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We consider the Cauchy problem for the wave equation with overdamping and semilinear source terms: ( P ) { u t t − Δ u + b ( t ) u t = N ( u ) , ( t , x ) ∈ R + × R N ( u , u t ) ( 0 , x ) = ( u 0 , u 1 ) ( x ) , x ∈ R N , with b ( t ) = b 0 ( t + 1 ) − β , b 0 > 0 , β − 1 and N ( u ) = | u | p − 1 u , p > 1 . Ikeda and Wakasugi [8] have recently showed that, when | N ( u ) | ≤ C | u | p for any p > 1 , there is a global-in-time solution to (P) for suitable small data, and that, when N ( u ) = ± | u | p , the local-in-time solution blows up within a finite time for suitable large data. To show the blow-up result, their method seems to be not applicable to our semilinear term. Our aim is to show the blow-up of solutions for suitable large data, by the method much different from theirs.

A new nuclear data library, JENDL/ImPACT-2018, was developed for an innovative study on the transmutation of long-lived fission products. Nuclear reaction cross-sections were newly evaluate...

The Japan Atomic Energy Agency's accelerator-driven subcritical system (JAEA-ADS) proposes the reduction of nuclear waste through the transmutation of minor actinides. The JAEA-ADS drives a 30-MW proton beam to a spallation target to produce neutrons for a subcritical 800-MWth reactor. The beam must be transported from the end of the linear accelerator (linac) to the target located inside the reactor core with high beam power stability and low peak density to ensure beam window integrity, which is a primary concern for the ADS project. Additionally, the design of the beam transport to the target (BTT) must be compatible with the established reactor design, and the elements that comprise the BTT must facilitate the maintenance and replacement of the fuel and the beam window. To this end, a robust-compact BTT design was developed through massive multiparticle simulations. First, the beam optics was optimized to guarantee beam window feasibility requirements by providing a low peak density of less than 0.3 μA/mm2. Then, beam stability was evaluated and improved by a simultaneous application of input beam and element errors. The input beam errors were based on the beam degradation obtained by implementing fast fault compensation in the linac, which is a key strategy to attain high-reliability operation. The results show that the BTT design fulfills reactor and beam window requirements for JAEA-ADS.

Nuclear Material Balance code version 4.0 (NMB4.0) has been developed through collaborative R&D between TokyoTech&JAEA. Conventional nuclear fuel cycle simulation codes mainly analyze actinides and are specialized for front-end mass balance analysis. However, quantitative back-end simulation has recently become necessary for considering R&D strategies and sustainable nuclear energy utilization. Therefore, NMB4.0 was developed to realize the integrated nuclear fuel cycle simulation from front- to back-end. There are three technical features in NMB4.0: 179 nuclides are tracked, more than any other code, throughout the nuclear fuel cycle; the Okamura explicit method is implemented, which contributes to reducing the numerical cost while maintaining the accuracy of depletion calculations on nuclides with a shorter half-life; and flexibility of back-end simulation is achieved. The main objective of this paper is to show the newly developed functions, made for integrated back-end simulation, and verify NMB4.0 through a benchmark study to show the computational performance.

The Japan Atomic Energy Agency (JAEA) has investigated an accelerator-driven system (ADS) to transmute minor actinides which will be partitioned from the high level waste. There are various inherent issues for the research and development on the ADS. The recent two activities to realize a feasible and reliable ADS concept are introduced in this paper. For the feasibility, the design of a beam window which is a boundary of the accelerator and the subcritical core, is one of the most important issues. To mitigate the design condition of the beam window, namely to reduce the necessary proton beam current, the subcritical core concept with subcriticality adjustment rods was investigated. Through the coupled analysis of neutronics, particle transport, thermal-hydraulics and structure, a feasible beam window concept was presented. For the reliability, a beam-trip is the inherent and serious issue for the ADS design because it induces a rapid temperature change in structures in the subcritical core. In the previous study, the beam-trip frequency of the ADS accelerator was estimated based on the operation data of the existing accelerators. To minimize the beam-trip frequencies, a double-accelerator concept was proposed and its beam-trip frequency was estimated. The results indicated that the double-accelerator concept was useful to realize reliable accelerator operation for the ADS although the cost of the accelerators and those buildings would be double.

At the Kyoto University Critical Assembly, a series of reaction rate experiments is conducted on the accelerator-driven system (ADS) with spallation neutrons generated by the combined use of 100 Me...

Sensitivity and uncertainty analyses of lead (Pb) isotope cross sections are conducted with the use of sample reactivity experiments at the Kyoto University Critical Assembly (KUCA). With the combi...

To investigate the applicability of the pulsed neutron source method using a pulsed spallation neutron source for an on-line subcriticality monitoring system of an accelerator-driven system, a subc...

The paper presents the measured cumulative yields of 44 Ti for nat Cr, 56 Fe, nat Ni and 93 Nb samples irradiated by protons at the energy range 0.04–2.6 GeV. The obtained excitation functions are compared with calculations of the well-known codes: ISABEL, Bertini, INCL4.2+ABLA, INCL4.5+ABLA07, PHITS, CASCADE07 and CEM03.02. The predictive power of these codes regarding the studied nuclides is analyzed.

In order to perform the parametric survey for an accelerator-driven system (ADS) core with the subcriticality adjustment mechanism, a new calculation code system, ADS3D, was developed on MARBLE which is a comprehensive and versatile framework for reactor analysis. The application of ADS3D was also demonstrated on the neutronics design of ADS operated by control rod (CR) movement. Through the neutronics calculation, it was shown that the maximum proton beam current was decreased from 20.5 to 11.6 mA due to the switch from beam-operated to CR-operated core.

The impact of different spallation models and parametrisation of nucleon–nucleus interactions in the particle transport code PHITS on the nuclear characteristics of an accelerator-driven system (ADS) is investigated. Cut-off neutrons below 20 MeV calculated using the default option of the current spallation model (i.e. Liege intranuclear cascade (INC) model version 4.6, INCL4.6) are found to be 14% less than those calculated by the old spallation model (i.e. Bertini INC model). This decrease increases the proton beam current that drives the 800-MW thermal power and impacts various ADS parameters, including material damage, nuclear heating of the proton beam window and the inventory of spallation products. To validate these options based on the ADS neutronics design, we conduct benchmark calculations of the total and non-elastic cross sections, thick target neutron yields and activation reaction rate distributions. The results suggest that Pearlstein–Niita systematics, which is a default option of ...

The Japan Atomic Energy Agency (JAEA) has been conducting the research and development (R&D) on accelerator-driven subcritical system (ADS) as a dedicated system for the transmutation of long-lived radioactive nuclides. To foster the R&D of ADS, the Transmutation Physics Experimental Facility (TEF-P) in the J-PARC project has been planned to build by JAEA [1]. The TEF-P is used minor actinide (MA) fuel which has large decay heat, so during the failure of the core cooling system, the evaluation of the core temperature increase is important. This study aims to evaluate the natural cooling characteristics of TEF-P core and to achieve a design that does not damage the core and the fuels during an accident (the failure of the core cooling system). The experiments using mockup device was performed to validate the heat transfer characteristics in the empty rectangular lattice tube. It was obtained that the actual heat transfer coefficient of empty rectangular lattice tube was about 2.2 times larger than the theoretical free convection model. It was also confirmed that the insertion of any block into the empty rectangular lattice tube could achieve the higher heat transfer coefficient. Using the heat transfer coefficient obtained by experiment results, thermal analysis was performed by the three-dimensional heat transfer analysis. As a result, the calculation results showed that the maximum core temperature will be 294 °C which is less than the design criterion of temperature, 327 °C. It was presented that the design condition which the core temperature will be below the design criterion during the failure of the core cooling system through this study.

Sample reactivity experiments on the uncertainty analyses of Pb nuclear data are carried out by substituting Al plates for Pb ones at the Kyoto University Critical Assembly, as part of basic research on Pb–Bi for the coolant. Numerical simulations of sample reactivity experiments are performed with the Monte Carlo calculation code MCNP6.1 together with four nuclear data libraries JENDL-3.3, JENDL-4.0, ENDF/B-VII.0 and JEFF-3.1, to examine the accuracy of cross-section uncertainties of Pb isotopes by comparing measured and calculated sample reactivities. A library update from JENDL-3.3 to JENDL-4.0 is demonstrated by the fact that the difference between Pb isotopes of the two JENDL libraries is dominant in the comparative study, through the experimental analyses of sample reactivity by the MCNP approach. In addition, JENDL-4.0 reveals a slight difference from ENDF/B-VII.0 in all Pb isotopes and 27Al, and from JEFF-3.1 in 238U and 27Al. Based on these results, further experiments are needed to investigate t...

We consider the Cauchy problem of the semilinear wave equation with a damping term \begin{align*} u_{tt} - \Delta u + c(t,x) u_t = |u|^p, \quad (t,x)\in (0,\infty)\times \mathbb{R}^N,\quad u(0,x) = \varepsilon u_0(x), \ u_t(0,x) = \varepsilon u_1(x), \quad x\in \mathbb{R}^N, \end{align*} where $p>1$ and the coefficient of the damping term has the form \begin{align*} c(t,x) = a_0 (1+|x|^2)^{-\alpha/2} (1+t)^{-\beta} \end{align*} with some $a_0 > 0$, $\alpha < 0$, $\beta \in (-1, 1]$. In particular, we mainly consider the cases $ \alpha < 0, \beta =0$ or $\alpha < 0, \beta = 1$, which imply $\alpha + \beta < 1$, namely, the damping is spatially increasing and effective. Our aim is to prove that the critical exponent is given by $ p = 1+ \frac{2}{N-\alpha}$. This shows that the critical exponent is the same as that of the corresponding parabolic equation $c(t,x) v_t - \Delta v = |v|^p$. The global existence part is proved by a weighted energy estimates with an exponential-type weight function and a special case of the Caffarelli-Kohn-Nirenberg inequality. The blow-up part is proved by a test-function method introduced by Ikeda and Sobajima (arXiv:1710.06780v1). We also give an upper estimate of the lifespan., Comment: 28 pages

In this paper, we consider a system of weakly coupled semilinear damped wave equations. We determine the critical exponent for any space dimensions. Our proof of the global existence of solutions for supercritical nonlinearities is based on a weighted energy method, whose multiplier is appropriately modified in the case where one of the exponent of the nonlinear term is less than the so called Fujita’s critical exponent. We also give estimates of the lifespan of solutions from above for subcritical nonlinearities.

After the severe accident at the Fukushima-1 nuclear power plant, large amounts of contaminated stagnant water have accumulated in turbine buildings and their surroundings. This rapid communication...

For the safe storage of zeolite wastes generated by the treatment of radioactive saline water at the Fukushima Daiichi Nuclear Power Station, this study investigated the fundamental properties of h...

A sensitivity and uncertainty analysis was performed for the minor-actinide transmuter proposed by the Japan Atomic Energy Agency with JENDL-4.0. Analysis with sensitivity coefficients and the JENDL-4.0 covariance data showed that the covariances of the capture cross sections and fission-related parameters of MAs and Pu isotopes have considerable impact on the uncertainties of reactor physics parameters, and covariances of the inelastic scattering cross section of lead-bismuth eutectic (LBE) materials significantly affect the uncertainty of coolant-void reactivity.

Utilization of rock-like oxide (ROX) fuel in light water reactors for plutonium (Pu) burning was studied by nuclear material balance (NMB) analysis for a case of Japanese phase-out scenario under investigation after the Fukushima accident. For the analysis, the NMB code was developed with features of accurate burn-up calculation, flexible combination of reactors and fuels, and an ability to estimate waste and repository. Three scenario groups of once-through Pu burning in mixed oxide (MOX) fuel and in ROX fuel were analyzed. Using two full-MOX or full-ROX reactors the Pu amount is reduced to about one-half and the isotopic vector of Pu deteriorated for being used as a nuclear weapon, especially in terms of spontaneous fission neutron generation. Effects of ROX reactors are more significant than MOX reactors in terms of both reduction in the Pu amount and deterioration of the isotopic vector. Repository footprint and potential radiotoxicity are not reduced by the MOX and ROX reactors because the heat and t...

A sensitivity and uncertainty analysis was performed for the accelerator-driven system (ADS) proposed by the Japan Atomic Energy Agency (JAEA) with the latest version of the Japanese Evaluated Nuclear Data Library (JENDL-4.0). Significant discrepancies have been found between the reactor physics parameters of JENDL-4.0 and those of JENDL-3.3. An analysis with the sensitivity coefficients showed that the major contributors to these discrepancies are the differences in the inelastic scattering cross sections of 206Pb and 207Pb, and the capture and inelastic scattering cross sections and ν value of 241Am. The uncertainty analysis with the JENDL-4.0 covariance data found that the covariances of the fission neutron spectrum of minor actinides (MAs) have a considerable impact on the uncertainties of the reactor physics parameters.

The transmutation of Iodine-129 in accelerator driven system (ADS) is studied. The sodium iodide assembly loadings inside the core of ADS and in the surrounding core region are considered. The introduced concept of ADS with a power of 800 MWt is able to transmute 250 kg/y of minor actinides (MAs) and 46 kg/y of Iodine-129 that supports ten PWRs. The initial loading masses of MAs and I-129 in ADS were equal to 3810 kg and 824 kg, respectively.

The transient analyses for the lead–bismuth cooled Accelerator-Driven System (ADS) were performed with the use of the SIMMER-III and RELAP5/mod3.2 codes to investigate the possibility of the core damage. Five accidents; the beam window breakage, the protected loss of heat sink, the beam overpower, the unprotected loss of flow and the unprotected blockage accident were analyzed as the typical accidents in the ADS. Through these calculations, it was confirmed that all calculation results except the protected loss of heat sink satisfied the no-damage criteria. In the protected loss of heat sink, the cladding tube temperature reached at the melting temperature after 20 h although the calculation condition was very conservative. It is required to design a safety system of the ADS to decrease the frequencies of the accidents and to ease the accidents.

The design of Accelerator Driven Systems (ADS) requires the development of simulation tools that are able to describe in a realistic way their nuclear performance and transmutation rate capability. In this publication, we present an evaluation of state of the art Monte Carlo design tools to assess their performance concerning transmutation of long-lived fission products. This work, performed under the umbrella of the International Atomic Energy Agency, analyses two important aspects for transmutation systems: moderation on Lead and neutron captures of 99 Tc, 127 I and 129 I. The analysis of the results shows how shielding effects due to the resonances at epithermal energies of these nuclides affects strongly their transmutation rate. The results suggest that some research effort should be undertaken to improve the quality of Iodine nuclear data at epithermal and fast neutron energy to obtain a reliable transmutation estimation.

We consider the Cauchy problem for the semilinear wave equation with time-dependent damping $$ \left\{ \begin{array}{ll} u_{tt} - \Delta u + b(t)u_t=|u|^{\rho}, & (t,x) \in \mathbb{R}^+ \times \mathbb{R}^N \\ (u,u_t)(0,x) = (u_0,u_1)(x), & x \in \mathbb{R}^N. \end{array}\right. (*) $$ When $b(t)=b_0(t+1)^{-\beta}$ with $b_0>0$ and $-1 < \beta 0\,(i=0,1)$, we show that the time-global solution of ($*$) does not exist provided that $1

Frequent beam trips as experienced in the existing high-power proton accelerators may cause thermal fatigue in accelerator-driven system (ADS) components, which may lead to degradation of their structural integrity and reduction of their lifetime. In this study, acceptable beam-trip frequencies of the ADS accelerator were evaluated and compared with the performance of the ADS accelerator, which was estimated based on the operational data on existing accelerators. Thermal transient analyses were performed to investigate the effects of beam trips on the reactor components, with the objective of determining the feasibility of engineering the ADS and the reliability of the accelerator. These analyses were based on the thermal responses of the following reactor components: the beam window, the fuel cladding, the inner barrel and the reactor vessel. Assuming that the annual plant availability was 70%, our results indicated three acceptable beam-trip frequencies, depending on the beam-trip duration, τ b : 2 × 10...

After the severe accident in the Fukushima-1 nuclear power plant, a large amount of contaminated stagnant water has been produced in turbine buildings and surrounding areas. This rapid communication reports the calculation of the radionuclide inventory in the core, the collection of the measured inventory in the stagnant water, and the estimation of the radionuclide release ratios from the core to the stagnant water. The present evaluation is based on data obtained before June 3, 2011. It was revealed that the release ratios of tritium, iodine and cesium were several tens of percent, while those of strontium and barium were smaller by one or two orders of magnitude. These release ratios of the Fukushima accident were equivalent to those of the TMI-2 accident.

In present study the benchmarking of TARC experiment was done. MCNPX code with JENDL-HE, JENDL-4, JENDL-3.3 and LA-150 nuclear data libraries were used for simulations. The calculated neutron fluences, energy-time correlations of resonance reactions and transmutation rates of 99 Tc and 127,129 I were compared with experimental measurements. On the base of comparative analysis of simulations the validation of nuclear data libraries was performed and recommendations were given, as follows: – the JENDL-4 library contains better neutron elastic cross-section data for natural lead in energy range from ∼0.01 eV to ∼100 eV comparing to other considered data libraries; – the JENDL-HE (JENDL-3.3) has better neutron inelastic scattering cross-sections for lead in energy range of 1–3 MeV, although the experimental errors in neutron fluence are rather large; – neutron capture cross-sections of 129 I in JENDL-4 library should be used for 129 I transmutation process simulations.

A feasibility study on natural uranium spallation target in accelerator-driven system (ADS) for minor actinide (MA) transmutation was performed. As a result of comparative study of uranium and lead-bismuth (PbBi) targets in the bare case without blanket surrounding, it was found that uranium target had better neutron generation performance, but limited by the geometrical size due to high neutron absorption in 238U. In ADS for MA transmutation, uranium used as target instead of PbBi also absorbs neutrons passing the target area. More realistic concept of pin type uranium spallation target cooled by liquid PbBi was considered aiming at enhancing spallation target performance in terms of neutron generation efficiency and operation temperature. The uranium pin target design had nothing better effects on neutron balance of such system than a conventional PbBi target in ADS and it was concluded that uranium target was not suitable for the full-scale ADS.

Reduction of burden caused by radioactive waste management is one of the most critical issues for the sustainable utilization of nuclear power. The Partitioning and Transmutation (P&T) technology provides the possibility to reduce the amount of the radiotoxic inventory of the high-level radioactive waste (HLW) dramatically and to extend the repository capacity. The accelerator-driven system (ADS) is regarded as a powerful tool to effectively transmute minor actinides (MAs) in the “double-strata” fuel cycle strategy. The ADS has a potential to flexibly manage MA in the transient phase from light water reactors (LWRs) to fast breeder reactors (FBRs), and can co-exist with FBR symbiotically and complementarily to enhance the reliability and the safety of the commercial FBR cycle. The concept of ADS in JAEA is a lead–bismuth eutectic (LBE) cooled, tank-type subcritical reactor with the power of 800 MWth driven by a 30 MW superconducting LINAC. By such an ADS, 250 kg of MA can be transmuted annually, which corresponds to the amount of MA produced in 10 units of LWR with 1 GWe. The design study was performed mainly for the subcritical reactor and the spallation target with a beam window. In Japan, Atomic Energy Commission (AEC) has implemented the check and review (C&R) on P&T technology from 2008 to 2009. In the C&R, the benefit of P&T technology, the current status of the R&D, and the way forward to promote it were discussed.

We consider the Cauchy problem for the damped wave equation with space–time dependent potential b ( t , x ) and absorbing semilinear term | u | ρ − 1 u . Here, b ( t , x ) = b 0 ( 1 + | x | 2 ) − α 2 ( 1 + t ) − β with b 0 > 0 , α , β ⩾ 0 and α + β ∈ [ 0 , 1 ) . Using the weighted energy method, we can obtain the L 2 decay rate of the solution, which is almost optimal in the case ρ > ρ c ( N , α , β ) : = 1 + 2 / ( N − α ) . Combining this decay rate with the result that we got in the paper [J. Lin, K. Nishihara, J. Zhai, L 2 -estimates of solutions for damped wave equations with space–time dependent damping term, J. Differential Equations 248 (2010) 403–422], we believe that ρ c ( N , α , β ) is a critical exponent. Note that when α = β = 0 , ρ c ( N , α , β ) coincides to the Fujita exponent ρ F ( N ) : = 1 + 2 / N . The new points include the estimate in the supercritical exponent and for not necessarily compactly supported data.

Zeolite is used for decontamination of radioactive water containing salts from seawater in the Fukushima Dai-ichi Nuclear Power Station. Evaluation of hydrogen production by water radiolysis during the process is important for a safe operation. Thus, hydrogen production from the mixture of zeolite and seawater was studied by γ-radiolysis experiment, and the hydrogen production during the process was evaluated. The measured yield of hydrogen from seawater was comparable to the primary yield in the γ-radiolysis of water. This result indicates that oxidation of hydrogen by radical products of water radiolysis is not effective in seawater. The measured yield from the mixture decreased at a high weight fraction of zeolite. However, the measured yield was higher than that expected from the direct radiolysis of water in the mixture, which would decrease proportionally to the weight fraction of water. This result suggests that the radiation energy deposited on zeolite is involved in hydrogen formation. From the measured yields, the hydrogen production rate was evaluated to be 3.6 mL/h per ton of radioactive water before the process and 1.5 L/h per ton of the waste adsorbent after the process.

The impact of partitioning and/or transmutation (PT) technology on high-level waste management was investigated for the equilibrium state of several potential fast breeder reactor (FBR) fuel cycles. Three different fuel cycle scenarios involving PT technology were analyzed: 1) partitioning process only (separation of some fission products), 2) transmutation process only (separation and transmutation of minor actinides), and 3) both partitioning and transmutation processes. The conventional light water reactor (LWR) fuel cycle without PT technology, on which the current repository design is based, was also included for comparison. We focused on the thermal constraints in a geological repository and determined the necessary predisposal storage quantities and time periods (by defining a storage capacity index) for several predefined emplacement configurations through transient thermal analysis. The relation between this storage capacity index and the required repository emplacement area was obtained. We foun...

The conceptual design study of the beam window of the spallation target is one of the critical issues in the R&D of the accelerator-driven system (ADS). In this study, the investigation to create a feasible concept of the beam window for the ADS was performed by changing the proton beam profile from the Gaussian distribution to the parabolic and the flat distributions. Detailed analyses were performed by considering the particle transport of protons and neutrons in the spallation target region, the thermal hydraulics of lead bismuth eutectic (LBE) around the beam window, and the structural strength of the beam window. The calculation results showed that the difference in the temperature between the inner and outer surfaces at the top of the beam window was reduced by changing the beam profile from the Gaussian to the parabolic and the flat distributions. By this reduction, in the parabolic case, the thermal stress at the top position was also reduced and the buckling pressure increased by about 20%. On the other hand, in the flat case, the thermal stress at the peripheral region instead of at the central position increased and the buckling pressure slightly deteriorated. The buckling mode was also changed in the flat case. From these calculation results and discussions, it was confirmed that all three cases, the Gaussian, the parabolic, and the flat, were feasible and it was concluded that the concept with the parabolic distribution would be the most feasible under the current ADS design condition.

We consider the Cauchy problem for the damped wave equation with space dependent potential V(x)u t and absorbed semilinear term |u|ρ−1 u in R N . Our assumption on V(x) ∼ (1 + |x|2)−α/2 (0 ≤ α

To confirm the reliability of calculated reactor physics parameters for the nuclear transmutation systems, the uncertainty deduced from the covariance data prepared in JENDL-3.3 is compared with the differences in the reactor physics parameters in the Monte-Carlo calculation using different nuclear data libraries, ENDF/B-VII.0 and JEFF-3.1.1. The Accelerator-Driven System (ADS) and the Minor Actinide (MA)-loaded Fast Reactor (FR) are selected as the representative transmutation systems. The criticality and void reactivity of these systems are discussed. The results show that the uncertainties deduced from the JENDL-3.3 covariance data are smaller than the differences in the reactor physics parameters among the nuclear data libraries. The cause of this discrepancy is that the covariance data of main nuclides and reactions in JENDL-3.3 are smaller than the relative differences in the cross sections among the nuclear data libraries. It is required to verify the uncertainty of the reactor physics parameters by integral experiments and to discuss the uncertainty utilization for the nuclear design accuracy.

The investigation of the beam window, which is a key component in the conceptual design of an Accelerator Driven System, has been performed. In the past studies, it was found that buckling failure due to hydrostatic pressure in the liquid lead bismuth was critical failure mode for the beam window and detailed structural analyses were performed. These investigations, however, did not consider irradiation effects by neutrons and protons. In this study, investigations based on the latest knowledge for irradiation effects obtained in the spallation target irradiation program are presented. By using the experimental data, it was found that the buckling pressure increased about 80% by the irradiation (20 dpa). It was assumed that if the beam window had integrity in the unirradiated condition, the buckling failure would not be critical issue during the ADS operation.

In the present neutronics design of the Accelerator-Driven System (ADS) cooled by lead-bismuth eu-tectic (LBE), we investigated several methods to reduce the power peak and beam current, and estimated the temperature reductions of the cladding tube and beam window from the conventional design. The methods are adjustment of inert matrix ratio in fuel in each burn-up cycle, multiregion design in terms of pin radius or inert matrix content, and modification of the level of the beam window position and the height of the central fuel assemblies. As a result, we optimized the ADS combined with the adjustment of the inert matrix ratio in each burn-up cycle, multiregion design in terms of inert matrix content and deepened window level. The maximum temperatures of the optimized ADS at the surface of the cladding tube and the beam window were reduced by 91 and 38°C, respectively. The maximum beam current was improved from 20.3 to 15.6 mA.

Irradiation damage to the beam window in the concept of 800MWth accelerator-driven system is evaluated. Heat produced in the window is also evaluated. Transport of proton and neutron up to 3.0 GeV is calculated by both PHITS that is the Monte Carlo code for particles and heavy ions and TWODANT that is two-dimensional deterministic transport code. The beam window is irradiated at the center of the accelerator-driven system with 20 MW proton beam power and neutron from the core during 300 full power days. Heat, displacement per atom, production rate of hydrogen and helium isotopes, and neutron and proton fields are estimated, assuming the Gaussian and flat beam profiles.

The feasibility for the lead-bismuth-cooled accelerator-driven system (ADS) to transmute minor actinides partitioned from high-level radioactive waste is discussed. Since lead-bismuth will cause co...

We consider the Cauchy problem for the damped wave equation u t t − Δ u + u t = | u | ρ − 1 u , ( t , x ) ∈ R + × R N and the heat equation ϕ t − Δ ϕ = | ϕ | ρ − 1 ϕ , ( t , x ) ∈ R + × R N . If the data is small and slowly decays likely c 1 ( 1 + | x | ) − k N , 0 k ⩽ 1 , then the critical exponent is ρ c ( k ) = 1 + 2 k N for the semilinear heat equation. In this paper it is shown that in the supercritical case there exists a unique time global solution to the Cauchy problem for the semilinear heat equation in any dimensional space R N , whose asymptotic profile is given by Φ 0 ( t , x ) = ∫ R N e − | x − y | 2 4 t ( 4 π t ) N / 2 c 1 ( 1 + | y | 2 ) k N / 2 d y provided that the data ϕ 0 satisfies lim | x | → ∞ 〈 x 〉 k N ϕ 0 ( x ) = c 1 ( ≠ 0 ) . Even in the semilinear damped wave equation in the supercritical case a time global solution u with the data ( u , u t ) ( 0 , x ) = ( u 0 , u 1 ) ( x ) is shown in low dimensional spaces R N , N = 1 , 2 , 3 , to have the same asymptotic profile Φ 0 ( t , x ) provided that lim | x | → ∞ 〈 x 〉 k N ( u 0 + u 1 ) ( x ) = c 1 ( ≠ 0 ) . Those proofs are given by elementary estimates on the explicit formulas of solutions.

Partitioning and/or transmutation (PT) technology affects the disposal concept of high-level radioactive waste (HLW). We studied how cooling in the predisposal storage period may affect the design of the emplacement area in a repository for radioactive wastes produced by a light-water-reactor nuclear system that uses PT technology. Three different fuel cycle scenarios involving PT technology were analyzed: 1) partitioning process only (separation of some fission products), 2) transmutation process only (separation and transmutation of minor actinides), and 3) both partitioning and transmutation. The necessary predisposal storage periods for some predefined emplacement configurations were determined through transient thermal analysis, and the relation between the storage period and the emplacement area was obtained. For each scenario, we also estimated the storage capacity required for the dry storage of the heat-generating waste forms. The contributions of PT technology on the storage and disposal were di...