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A new nuclear data library for the advanced lattice physics code CASMO5 has been prepared based on JENDL-5. In JENDL-5, the range of data, such as the number of nuclides, has been dramatically expanded for general-purpose applications. At the same time, many essential nuclides for conventional light water reactor (LWR) analysis have also been modified based on state-of-the-art evaluations. The new JENDL-5–based CASMO5 library was prepared by replacing as much of the nuclear data of the current CASMO5 ENDF/B-VII.1–based library as possible with JENDL-5. This study performs initial verification and validation of the new library under typical LWR conditions. Verifications were performed based on the Organisation for Economic Co-operation and Development/Nuclear Energy Agency burnup credit criticality safety benchmark phase III-C, and the calculated kinf and fuel compositions of the boiling water reactor fuel assembly were compared with reported benchmark results. Comparison with the MCNP6.2 result was also performed using the same benchmark model. In addition, a tank-type critical assembly critical experiment and Takahama-3 postirradiation experiment were used for validation. The results indicate that the new library performs well and is comparable to the ENDF/B-VII.1–based library in predictions of reactivity and fuel compositions for typical LWR systems. [ABSTRACT FROM AUTHOR]

• The effect of addition U233O 2 to (Th, Pu) fuel on the safety parameters of PWR assembly. • Reactivity variation in different thorium fuels. • The power distribution in thorium-based fuels. • The behavior of conversion ratio with burnup in the considered fuels. • The effect of (Th, U233, Pu) on the minor actinide production. The aim of this work is not only to demonstrate how to introduce thorium as an alternative nuclear fuel into pressurized water reactors, but also to examine the several advantages that (Th, U233, Pu) has over (Th, U233) and (Th, Pu) fuels when (U233O 2) is mixed with (Th-Pu)O 2 fuel. To investigate these advantages, three thorium fuels, (Th, U233), (Th-U233-Pu) and (Th-Pu) are simulated and analyzed by the MCNPX 2.7 code. The reactor safety aspects of these fuels are compared with those of the reference UO 2 fuel. The concentration of Th-233 and fissile isotopes is adjusted to obtain criticality (Keff = 1.03) at the same burnup value (35 GWd/tonne) for all fuel types. Based on the predicted results, the author revealed that: 1) 56 % of plutonium (238-242) is depleted in the case of (Th-U233-Pu), while 41 % of plutonium is depleted in the (Th-Pu) at 60 GWd/tonne. This result indicates that (Th-U233-Pu) is more proliferation-resistant than (Th-Pu). 2) In terms of improving the quantity of radioactive waste, (Th, U233, Pu) model produces 0.6 and 0.9 times the amount of Am (241-243) and Cm (242-244) in UO 2 , respectively. On the other hand, (Th, Pu) fuel produces about 7 times the amount of Am and 8 times the amount of Cm in UO 2 fuel. 3) The fuel temperature coefficient (FTC) and moderator temperature coefficient (MTC) of (Th-U233-Pu) are more negative than that of UO 2 and (Th-U233) at the early stages of fuel burnup (0–5) GWd/tonne.4) the total mass of fissile isotopes in (Th-U233-Pu) is greater than that of UO 2 and (Th-U233) and as a result, the fuel life time is longer in the case of (Th-U233-Pu). 4) The conversion ratio provided by (Th-U233-Pu) is greater than that provided by UO 2 and (Th-U233).This behavior is expected owing to the hardening of neutron spectrum in (Th-U233-Pu). 5) The peaking factors are lower in the case of (Th-U233-Pu) compared to (Th-Pu). 6) The concentration of xenon and samarium is noticeably lower for (Th-U233-Pu) in comparison with (Th-Pu) and consequently, the time required for restarting the reactor is accelerated. [ABSTRACT FROM AUTHOR]

This paper proposes a methodology for burnup credit of spent nuclear fuel (SNF) pool storage based on the axial burnup of the fuel assemblies. Following the NRC guidance and normative for NPP safety, axial burnup profiles from a database were analysed, and a conservative theoretical profile that fits the safety conditions was obtained. This profile sets a limit on the upper criticality of the assemblies, and it is used to obtain the pool loading curve, which dictates if a fuel assembly fits to be loaded in the studied SNF pool based on its initial enrichment level and burnup. Although it takes burnup credit, this methodology follows strict safety normative and is deemed suitable for further analysis in SNF pools. This methodology can be used to conservatively reduce the requisites for the storage of spent fuel by establishing a burnup profile that bounds all other profiles. This approach allows for more flexibility in the spent fuel pool loading. [ABSTRACT FROM AUTHOR]

The article offers a critical look on the prospects and production of nuclear fuel. According to the author, most of the significant uranium-producing nations have slew virtually all security restrictions on information relating to nuclear raw materials since 1955. For him, it is important that resources of nuclear fuels be considered from a world point of view. All nations planning or considering atomic power development need access to the maximum information on the availability and cost of the nuclear fuel.

The specific features of calculation of the effective multiplication factor using the Monte Carlo method for weakly coupled and non-asymptotic multiplying systems are discussed. Particular examples are considered and practical recommendations on detection and Monte Carlo calculation of systems typical in numerical substantiation of nuclear safety for VVER fuel management problems are given. In particular, the problems of the choice of parameters for the batch mode and the method for normalization of the neutron batch, as well as finding and interpretation of the eigenvalue spectrum for the integral fission matrix, are discussed. [ABSTRACT FROM AUTHOR]

The effective neutron multiplication factor (k eff ) as a function of burnup for different volume coolant (CoR) and fuel (FR) to cell ratio is presented. Additionally the Conversion Ratio (CR) of Th-232 to U-233, concentration of U-233, fissile and fission products calculation as a function of burnup are presented. The assembly is a critical reactor which makes volumes of coolant and fuel changes possible. In addition, an analytical model of calculation of k eff as a function of U-233 and a poison concentration in equilibrium state are presented. One can achieve the criticality of Thorium Breeder Reactor (TBR) for enough high average neutron energy which one can obtain in Fast Breeder Reactor (FBR) only. The maximal value of CR and burnup for case of k eff ≥ 1 achieves 1.4 and 360 GWd/MTU, correspondently. The calculations were done with a MCNPX 2.7 code using F 2 Be, Na and Pb coolants. [ABSTRACT FROM AUTHOR]

A calculation code was developed for transient analysis of the total released energy in a criticality accident with two fuel-solution tanks. The calculation method was confirmed to be effective for transient analysis in a fuel-solution-tank system. The verification of the code was performed by comparing the calculation results with those of TRACY experiments. In the code, neutronic coupling between the tanks is treated. Using this code, transient analyses were performed for a two-tank system using different feedback models. The analyses confirmed that the total energy released in criticality accident with two fuel-solution tanks can be estimated using the knowledge of the total energy released in a single-fuel-solution-tank system. [ABSTRACT FROM AUTHOR]

In this study, the characteristics of changes in reactivity due to increasing turnup of uranium-fueled fast reactors are analyzed. A new classification system for nuclear reactor cores based on their uncontrolled tendency for reactivity changes during turnup was introduced and the design-optimization strategy for any fast reactor core aimed at a minimized reactivity swing is outlined. The 235U feed-fuel enrichment level that minimizes the burnup reactivity swing of a sodium-cooled metallic-fueled core is 10% to 12.5% for an average target fuel burnup of 1% to 20% FIMA (fission of initial metal atom). The higher the target burnup of the system, the lower the feed-fuel enrichment level that minimizes swing. The minimum attainable swing for a 125-MW (thermal) metallic-fueled sodium-cooled core is found to be ~200 pcm for 5% FIMA burnup and increases to ~ 800 pcm for a system aiming at 10% FIMA. In general, if the target discharge burnup is doubled, the minimum attainable burnup reactivity swing quadruples. Any optimized minimum reactivity swing core will form a positive parabolic uncontrolled reactivity trajectory with burnup, where the beginning of cycle and end of cycle reactivities are equal. Uranium-fueled fast cores with minimized burnup reactivity swing are net consumers of fissile material, with a fissile conversion ratio in the range of 0.7 to 0.9. [ABSTRACT FROM AUTHOR]

We propose a method to estimate sensitivity profiles of neutronics parameters with respect to nuclear data in random sampling-based uncertainty quantification calculations. The proposed method is tested to estimate sensitivity profiles of fast neutron systems criticalities. A high order effect in sensitivity profile estimation is found to be quite important, so a reverse sampling method is developed to mitigate the high order effect. With this reverse sampling method, detail energy group-wise sensitivity profiles can be estimated even though some fluctuations are observed in specific sensitivity profiles. Energy-integrated sensitivity profiles can be accurately calculated with the proposed method. With the estimated sensitivity profiles, partial uncertainties, that are neutronics parameters uncertainties induced by specific nuclear data uncertainties, are also calculated. Numerical tests reveal that the proposed method reproduces quite well the reference partial uncertainties. A simple and practical partial uncertainty estimation method, which only requires a covariance matrix between neutronics parameter and nuclear data, is also tested and assessed. [ABSTRACT FROM AUTHOR]

Highlights: [•] A three-dimensional model of a PWR fuel were simulated. [•] Results using TRITON/T6-DEPL module in SCALE 6.0 and two libraries (238 and 44 groups) were compared. [•] Variations in the infinite neutron multiplication factor and the nuclides concentrations, both under control rod insertion effects were analysed. [•] Results show very good agreement with those published by OECD. [Copyright &y& Elsevier]

Highlights: [•] We have identified areas where nuclear safety and security share similarities as well areas where these two differ. [•] Application of concepts, criteria, and operational practices used for safety, to nuclear security, enhances protection. [•] Integration of safety and security measures such that implementation of one does not compromise the other. [•] To improve the synergy, awareness of the intersection between nuclear safety and security should be increased. [Copyright &y& Elsevier]

Highlights: [•] An internally and externally cooled annular fuel was investigated and adapted to a typical VVER-1000 reactor. [•] Ordinary solid pins as well as annular pins are fully investigated. [•] MDNBR calculation was carried out for these hot annular fuel rods. [•] Pitch optimization was performed for each annular case. [•] An annular pins fuel assembly configuration is proposed for the VVER-1000 reactor. [Copyright &y& Elsevier]

Highlights: [•] VVER-1000 mock-up criticality coefficients determination. [•] Calculations of criticality using various libraries. [•] Calculations to experiment comparison of criticality coefficients. [•] Influence of thermal scattering law sublibraries on resulted k eff. [Copyright &y& Elsevier]

Abstract: The concepts of static and dynamic reactivity are discussed and the practical differences between them are noted. Special emphasis is given to the use of these concepts in defining uncertainties in reactivity calculations due to uncertainties in cross sections. Currently, in order to define uncertainty, it is the static reactivity that is used. But we show that this may give a non-conservative result compared with the dynamic method. For example, we show that while a static calculation of reactivity involving uncertainty may give a mean value of reactivity which is less than zero, the corresponding dynamic method can yield a value for the time constant which is greater than zero. As the time constant is the physical parameter of interest then the static method must be used with caution. [Copyright &y& Elsevier]

Spent nuclear fuel transported in large casks must remain subcritical in all credible configurations for normal operation and hypothetical accident conditions. The effects on spent nuclear fuel reactivity from "worst-case" accident scenarios were surveyed in NUREG/CR-6835, "Effects of Fuel Failure on Criticality Safety and Radiation Dose for Spent Fuel Casks." The survey used scenarios that were postulated to provide theoretical upper limits for reactivity effects of fuel relocation, although they were described as going "beyond credible conditions." These scenarios involved physical changes either to fuel assembly rod arrays or to collections of fuel pellets with the fuel skeleton removed. To provide more credible estimates of the probability and maximum reactivity changes, a process is presented that deconstructs each scenario into a set of subscenarios and identifies the physical phenomena required to create the subscenario. The boundary between credible but unlikely scenarios and incredible scenarios is more easily discernible with this process. For marginally credible worst-case scenarios, it is concluded that the maximum reasonable reactivity increase either is less than the mandated administrative nuclear criticality safety margin for scenarios involving physical changes to fuel assembly rod arrays or is a substantial reactivity decrease for scenarios involving collections of fuel pellets. A cask designer could apply scenario deconstruction to evaluate the physical limits that apply to a particular transportation cask, and perform calculations specific to a particular cask design to show that criticality safety requirements are met. [ABSTRACT FROM AUTHOR]

The present paper describes the evaluation of the control rod worth data obtained in the Monju restart core. The best-estimate value and its uncertainty are evaluated in detail. As in the criticality evaluation, data obtained in the previous test are evaluated in the same level of detail. Evaluated results of control rod worth are consistent among control rods at symmetrical positions for a random uncertainty. The experimental uncertainty is ±2%, where the dominant uncertainty originates from the delayed neutron parameters used in the period method. The correlation in the uncertainties is also evaluated among different control rods and tests. One can discuss a difference in worth among different control rods and cores without a detailed knowledge of the original uncertainty evaluation. Based on the evaluated data, calculation accuracy is investigated with JENDL-3.3 and JENDL-4.0. It is confirmed that the calculation accuracy is within an experimental uncertainty of ±2% for each layer and 10B content. [ABSTRACT FROM AUTHOR]

The Japanese prototype fast breeder reactor Monju restarted its system startup test in May 2010 after a 14-year interruption. In the first stage of the test, reactor physics parameters have been measured at a zero power level. The present paper describes the evaluation of the criticality data. The best-estimate value and its uncertainty are evaluated as accurately as possible, following the guidelines recommended by the International Criticality Safety Benchmark Evaluation Project. The restart core contains 1.5 wt% of 241Am, which is three times more than the previous test. To extract the influence of the 241Am accumulation on calculation accuracy, criticality data obtained in the previous test are evaluated at the same level of detail. The calculation accuracy is investigated with four major nuclear data libraries. It is confirmed that the accuracy is within 0.3% Δk/k, a 2σ value of experimental uncertainty, with JENDL-3.3, JENDL-4.0, and ENDF/ B-VII.0. The reactivity change due to the 241Pu decay can be simulated within an accuracy of 1% with JENDL-4.0 and JEFF-3.1. [ABSTRACT FROM AUTHOR]

The present technical note describes the evaluation of the isothermal temperature coefficient data obtained in the Monju restart core. As in the preceding evaluations on the criticality and the control rod worth, the best-estimate value and its uncertainty are evaluated as accurately as possible. Data obtained in the previous test are evaluated in the same level of detail. The measured data show that the fuel composition change from the previous test decreases the magnitude of the temperature coefficient by ~8%. Through a sensitivity analysis, it is confirmed that the decrease is mainly brought by the composition of 241Pu and 241Am. The best calculation accuracy within the experimental uncertainty of 2% is attained with JENDL-4.0 for the previous core; however, such a good accuracy is not achieved for the restart core. A further experimental investigation is required to solve the dependence of calculation accuracy on the core configurations. [ABSTRACT FROM AUTHOR]

Abstract: This study is being carried on as a provision of achieving enhanced irradiation facilities in the TRIGA MARK II research reactor core. Two reconfigured cores, each one having three irradiation channels are being analyzed with the Monte Carlo code MVP and the nuclear data library JENDL-3.3. The existing core has only one irradiation channel at the center. The graphite dummies are rearranged to surround the new irradiation channels in the new configurations. The results for each reconfigured core are compared with the existing core as reference. From this study important information are obtained for enhanced as well as economic utilization of TRIGA core. The results show positively the possibility to introduce new irradiation channels without disturbing the core excess reactivity and neutron flux significantly as well as keeping the shutdown margin and peaking factor close to the reference. [Copyright &y& Elsevier]

Abstract: Recently, annular fuel rods are proposed for both PWRs and BWRs to achieve more average power density and therefore many neutronic as well as thermal-hydraulics calculations have been made to find more performance in the future reactors. Also, some safety margins are studied for the proposed uranium-nitride (and/or other fissile materials) annular pins. Our aim herein is to study two important safety coefficient of the annular fuel core. These are “prompt reactivity” and “power coefficients”, where all investigations are made using MCNP-5 code. Also, a thermal resistance model for annular fuel heat transferring were given and then we have calculated its (a case) thermal resistance numerically. [Copyright &y& Elsevier]

Abstract: The current objective of coprecipitating uranium, and minor actinides in order to fabricate a new nuclear fuel by direct (co)precipitation for further transmutation, requires to develop specific technology in order to meet the following requirements: nuclear maintenance, criticity, and potentially high flowrates due to global coprecipitation. A new type of device designed and patented by the CEA was then tested in 2007 under inactive conditions and with uranium. The patent is for organic confinement in a pulsed column (PC). Actually, pulsed columns have been working for a long time in a nuclear environment, as they allow high capacity, sub-critical design (annular geometry) and easy high activity maintenance. The precipitation reaction between the oxalate complexing agent and a surrogate nitrate – cerium(III) or neodymium(III) alone, or coprecipitated uranium(IV) and cerium(III) – occurs within an emulsion created in the device by these two phases flowing with a counter-current chemically inert organic phase (for example tetrapropylene hydrogen—TPH) produced by the stirring action of the column pulsator. The precipitate is confined and thus does not form deposits on the vessel walls (which are also water-repellent); it flows downward by gravity and exits the column continuously into a settling tank. The results obtained for precipitation of cerium or neodymium alone in a short column of small diameter have demonstrated that high throughputs are feasible without system malfunctions. The measured particle size of the precipitates ranges from 20 to 40μm on average, and the measured device outflow indicates that the precipitation reaction is complete. These results suggest that this laboratory design can be extrapolated to an industrial column. Moreover, a recent test campaign demonstrated that a uranium–cerium coprecipitate easily forms when the two nitrates are mixed in a pulsed column of the same size operating under very similar process conditions. Qualitatively, the coprecipitate meets the process requirements to filter, calcine and fabricate the new nuclear fuel in the downstream steps. [Copyright &y& Elsevier]

U isotope is used as a booster fissile fuel material in the form of mixed ThC/UC fuel in a Canada Deuterium Uranium (CANDU) fuel bundle in order to assure the initial criticality at startup. Three different fuel compositions have been used: (1) 97% ThC+3% UC, (2) 98% ThC+2% UC and (3) 99% ThC+1% UC. The temporal variation of the criticality k and the burn-up values of the reactor have been calculated by full-power operation for a period of 20 years. The criticality starts by k=1.541, 1.355 and 0.995 for modes of (1), (2) and (3) fuel compositions, respectively. A sharp decrease in the criticality has been observed in the first 2 years as a consequence of rapid U burnout fuelling with (1) and (2) modes. The criticality becomes quasi-constant after the second year and remains above k∼1.06 for 20 years. After the second year, the CANDU reactor begins to operate practically as a thorium burner. Very high burnup could be achieved with the same fuel materials (up to 500 000 MWday t), provided that the fuel rod claddings would be replaced periodically (after every 500 00 or 100 000 MWday t). The reactor criticality will be sufficient for fuelling with (1) and (2) modes until a great fraction of the thorium fuel is burnt up. This would reduce fuel fabrication costs and nuclear waste mass for final disposal per unit energy drastically. Copyright © 2010 John Wiley & Sons, Ltd. [ABSTRACT FROM AUTHOR]

Abstract: SIMMER-III, a neutronics and thermal-hydraulics coupled code, was originally developed for core disruptive accident analyses of liquid metal cooled fast reactors. Due to its versatility in investigating scenarios of core disruption, the code has also been extended to the simulation of transients in thermal neutron systems such as the criticality accident at the JCO fuel fabrication plant, and, in recent years, applied to water-moderated thermal research reactor transient studies, too. Originally, SIMMER considered only cylindrical fuel pin geometry. Therefore, implementation of a plate-type fuel model to the SIMMER-III code is of importance for the analysis of research reactors adopting this kind of fuel. Furthermore, validation of the SIMMER-III modeling of light water-cooled thermal reactor reactivity initiated transients is of necessity. This paper presents the work carried out on the SIMMER-III code in the framework of a KIT and IRSN joint activity aimed at providing the code with experimental reactor transient study capabilities. The first step of the job was the implementation of a new fuel model in SIMMER-III. Verification on this new model indicates that it can well simulate the steady-state temperature profile in the fuel. Secondly, three cases with the shortest reactor periods of 5.0ms, 4.6ms and 3.2ms among the Special Power Excursion Reactor Tests (SPERT) performed in the SPERT I D-12/25 facility have been simulated. Comparison of the results between the SIMMER-III simulation and the reported SPERT results indicates that although there is space for further improvement on the modeling of negative feedback mechanisms, with this plate-type fuel model SIMMER-III can well represent the transient phenomena of reactivity driven power excursion. [ABSTRACT FROM AUTHOR]

Abstract: Ignalina NPP (Lithuania) comprises two Units with RBMK-1500 reactors. After the Unit 1 of the Ignalina Nuclear Power Plant was shut down in 2004, approximately 1000 fuel assemblies from Unit 1 were available for further reuse in Unit 2. The fuel-transportation container, vehicle, protection shaft and other necessary equipment were designed in order to implement the process for on-site transportation of Unit 1 fuel for reuse in the Unit 2. The developed equipment can be used also in decommissioning phase for fuel transportation to fuel storage facilities. The set of this equipment can be applied for NPPs with RBMK type reactors. The purpose of this paper is to introduce the content and main results of safety analysis, focusing attention on the container used for spent fuel transportation. The structural integrity, thermal, radiological and nuclear criticality safety calculations are performed to assess the acceptance of the proposed set of equipment. The safety analysis demonstrated that the proposed nuclear fuel-transportation container and other equipment are in compliance with functional, design and regulatory requirements, assure the required safety level for both normal operation and accident conditions and can be used for fuel transportation. [ABSTRACT FROM AUTHOR]

Abstract: Time evolution of criticality and burn-up grades of the Fixed Bed Nuclear Reactor (FBNR) are investigated for alternative fuels. These are: (1) low enriched uranium, (2) weapon grade plutonium, (3) reactor grade plutonium, and (4) minor actinides in the spent fuel of light water reactors (LWRs). The criticality calculations are conducted with SCALE 5.1 using S8-P3 approximation in 238 neutron energy groups with 90 groups in thermal energy region. The main results of the study can be summarized as follows: [(1)] Low enriched uranium (UO2): FBNR with an enrichment grade of 9% and 19% will start with k eff =1.2744 and k eff =1.36 and can operate ∼8 and >15years with the same fuel charge, where criticality drops to k eff =1.06 and a burn-up grade of 54000 and >110000 MW.D/t can be attained. [(2)] Weapon grade plutonium: Such a high quality nuclear fuel suggests to be mixed with thorium. Second series of criticality calculations are conducted with fuel compositions made of thoria (ThO2) and weapon grade PuO2, where PuO2 component has been varied from 1% to 100%. Criticality with k eff >1.0 is achieved by ∼2.5% PuO2. At 4% PuO2, the reactor criticality will become satisfactory (k eff =1.1121), rapidly increasing with more PuO2. A reasonable mixture will by around 20% PuO2 and 80% ThO2 with a k eff =1.2864. This mixed fuel would allow full power reactor operation for >20years and burn-up grade can reach 136000MW.D/t. [(3)] Reactor grade plutonium: Third series of criticality calculations are conducted with fuel compositions made of thoria and reactor grade PuO2, where PuO2 is varied from 1% to 100%. Reactor becomes critical by ∼8% PuO2 content. One can achieve k eff =1.2670 by 35% PuO2 and would allow full power reactor operation also for >20years and burn-up grade can reach 123000MW.D/t. [(4)] Minor actinides in the spent fuel of LWRs: Fourth series of criticality calculations are conducted with fuel compositions made of thoria and MAO2, where MAO2 is varied from 1% to 100%. Reactor becomes critical by ∼17% MAO2 content. Reasonably high reactor criticality (k eff =1.2673) is achieved by 50% MAO2 for a reactor operation time of 15years with a burn up of 86000MW.D/t without fuel change. On that way, the hazardous nuclear waste product can be transmuted as well as utilized as fuel. [Copyright &y& Elsevier]

Abstract: There are many application fields for fast neutrons. The main application fields of the fast neutrons are accelerator-driven sub-critical systems (ADS) and fusion–fission (hybrid) reactor systems for fission energy production. Thorium (Th) and uranium (U) are nuclear fuels in fusion–fission (hybrid) reactor systems and bismuth (Bi) is also the target nucleus in the ADS reactor systems. In this study, neutron production cross sections produced by (d,xn) reactions for spallation targets such as 209Bi, 232Th, 235U and 238U have been investigated. New evaluated hybrid model and geometry dependent hybrid model have been used to calculate the pre-equilibrium neutron production cross sections. For the reaction equilibrium component, Weisskopf–Ewing model calculations have been preferred. The obtained results have been discussed and compared with the available experimental data and found in agreement with each other. [Copyright &y& Elsevier]

A collection of software tools developed for the Mayak worker study is described. IMPDOS is software for modelling, data analysis, and activity and dose calculations using the bioassay and postmortem data from Mayak workers provided by Southern Urals Biophysics Institute. The capabilities include: parameter fitting of data for individual cases, Bayesian dose calculations using the fit results for collections of cases with extensive data as a biokinetic prior, and database storage of results for retrieval, analysis and interpretation. [ABSTRACT FROM AUTHOR]

We examined the molecular mechanisms underlying the formation of the thalamocortical pathway in the cerebral neocortex of normal and reeler mutant mice. During normal development of the mouse neocortex, thalamic axons immunoreactive for the neural cell adhesion molecule L1 rarely invaded the cortical plate and ran centered in the subplate which is immunoreactive for neurocan, a brain-specific chondroitin sulfate proteoglycan. On the other hand, in homozygous reeler mutant mice, thalamic axons took an aberrant course to run obliquely through the cortical plate. Injection of bromodeoxyuridine at embryonic day 11 specifically labeled subplate neurons in normal mice, whilst in the reeler neocortex it labeled cells scattered in the cortical plate as well as in the superficial layer (superplate). Neurocan immunoreactivity was associated with the bromodeoxyuridine-positive cells in the superplate, as well as being present in oblique bands within the cortical plate, along which L1-bearing thalamic axons preferentially ran. The present results support our previous hypothesis proposed for normal rats that a heterophilic molecular interaction between L1 and neurocan is involved in determining the thalamocortical pathway within the neocortical anlage [T. Fukuda et al. (1997) Journal of Comparative Neurology, 382, 141–152]. [ABSTRACT FROM AUTHOR]

The aim of this work is to present a new approach based upon the application of the LTSN method to solve the eigenvalue problem in a slab, namely the problem of criticality. The advances are based on a new iterative method for the solution of the system of equations for the boundary conditions and a new search methodology for the keff based on the bisection method. This method proved to be reliable, specially after the introduction of the restarted generalized minimum residual (RGMRE). We report numerical simulations for keff and critical dimensions determination considering the SN multigroup problem in heterogeneous slab. [Copyright &y& Elsevier]

Self-Organized Criticality (SOC) phenomena could have a significant effect on the dynamics of ecosystems. The Bak–Sneppen (BS) model is a simple and robust model of biological evolution that exhibits punctuated equilibrium behavior. Here, we will introduce random version of BS model. We also generalize the single objective BS model to a multiobjective one. [ABSTRACT FROM AUTHOR]

The criticality problem of the one-speed transport equation with isotropic scattering for reflected spheres is studied using the PN method. It is observed that the method diverges as one increases the order of the PN approximation. Before this divergence, the method yields results that agree with those of the integral equation. [Copyright &y& Elsevier]

Solutions for a flat thermal flux in a two-group diffusion model are found for a range of assumptions, since this is a requisite for both a minimum and a maximum fuel loading in the model. It is shown that a maximum fuel loading then arises when such a region, at unity infinite multiplication factor, is complemented by an outer core to bring the finite reactor critical. Correspondingly, when a core with a flat flux is surrounded by a reflector, the core fuel may be so distributed as to flatten the flux using the ‘flux-trap’ phenomenon and provide a minimum loading. Critical conditions and necessary fuel density are derived in finite and infinitely thick reflectors. A curiosity is the possibility of observing ‘negative’ reflector savings. Fuel savings are estimated for a range of models for the minimum loading compared to a correspondingly critical uniform core loading. In some circumstances a saving of up to 70% is indicated with economic and safety implications. It is shown how the reduction of the minimum loading solution from a two- to a one-group model retains the flat flux in the core but fails to satisfy the thermal boundary condition unless a reflector, with the flux-trap, is used so that without it, the two-group minimum fuel loading solution cannot fully transform to a minimum loading in one group. The full solution for a two-group model flat flux core with finite thickness reflector is given. It is shown that as the reflector is reduced it becomes necessary to increase the fuel density at the centre to a point where this exceeds the capability of the chosen fuel, thus providing a secondary criticality equation. The increasing steepness (negative slope) of the flux distribution is required to make the flux trap phenomenon possible in the reducing reflector region. The range in which there may be two reflector thicknesses leading to the same size core, but different fuel distributions observed by Williams, is determined. Constrained solutions that limit either the size of core or the maximum fuel density are considered, generalising the original work of Goertzel to a practicable core design. [Copyright &y& Elsevier]

This paper examines the possibility of criticality in a nuclear waste repository. The estimated probabilities are rough bounds and do notentirely dismiss the possibility of a critical condition; however, they do point to the difficulty of creating conditions under which a critical mass could be assembled (i.e., corrosion of containers, separation of neutron absorbers from the fissile material and collapse or precipitation of the fissile material). In addition, should a criticality occur in or near a container, the bounding consequence calculations showed that fissions from one critical event are quite small (< ~1020 fissions, if similar to aqueous and metal accidents and experiments). Furthermore, a reasonable upper bound of total critical events of 1028 fissions corresponds to only 0.1% of the number of fissions represented by the spent nuclear fuel inventory in a repository containing 70,000 metric tons of heavy metal (MTHM) (the expected size for the proposed repository at Yucca Mountain, Nevada. [ABSTRACT FROM AUTHOR]

This paper presents preliminary criticality safety assessments performed by the Paul Scherrer Institute (PSI) in cooperation with the Swiss National Cooperative for the Disposal of Radioactive Waste (Nagra) for spent nuclear fuel disposal canisters loaded with Swiss Pressurized Water Reactor (PWR) UO2 spent fuel assemblies. The burnup credit application is examined with respect to both existing concepts: taking into account actinides only and taking into account actinides plus fission products. The criticality safety calculations are integrated with uncertainty quantifications that are as detailed as possible, accounting for the uncertainties in the nuclear data used, fuel assembly and disposal canister design parameters and operating conditions, as well as the radiation-induced changes in the fuel assembly geometry. Furthermore, the most penalising axial and radial burnup profiles and the most reactive fuel loading configuration for the canisters were taken into account accordingly. The results of the study are presented with the help of loading curves showing what minimum average fuel assembly burnup is required for the given initial fuel enrichment of fresh fuel assemblies to ensure that the effective neutron multiplication factor, k e f f , of the canister would comply with the imposed criticality safety criterion. [ABSTRACT FROM AUTHOR]