Tritium and deuterium depth profiles in a bumper limiter of TFTR used in the D–T experiments were measured as an application of d-NRA. The T(d, α)n and D(d, t)H reactions were used to determine the depth profiles. The tritium concentration had a peak, whereas deuterium showed a broad distribution in the profile. Also, the application of NERDA is demonstrated. A proof-of-principle experiment was performed using a standard sample of deuterated polyethylene. The depth profile and resolution for NERDA are discussed. [Copyright &y& Elsevier]

Abstract: Experiments have been conducted with breeding blanket mock-ups consisting of layers of candidate materials for JAERI’s pebble bed blanket design: lithium titanate, beryllium and reduced activation ferritic steel F82H. The mock-ups were irradiated with 14 MeV neutrons from the DT neutron source at FNS. The lithium titanate was enriched to 40 and 95% in . The local tritium production rate in the 12 mm thick breeding layers was measured by means of pellet detectors of lithium titanate, Li2CO3 and liquid scintillation counting techniques. Additional verification of the neutron spectrum was obtained from material activation such as activation foils. The results were analyzed with the Monte Carlo neutron transport code MCNP-4C and several transport cross section libraries such as FENDL/MC-2.0. The calculation/experiment ratio was found to be within the combined error estimate of calculation and experiment but in some cases, the calculations showed an overestimation tendency. [Copyright &y& Elsevier]

Abstract: For the evaluation of the damages, nuclear heating, and also tritium inventory in the fusion materials, the precise information of the energetic charged particles produced in the materials by nuclear reactions is important. Fusion neutronics source in Japan Atomic Engineering research Association, JAERI/FNS, has started the measurements of the double differential cross-sections (DDXs) and the angular differential cross-sections (ADXs) for the candidate materials of fusion reactors using the collimated D–T neutron beam system and the precise results of the DDXs and ADXs for the 9Be(n,αx) and 9Be(n,t) reactions have been obtained. Also, from the comparison of the evaluated nuclear data for the beryllium, it was found that the calculated DDX of 9Be(n,α) above about 3MeV was underestimated and the total 9Be(n,t) cross-section obtained with the measurement of the DDX and ADX was in good agreement with the evaluated nuclear data. In the following sections, the details of the measurement methods for the charged particles and the obtained DDX and ADX for the beryllium are described. [Copyright &y& Elsevier]

At Fusion Neutronics Source (FNS) of JAERI, tritium breeding experiments with blanket mock-ups consisting of advanced fusion reactor materials are in progress. The breeding zones are thin layers of lithium titanate which is one of the candidate tritium breeder materials for the DEMO fusion power reactor. It is anticipated that the application of small pellet-shaped lithium titanate detectors manufactured from the same material as the breeding layer will reduce experimental uncertainties arising from necessary corrections due to different isotopic lithium volume concentrations in breeding material and detector. Therefore, a method was developed to measure the local tritium production by means of lithium titanate pellet detectors and a liquid scintillation counting technique. The lithium titanate pellets were dissolved in concentrated hydrochloric acid solution and the resulting acidic solution was neutralized. Two ways of further processing were followed: direct incorporation into a liquid scintillation cocktail and distillation of the solution followed by mixing with liquid scintillator. Two types of lithium titanate pellets were investigated with different 6Li enrichment and manufacturing procedure. It was found that lithium titanate is suitable for tritium production measurements. However some discrepancies in the measurement accuracy remained with one of the investigated pellet detectors when compared with a well-established lithium carbonate measurement technique and this issue needs further investigation. [Copyright &y& Elsevier]

A shutdown dose experiment was conducted by using the FNS at JAERI under the ITER R&D Task T-426, in order to evaluate the accuracy of the shutdown dose evaluation method. The activation and the delayed gamma-ray distribution after passing through the straight duct were simulated in the experiment. The experiment was conducted at the 80° beam line of the FNS and the irradiation was carried out over 6 days. The experimental analyses were carried out by using the new direct one step Monte Carlo method that is used as the design tool to estimate shutdown dose of the ITER/FEAT. The nuclear data packages, FENDL-2/MC and FENDL-2/A, were compared through the analysis of the experiment. As a result, the following facts were found for the accuracy of the shutdown dose evaluation: (1) the C/E values of the shutdown dose at around 106 s after the irradiation were 0.92–1.07; (2) the shutdown dose rate could be evaluated by the new direct one-step Monte Carlo method within the experimental error, and the accuracy of the method is sufficient for shielding design of the International Thermonuclear Experimental Reactor (ITER). [Copyright &y& Elsevier]

Abstract: Neutron benchmark experiments carried out at 14MeV neutron generators on ITER relevant materials and components have provided validation of nuclear data used in shielding and activation calculations for ITER, as well as an assessment of the related uncertainties. Further validation activity is still needed which could be performed in ITER, especially for activation cross sections, for dose rate calculations and for tritium production rates. In particular, concerning the nuclear performance of breeder blankets to be tested in ITER, the paper discusses the extensive preparatory work, to be carried out in a coordinated way among the participating Parties, required to finalize the design of neutronics Test Blanket Modules (TBMs), so that these tests can provide a complete and comparable, therefore useful, picture of the different concepts. Neutronics experiments being carried out at neutron generators on TBM mock-ups are providing important results on the quality of relevant neutron cross sections and for the preparation of the measurement techniques. [Copyright &y& Elsevier]

Abstract: Double-differential cross-section (DDX) for emitted charged particles is necessary to estimate material damage, gas production and nuclear heating in a fusion reactor. Detailed measurements of the cross-sections for beryllium, carbon and fluorine, which are among the composition materials of expected fusion blankets and first walls, were carried out with a charged-particle spectrometer using a pencil-beam DT neutron source. As verification of the cross-sections evaluated in three nuclear libraries (JENDL-3.3, ENDF/B-VI and JEFF-3.1), our measured data were compared with the data evaluated in the libraries. From the comparison, the following problems were pointed out: Beryllium: Remarkable differences in energy and angular distribution for α-particles were observed between the measured data and the libraries. The estimated total cross-section for α-particle production well agreed with the libraries. Carbon: There was a discrepancy of about 20% between JENDL-3.3 and ENDF/B-VI (JEFF-3.1) for α-particle production cross-section, and no DDX for α-particles is given in the libraries. Our obtained total cross-section for α-particle production was rather consistent with ENDF/B-VI (JEFF-3.1), and the value evaluated in JENDL-3.3 seemed too large. Fluorine: The remarkable differences for DDX of protons and α-particles were observed between the obtained result and JENDL-3.3, although detailed DDX was stored only in JENDL. The obtained total cross-sections mostly supported the evaluation of ENDF/B-VI (JEFF-3.1). [Copyright &y& Elsevier]

Abstract: Sequential charged particle reactions, that is, reactions induced by secondary charged particles generated due to neutron reactions, may cause non-negligible radioactivity in low activation material. A computer code ACT4, which had been developed to calculate radioactivity, decay heat and gamma ray source for fusion reactors was supplemented an ability to deal with the reactions. Experiments conducted with Fusion Neutronics Source in Japan Atomic Energy Research Institute for the radioactivity by sequential charged particle reactions were analyzed with the newly revised code of ACT4. As the result, the activations via sequential reaction in LiF showed reasonable agreement with experimental data, which means that the ACT4 program will work well. However, the other activations in vanadium and low activation ferritic steel did not. The disagreement must be caused by incorrectness in neutron cross sections, charged particle cross sections, charged particle emission spectra, etc., and they need to be fully reviewed for the detailed and reliable design analysis of fusion reactors. [Copyright &y& Elsevier]

Abstract: The neutronics study in support of the DEMO blanket concept has been performed using the DT neutrons at the FNS facility in JAERI on blanket mockups and a bulk beryllium assembly. From the results obtained in the blanket mockups, it has been clarified that the C/Es of the integrated tritium production are 1.01–1.04 and 1.11–1.13 in the experiments without and with the neutron reflector, respectively. The tritium production can be very accurately predicted in the experiment without the reflector. In the mockups with the 12.6 and 25.2mm thick tungsten armors, the integrated tritium productions are reduced by 3 and 6% relative to the case without the armor, respectively. Based on analysis of the integral experiment on the beryllium assembly, it is possible to conclude that the calculation uncertainty of the reaction rates in beryllium provides an acceptable agreement with the experimental data related to tritium breeding. Tritium techniques used for TPR measurements have been tested in the international benchmark experiment. There is a very good agreement within the estimated uncertainties between results obtained at JAERI and at TUD, though the results obtained at ENEA show an underestimation (about 15–20%) in comparison with FNS and TUD results. [Copyright &y& Elsevier]

Abstract: An integral breeding blanket experiment has been done at Fusion Neutronics Source of JAERI with an assembly of three layers of lithium titanate, beryllium, and low activation ferritic steel F82H. The experimental assembly and the neutron source were enclosed in a SS316 stainless steel can and irradiated with D–T neutrons. The tritium production rate was measured with small Li2CO3 pellet detectors. The tritium production from7Li was estimated by means of the reactions32S(n, p)32P and35Cl(n, )32P which have effective cross-sections similar to7Li(n, n)T but are more sensitive methods. The neutron flux in the assembly was checked with activation foils. The measurement results were analyzed with the three-dimensional Monte Carlo code MCNP-4C and three different beryllium cross-section evaluations taken from FENDL/MC-2.0, FENDL/E-2.0, and EFF-3.03. The library FENDL/MC-2.0 was also used for the other materials in the assembly. All three beryllium libraries model the fast flux in the assembly well. Differences were seen for indium foils. An overestimation was observed for the tritium production. [Copyright &y& Elsevier]

The D–T neutron skyshine experiments have been carried out at the Fusion Neutronics Source (FNS) of JAERI with the neutron yield of ∼1.7×1011 n/s. The concrete thickness of the roof and the wall of a FNS target room are 1.15 and 2 m, respectively. The FNS skyshine port with a size of 0.9×0.9 m2 was open during the experimental period. The radiation dose rate outside the target room was measured a maximum distance of 550 m from the D–T target point with a spherical rem-counter. Secondary gamma-rays were measured with high purity Ge detectors and NaI scintillation counters. The highest neutron dose was about 9×10−22 Sv/(source neutron) at a distance of 30 m from the D–T target point and the dose rate was attenuated to 4×10−24 Sv/(source neutron) at a distance of 550 m. The measured neutron dose distribution was analyzed with Monte Carlo code MCNP-4B and a simple line source model. The MCNP calculation overestimates the neutron dose in the distance range larger than 230 m. The line source model agrees well with the experimental results within the distance of 350 m. [Copyright &y& Elsevier]

A key conception of the remote maintenance (RM) method is presented of the target assembly (TA) of Advanced Fusion Neutron Source (A-FNS), that is an accelerator-driven fusion neutron source utilizing the Li(d,xn) nuclear reactions. The RM method, newly proposed here, is aimed at (i) enhancing the modularity of the TA RM and (ii) keeping the compatibility with the materials irradiation tests, compared to the preceding TA RM design of International Fusion Material Irradiation Facility (IFMIF). A 3-dimensional kinematics simulation indicates that the large clearance of 52 mm between the TA and high flux test module (HFTM) brings the independent removal from and the installation in the test cell, regardless of the presence of the test modules in the test cell. A neutronics performance of the HFTM with the enlarged clearance is also studied. The gradient of the calculated neutron irradiation damage is found to be decreased and satisfies a users' requirement of the irradiation environment when the clearance is enlarged from 2 to 52 mm. (The former value was the IFMIF engineering design value.) These simulation results suggest that the proposed TA RM method will enhance not only the modularity of the TA, but also will improve the materials irradiation test environment. [ABSTRACT FROM AUTHOR]

• Alleviation of impurities levels in Li for IFMIF-type neutron source are suggested. • No fatal problems are found by applying the alleviation of impurities levels. • Level of Be-7 in Li under process can be diminished by the alleviation. The alleviated limits of impurities of nitrogen and hydrogen isotopes in lithium are suggested to be 400wppm and 550appm, respectively for early realization of Advanced Fusion Neutron Source planned in Japan. The applicability of the suggestion is discussed by comparing to that of IFMIF. As the results, not less than 4-year buffer for R&D of nitrogen trap are brought in, which is thought as one of bottle neck. Additionally, the application of this suggestion would diminish the beryllium (a radioactive isotope will generate in A-FNS process) concentration in processing lithium, as well. On the other hand, it is indicated that there would be no fatal problems by application of the present suggestion. [ABSTRACT FROM AUTHOR]

Abstract: A neutronics experiment has been performed in the frame of European Fusion Technology Program on a mock-up of the EU Test Blanket Module (TBM), HCPB concept, with the objective to validate the capability of nuclear data to predict nuclear responses, such as the tritium production rate (TPR), with qualified uncertainties. In the experiment, the TPR has been measured using Li2CO3 pellets located at various depths at two symmetrical positions at each depth, one in the upper and one in the lower breeder cassette. Three independent measurements were performed by ENEA, TUD/VKTA and JAEA. The neutron flux in the beryllium layer was measured as well using activation foils. The measured tritium production in the TBM (E) was compared with the same quantity (C) calculated by the MCNP-4C code using a very detailed model of the experimental set up. A sensitivity and uncertainty analysis has also been performed to evaluate the calculation uncertainty due to the uncertainty on neutron cross sections. The results of such analysis, carried out using both Monte Carlo and deterministic codes, shows that the uncertainty due to cross sections amounts to 4–5% (2σ). C/E ratios were obtained with a total uncertainty on the C/E comparison of 9–10% (2σ). The C/E comparison shows that the calculation of the tritium production is underestimated by about 10% in average and is within the combined 2σ uncertainty margin of the measurements and the calculations. The obtained results indicate that design calculations of the tritium breeding ratio provide conservative results for the fusion reactor employing a HCPB type breeder blanket. [Copyright &y& Elsevier]

Abstract: Double-differential charged-particle emission cross section (DDXc) is a fundamental value to estimate nuclear heating and material damages in fusion reactor development. In the present study, measurements of DDXc for several light elements have been carried out with a newly developed spectrometer. In the spectrometer, a good signal-to-noise ratio and very low minimum measurable energy were realized by using a pencil-beam DT neutron source of FNS in JAERI and a counter-telescope system with very thin silicon surface barrier detectors of appropriate thicknesses. The detailed DDXc data in a wide emission energy range were obtained for 27Al, 9Be, natC and 19F. For 27Al(n,xα) reactions, the measured total cross section agreed well with the evaluated data. For 9Be(n,xα) reactions, there were differences in the α-particle emission DDX between measurement and the evaluated data in the detail of energy structure. For natC(n,xα) reaction, a large discrepancy between the present result and the previous experimental data appeared. For 19F(n, charged-particle) reactions, some peaks corresponding to excited states of the residual nuclei were observed in the energy spectrum of emitted proton, deuteron, and triton particles. [Copyright &y& Elsevier]

Abstract: A neutronics experiment has been launched in the European fusion technology program on a mock-up of a test blanket module with the objective to validate the capability of nuclear data to predict nuclear responses with qualified uncertainties. In the experiment, the mock-up will be irradiated with a 14-MeV neutron source and the tritium production rate will be measured and compared with the calculated value by neutron cross sections from the European Fusion File. Given the difficulty of performing a reliable measurement of tritium production rate in such experiments, an international benchmark comparison on tritium measurement techniques has been carried out in collaboration between ENEA, TUD/VKTA and also JAERI that has already performed such experiments. The main purpose of the comparison is to establish the reference technique of TPR measurements to be used later in the TBM experiment. Three steps of the comparison procedure are presented in the paper as well as the preliminary results. The implications for the TBM experiment are also discussed. [Copyright &y& Elsevier]

Abstract: Neutron calculations have been carried out for a fusion–fission (FF) hybrid reactor system, focused on the Japanese situation, to be realized as soon as possible. The ITER facility configuration has been taken as basis for the calculations. In the blanket region reprocessed plutonium and natural uranium oxide were taken as nuclear fuel for neutron/energy multiplication and lithium oxide with beryllium for tritium breeding. The plasma conditions were derived from the JT-60 experience. The neutron analysis has been done with a three-dimensional Monte Carlo code and a point burnup code. As a result, it was found that the proposed FF concept can be operated for at least 5 years with acceptable good performance from the point of view of tritium breeding, sub-criticality, energy multiplication and used for transmutation of long-lived fission products without nuclear fuel exchange and shuffling. [Copyright &y& Elsevier]

Degradation of electrical insulation property of mineral-insulated (MI) cables under fusion neutron irradiation was examined by use of Fusion Neutronics Source at Japan Atomic Energy Research Institute. Observed increase of leakage current in an MI cable (diameter; 4.8 mm, length; 2.0 m) was ∼5.0 pA for the DT neutron flux of 2.9×107 n/cm2/s and bias voltage of 200 V between the central wire and sheath. The bias voltage and neutron flux dependence of the leakage current indicates that the degradation of insulation property of the cable is dominated by the production rate and drift of electric charge in the insulator. Furthermore, the appearance of large transient leakage currents immediately after the start and stop of the irradiation and the growth of leakage current even for the bias voltage of 0 V imply that electrical properties of MI cables under irradiation are considerably affected by the distribution of electric charge trapped in the insulator. [Copyright &y& Elsevier]