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A method to monitor the fouling resistance of a heat exchanger, especially the thermosiphon reboiler, is presented. The historical hourly data of reboiler and distillation column are used to analyze fouling factor. The initial value of overall heat transfer coefficient and the current overall heat transfer coefficient are needed to calculate fouling resistance. Since the initial heat transfer coefficient is not constant but is affected by operating conditions such as shell/tube flow rate and temperature, a relationship between initial heat transfer coefficient and these operating conditions has to be developed to calculate the fouling factor. However, sometimes important data such as tube-side flow, inlet/outlet temperature of hot/cold fluid, and liquid level in shell/tube, which are needed to bring out the fouling condition of equipment, are missing. With such insufficient information, the fouling behavior can be analyzed by building a new equation based on measurable data. Dimensional analysis determine...

An alternative definition of neutron multiplication factor measured by the neutron source multiplication (NSM) method is newly proposed. This newly defined neutron multiplication factor, kdet, is derived on the basis of neutron detection process in a subcritical system with an external neutron source. The definition of kdet is expressed as a ratio of total number of detected fission-neutrons to total number of detected all neutrons. In this paper, a heuristic derivation of kdet is presented, and another interpretation of kdet is explained by using the detector importance function. Based on the idea of kdet, the measurement principle of NSM method is reinterpreted, and the correction factors in the NSM method are clarified. In order to verify our proposed NSM method, numerical analysis of the NSM method is carried out. The numerical results suggest that target neutron multiplication factors of the NSM method can be well estimated even without any corrections by putting a neutron detector where the effective neutron multiplication factor keff is well approximated by kdet.

Multicycle optimization was carried out by assuming power sharing of each fuel batch as an independent parameter; that is, the power sharing of each fuel batch was considered as an optimization variable. The steepest descent method was used to optimize the power sharing for multiple cycles. Two different optimizations were carried out, i.e., multicycle and successive single-cycle optimizations. In the former, the power sharing of each fuel batch in each cycle was simultaneously optimized for multiple cycles. In the latter, optimization of the power sharing in a cycle was carried out, and then optimization in the next cycle was carried out. Maximization of discharge burnup and minimization of the number of fresh fuel assemblies were considered as the objective functions. The calculation results qualitatively and quantitatively clarify the implicit adverse effect of the single-cycle optimization, which is usually used in current core designs. Under the calculation conditions of the present study, the difference in the number of fresh fuel assemblies between multicycle and successive single-cycle optimizations is 2–3 fuel assemblies per cycle. Comparison of the power sharing obtained by both methods would provide insights to correct the adverse effect of the single-cycle optimization.

The spatial discretization form of the space-dependent reactor kinetics equation is a first-order simultaneous ordinary differential equation in time. Conventional numerical methods of the space-dependent kinetics equation, i.e., the generalized Runge-Kutta method, the implicit method (backward Euler method), and the Theta method, are based on the time difference approximation. However, the present study adopts the analytical solution of the space-dependent kinetics equation expressed by the matrix exponential and no time difference approximation is used. In this context, our present approach is classified as an explicit method in which no iteration calculation on space and energy is necessary. The Krylov subspace method is used to evaluate the matrix exponential observed in the solution of the spatially discretized space-dependent kinetics equation. The Krylov subspace method is implemented into a space-dependent kinetics solver. In order to examine the effectiveness of the Krylov subspace method, the TW...

The pin-by-pin fine mesh BWR core analysis code SUBARU has been developed as a next-generation BWR core analysis code. SUBARU is based on the SP3 theory and the response matrix method is used for flux calculations. The SP3 theory consists of the 0th and 2nd order neutron fluxes. Therefore, the relations among the 0th and 2nd order partial neutron currents and the fluxes are required to apply the response matrix method. SUBARU is approximated the relations among the partial neutron currents and the fluxes are similar to that the diffusion theory. Our previous study revealed that the prediction accuracy of SUBARU is much higher than that of conventional core analysis codes. However, validity of the above approximation is not directly investigated so far. Therefore, relations among the partial neutron currents and the fluxes are theoretically derived and calculation results with the rigorous and the conventional formulations are compared. The calculation results indicate that the approximation of the conventional formulation is appropriate for the BWR core analysis.

An improved inverse analysis model for fuel loading pattern (LP) optimization is proposed to reproduce a desired in-core power distribution from available fuel assemblies based on the inverse analysis of diffusion equations. In the present approach, the inverse analysis is simplified as a fixed-source problem rather than an eigenvalue problem in a previous work. By using the one-to-one correspondence between the power distribution and reactivity distribution, a target reactivity distribution is solved from a desired power distribution through the inverse analysis. The reactivity of available assemblies is evaluated using a modified effective fast-group model with a higher order correction of leakage effect. A candidate LP is determined by individually matching the reactivity of available assemblies to the target in-core reactivity distribution using a stable matching algorithm. The process of predicting the reactivity of available assemblies and matching them to the target reactivity distribution is repeated until the convergence condition is satisfied. Validation of the model has been confirmed with a reference PWR core. The model is expected to be applicable to a problem of multicycle loading pattern optimization.

The purpose of this study is to measure the photoneutron emission rate after the shutdown of an operational BWR. The photoneutrons originate from the D(γ, n)H reaction in the moderator region, while the high-energy gamma rays are generated from 140Ba-140La transient equilibrium of fission products in irradiated fuel. The photoneutron emission rate is measured by means of the photoneutron signal ratio of the Start-up Range Neutron Monitor (SRNM). The ratio is defined as a ratio of the photoneutrons to neutrons originating from spontaneous fission and the oxygen (γ, n) reaction of actinides (242Cm, 244Cm, etc.) in irradiated fuel. The principle of the measurement of the photoneutron signal ratio is the large difference between the decay constants of the 140Ba-140La transient equilibrium and those of the actinides. The measurement of the SRNM signal was continuously carried out over several months, and the photoneutron signal ratio was evaluated by using the least-squares method to fit a theoretical model to...

The temporal evolution of the distribution of the number of neutrons in a time-varying multiplying system, producing only prompt neutrons, was treated recently with the master equation technique by some of the present authors. Such a treatment gives account of both the so-called zero power reactor noise and the power reactor noise simultaneously. In particular, the first two moments of the neutron number, as well as the concept of criticality for time-varying systems, were investigated and discussed. The present paper extends these investigations to the case when delayed neutrons are also taken into account. Due to the complexity of the description, only the expectation of the neutron number is calculated. The concept of criticality of a time-varying system is also generalized to systems with delayed neutrons. The temporal behaviour of the expectation of the number of neutrons and its asymptotic properties are displayed and discussed.

The purpose of this study is to compare the measured and calculated ratios of neutrons originating from 242Cm and 244Cm in commercial BWRs in order to evaluate the prediction error for a weakening start-up range neutron monitor (SRNM) signal due to the decay of in-core neutron sources over a ten year period. Measurements were carried out over a period of several months, using the least-squares method to fit the time series of SRNM signals with the theoretical model. Neutron sources of each fuel assembly were calculated by the macroscopic depletion model of a plant process computer. The ratios of neutrons originating from 242Cm and 244Cm detected by each SRNM were calculated by whole core fixed source calculation using the nodal expansion method. One BWR core measurement was taken at the beginning of the cycle and three BWR core measurements were taken in the middle of the cycle. A comparison of the measured and calculated ratios of 242Cm and 244Cm number densities showed reasonable agreement. The slight d...

The third-order neutron correlation technique is one of the subcriticality measurement techniques. The significant feature of this technique is that an absolute value of subcriticality can be evaluated from the measured neutron correlation factors, namely Y∞ and X∞. In order to extend the applicability of the third-order neutron correlation technique to actual complicated subcritical systems, we previously derived the spatial and neutron energy dependent theoretical formulas of Y∞ and X∞ by using the first-, second-, and third-order importance functions. These theoretical formulas depend on the locations of both an external neutron source and a neutron detector. The numerical solutions of these formulas are very useful to design the arrangements of the external neutron source and the neutron detector and to investigate the contamination of the spatial effect in the measured neutron correlation factors, Y∞ and X∞. Therefore, we developed a novel deterministic code that can calculate both Y∞ and X∞ based on the SN method. In this paper, we focus on the description of the numerical procedure of this code and its verification by comparing it with the results of Monte Carlo simulations.

A fast reactor cycle scheme that incorporates a thoria-based minor actinide-containing cermet fuel is given. The present cermet fuel consists of an oxide solid solution of Th and minor actinides and Mo-inert matrix. It has been proposed as a high-performance device that can enhance minor actinide incineration in a fast reactor cycle. It is used in an independent small sub-cycle, whereby dedicated cycle technologies are adopted. Two-step reprocessing process was proposed for the present cermet fuel; it consists of a pre-removal of Mo-inert matrix and an actinide recovery. A preliminary test for the pre-removal of Mo-inert matrix was carried out using a surrogate cermet fuel. Burnup characteristics of a fast reactor core loaded with the cermet fuel were investigated by using neutronic calculation codes. It was revealed that a heterogeneous composition of Mo-inert matrix in inner and outer cores may lead to an effective transmutation of minor actinides and a flattened power density. It was concluded that the present cermet fuel was potentially promising as a high-performance incineration device of minor actinides for fast reactors.

The validity of various simplified treatments of anisotropic scattering is investigated in typical LWR configurations that simulate PWR, BWR, and APWR. In addition to the explicit treatment of anisotropic scattering, the diagonal approximation of anisotropic scattering matrixes and the transport corrected cross section with the assumption of isotropic scattering are also tested. Calculation results indicate that the anisotropic scattering matrix of the P1 component would be explicitly treated to obtain accurate results in the present calculations. The P2 or higher order anisotropic scattering matrix could be approximated by the diagonal approximation while maintaining the calculation accuracy. In addition to the above investigation, the physical meaning of the transport corrected cross section used in a transport calculation is discussed through a simple one-group fixed-source benchmark problem with anisotropic scattering. Though the transport corrected cross section is usually used to consider the P1 sca...

The zero-power reactor noise theory in a steady neutron multiplying medium was extended recently to a medium randomly varying in time to bridge the fields of the zero-power and the power reactor noise. For a time-varying medium in which the transition probability randomly fluctuates, only the use of the probability generating function technique based on the forward master equation approach is practical. However, with the forward master equation approach, the treatment of the joint moments of the neutron number and the medium state leads to a closure problem. Recently, the capability of the moment calculation technique for such cases was extended such that the closure problem could be solved exactly. The present paper describes and demonstrates this closure-free moment calculation technique in a time-varying binary multiplying medium, in which the medium state has two possible realizations. In addition to the first two moments of the neutron number N alone (irrespective of the medium state η), the joint moments of Nn and ηm, i.e., 〈Nnηm〉, were also obtained in a compact form for n = 1, 2 and arbitrary values of m, without a closure assumption. It was found that, for even m values, the asymptotic values of Nn and ηm are uncorrelated, whereas, for odd m values, they are negatively correlated, namely, their covariance is less than zero. The first two moments of the neutron number theoretically obtained were verified by the Monte Carlo technique. A perfect agreement was found between the Monte Carlo and the theoretical solutions. The closure-free moment calculation technique demonstrated in the present paper is expected to be applicable to various other problems related to the birth-and-death process with fluctuations of the transition probability, in which a closure problem occurs.

A practical method is proposed for retrospective estimation of the spatial dose distribution and the number of fissions in a typical criticality accident where there is a single radiation source in an accessible room. In this method, two kinds of low-fading and tissue-equivalent dosimeters are used as area dosimeters: an alanine dosimeter and a thermoluminescence dosimeter made of enriched lithium tetra borate (7Li2 11 B4O7). The procedure of the method consists of four parts: (1) dosimetry using area dosimeters (2) search for the center of the radiation source, (3) estimation of the spatial dose distribution in the room, and (4) estimation of the number of fissions in the accident. Dosimetry of criticality accident situations simulated at the TRACY facility demonstrates the practicability of the method. Although a post-hoc analysis in principle, this method gives useful information on the magnitude and hazard level of the accident to determine the strategy of radiation emergency medicine and other post-a...

In order to verify the Feynman-alpha formula with stationary multiple emission sources [Pazsit, I., Yamane, Y., 1998. Nuclear Instruments and Methods A403, 431–441], a series of experiments were performed by using three different neutron sources, i.e., an Am–Be, a 252 Cf, and a randomly triggered D–T pulsed neutron generator. As a result, it was found that the prompt neutron decay constant evaluated from the measured Y -function was hardly affected by the difference of neutron sources that have respective source multiplicities. On the other hand, it was demonstrated that the amplitude of Y -function, i.e., the correlation amplitude, depended on both the subcriticality and the neutron source multiplicity. These experimental facts agreed with the theoretical prospects that were deduced from the Feynman-alpha formula with multiple emission sources. Therefore, it was concluded that the validity of this formula was experimentally shown.

The third order neutron correlation technique is one of the subcriticality measurement techniques. This technique was originally proposed by Furuhashi and utilizes the second and third order neutron correlation factors Y(T) and X(T), which are evaluated from the variance and third order central moment of neutron counts detected during a counting gate width T, respectively. We can obtain the absolute value of subcriticality from the Y∞ and X∞, which are saturation values of Y(T) and X(T) when T goes to infinity. In the previous paper, we derived the generalized theoretical formulas of Y∞ and X∞ that took account of both the spatial and neutron energy effects. Its derivation was based on a heuristic method in which we consider the branching process of neutron family, and we utilized the α-eigenfunction expansion technique. Then, the previous formulas are expressed by multiple sums involving both the α-eigenfunctions and their adjoint functions. In this paper, we derive the new compact expressions of Y∞ and X∞ by using the importance functions related to the neutron detection process. Present derivation is based on the stochastic equation of neutron transport. New expressions have two advantages. We can clarify the physical meanings of Y∞ and X∞. And, without α-eigenfunction expansion, we can calculate Y∞ and X∞ directly by calculating the importance functions which satisfy the adjoint neutron transport equations. Moreover, we prove that new expressions of Y∞ and X∞ are identical with the previous ones, so that we can conclude that our previous derivation based on a heuristic method is correct.

In this paper, new optimization algorithms for the in-core fuel shuffling sequence of a boiling water reactor (BWR) are proposed to reduce outage time. During the short outage of a BWR, fuel shuffling can be a critical path in the periodic overall plant inspection. Therefore, a reduction in operation time for in-core fuel shuffling is essential to improve the plant capacity factor. For BWR in-core fuel shuffling, the shuffling sequence should be selected carefully since a fuel shuffling operation may affect those following it. Furthermore, several constraints must be satisfied during the in-core fuel shuffling of a BWR; e.g., two fuel assemblies must be inserted diagonally in a cell to fix the position of a control blade in it. Therefore, it is difficult to optimize BWR in-core fuel shuffling. In order to resolve this issue, new optimization methods are proposed, and the performances of some optimization algorithms are compared. Test calculations in actual BWR plants reveal that the workload for in-core fuel shuffling can be reduced by the proposed methods. The results of this paper will contribute to increasing the plant capacity factor by reducing the outage time.

We have studied measurement techniques of subcriticality by using the neutron correlation techniques. Among various techniques, we have paid attention to the third order neutron correlation technique, which was proposed by Furuhashi originally and utilizes the third order fluctuations of neutron counts. By using this technique, we can obtain the absolute value of subcriticality with the aid of factorial moments of fission neutron ν and external source neutron q , without prior knowledge of the prompt neutron decay constant at a reference state. However, in order to apply the third order neutron correlation technique to actual experiments, we must consider the spatial and neutron energy effects in this technique. For this purpose, we derived more generalized theoretical formulas of the third order neutron correlation technique that took account of the spatial and neutron energy effects by virtue of the α -eigenfunction expansion. Making use of the newly derived formulas, we propose the practical formulas and the experimental procedures based on the fundamental mode approximation.

A novel concept for the americium-containing target, which consists of (Th,Am)O2-x and molybdenum support material, is proposed that can fulfill several requirements such as reduction of the environmental burden and effective use of natural resources. In particular, deployment of recovered molybdenum from the spent nuclear fuel is highlighted and its nuclear effects are focused on in comparison to other types of molybdenum, namely natural and enriched ones. A preliminary investigation for effects of molybdenum isotopic composition on fast reactor core characteristics were carried out using nuclear calculation codes. The results indicated the feasibility of the fast reactor core loaded with the present target by optimizing its amount and composition. Moreover, several additional gains were indicated such as possible use of the radioactive waste and production of a new energy source, 233U.

For a study on the applicability of a personal dosimetry method to criticality accident dosimetry, an assessment of the human body surface and internal dose estimations was performed by experimental and computational simulations. The experimental simulation was carried out in a criticality accident situation created at the TRACY facility. The neutron and gamma-ray absorbed doses in muscle tissue were separately estimated by a dosimeter set of an alanine dosimeter and a thermoluminescence dosimeter made of enriched lithium tetra borate with a phantom. The computational simulation was conducted with a Monte Carlo code taking account of dose components of neutrons, prompt gamma-rays and delayed gamma-rays. Prior to the assessment of the body surface and internal dose estimations, the computational simulation was ascertained to be valid by comparison between the calculated dose distributions in the phantom and the measured ones. The assessment of the body surface and internal dose estimations based on the exp...

The Rossi-alpha formula for subcritical systems driven by a pulsed neutron source was derived with a solution technique that can effectively take different shape functions of the pulsed neutron source into account. The result of derivation showed that the present formula consists of the correlated and uncorrelated terms and an oscillating term whose period was identical to that of the pulsed neutron source. To investigate the validity of the pulsed Rossi-alpha formula and its applicability for the evaluation of the prompt neutron decay constant alpha from measurements, an experiment was performed at the Kyoto University Critical Assembly with a pulsed D-T neutron source. Although it seemed that fitting the present formula to the experimental curves was not practical because of an infinite-series expansion structure of the oscillating term, it was possible to extract only the correlated term, that explicitly includes the alpha value, by a simple data handling procedure. It was found that the measured alpha values agreed with those obtained by the pulsed neutron source method very well. This finding means that the pulsed Rossi-alpha method will be an attractive candidate for the monitoring method of the subcritical reactivity of future accelerator driven systems.

In previous works, the authors have developed an effective solution technique for calculating the pulsed Feynman- and Rossi-alpha formulae. Through derivation of these formulae, it was shown that the technique can easily handle various pulse shapes of the pulsed neutron source. Furthermore, it was also shown that both the deterministic (i.e.{} synchronizing with the pulsing of neutron source) and stochastic (non-synchronizing) Feynman-alpha formulae can be obtained with this solution technique. However, for mathematical simplicity and the sake of insight, the formal derivation was performed in a model without delayed neutrons. In this paper, to demonstrate the robustness of the technique, the pulsed Feynman- and Rossi-alpha formulae were re-derived by taking one group of delayed neutrons into account. The results show that the advantages of this technique are retained even by inclusion of the delayed neutrons. Compact explicit formulae are derived for the Feynman- and Rossi-alpha methods for various pulse shapes and pulsing methods.

In this paper, improvement of the SPH method (the improved SPH method) is proposed. The SPH method is commonly used in pin-by-pin mesh core calculations to reduce cell-homogenization error. The investigation revealed that the normalization condition of the SPH factor in the conventional SPH method is not appropriate for multi-assembly calculations in which different assembly types are adjacent. Since the conventional normalization condition does not incorporate flux discontinuity between assemblies, cell homogenization error in assembly peripheral region becomes larger. In the improved SPH method, the SPH factor is divided by an averaged “cell-level” discontinuity factor obtained in each fuel assembly. Though the SPH factor is somewhat modified from the conventional value, no additional homogenization parameters (e.g. discontinuity factor) is necessary in core calculations. Test calculations were carried out in a simplified one-dimensional slab and two-dimensional PWR colorset geometries that simulate par...

In this paper, simultaneous optimization is carried out for successive two cycles of pressurized water reactors. At first, a simplified problem of the simultaneous optimization was studied by assuming the batch-wise power sharing as independent variable, i.e., batch-wise power sharing was optimized without considering corresponding loading patterns. The optimization of the batch-wise power sharing was carried out for the conventional single cycle, the equilibrium cycle and the two successive (tandem) cycles. The analysis indicated that the tandem cycle optimization well reproduce that of the equilibrium cycle optimization, which is considered as a typical case of the true multicycle optimization. Next, simultaneous optimization of loading patterns for tandem cycles is carried out using the simulated annealing method. Since the design space of the tandem cycles optimization is much larger than that of the conventional single cycle optimization, the optimization condition (i.e., number of calculated pattern...

In this paper, thermal expansion effect on neutronics characteristics is approximately taken into account by a correction on cross sections. Dimensions of reactor core components depend on their temperatures due to the thermal expansion phenomena. However, the variation of calculation geometry requires considerable computational load for trajectory based lattice transport calculations such as the characteristics method since ray tracing must be re-executed. Therefore, if a correction on cross sections can accurately capture the effect of geometrical variation due to the thermal expansion, computation time of a lattice transport calculation that treat temperature variation can be reduced. Three different corrections on cross sections were tested in PWR fuel assembly geometry using UO2 and MOX fuels. It was found that the correction of cross sections that preserves neutron attenuation in a region almost reproduce the reference calculation that explicitly considers geometrical variation due to the thermal ex...

In this paper, various approximations of the exponential function are tested from the viewpoint of computational efficiency, i.e. accuracy and computation time. Since considerable part of computation time is devoted to the evaluation of the exponential function in transport calculations using the characteristics method, reduction of its computation time is important. Evaluations of the approximations are carried out on a Linux operating system using two different processors, i.e. Xeon 2.8GHz with 2MB L3 cache and Pentium-III 800 MHz with 512 KB L2 cache. In order to grasp the impact of approximation error of the exponential function on transport calculation results, some sensitivity calculations are executed using the characteristics method. The evaluation results indicate that the table lookup method with linear interpolation or second-order interpolation is computationally efficient for practical calculations.

In this paper, performances of spatial homogenization methods for fuel or non-fuel cells are compared in slab geometry in order to facilitate pin-by-pin core calculations. Since the spatial homogenization methods were mainly developed for fuel assemblies, systematic study of their performance for the cell-level homogenization has not been carried out. Importance of cell-level homogenization is recently increasing since the pin-by-pin mesh core calculation in actual three-dimensional geometry, which is less approximate approach than current advanced nodal method, is getting feasible. Four homogenization methods were investigated in this paper; the flux-volume weighting, the generalized equivalence theory, the superhomogenization (SPH) method and the nonlinear iteration method. The last one, the nonlinear iteration method, was tested as the homogenization method for the first time. The calculations were carried out in simplified colorset assembly configurations of PWR, which are simulated by slab geometries, and homogenization performances were evaluated through comparison with the reference cell-heterogeneous calculations. The calculation results revealed that the generalized equivalence theory showed best performance. Though the nonlinear iteration method can significantly reduce homogenization error, its performance was not as good as that of the generalized equivalence theory. Through comparison of the results obtained by the generalized equivalence theory and the superhomogenization method, important byproduct was obtained; deficiency of the current superhomogenization method, which could be improved by incorporating the “cell-level discontinuity factor between assemblies”, was clarified.

A series of experiments was carried out by using the Kyoto University Critical Assembly and a pulsed neutron generator attached. The experiments were intended to investigate a variance-to-mean technique for measuring the prompt neutron decay constant α of the subcritical reactor system that was driven by the periodic and pulsed neutron source. As the results, it was confirmed that the α value measured by the present technique agreed with a reference value evaluated by the pulsed neutron technique. Furthermore, it was demonstrated that the present technique was available to measure the α value even when the product of α and the repetition period of pulsed neutrons was smaller than unity. Since measurement of α by the pulsed neutron technique is generally difficult under such conditions, it was concluded that the present technique was convenient for measuring the α value.

Derivation of a variance-to-mean formula was performed to obtain the prompt neutron decay constant α of the subcritical reactor system that is driven by the periodic and pulsed neutron source. The formula was derived for an experimental technique where the Y value is measured by choosing an injection of pulsed neutrons as the origin of the counting gate. By using the derived formula, it was indicated that the α value can be easily determined. From this fact, it was concluded that the present formula was useful to measure the α value for future experiments with the accelerator driven system.

Recent studies of various cancers, such as those of the breast, head and neck, bladder and lung, reported that 46-64% of somatic mutations in the D-loop region of mitochondrial DNA (mtDNA) are observed. However, in esophageal cancer, only a low rate (5%) of somatic mutations has so far been reported in one article (Hibi, K. et al., Int J Cancer 2001;92:319-321). Thus, to confirm this we analyzed the somatic mutations for hypervariable regions (HVR-I and HVR-II) in the D-loop of mtDNA to reevaluate the possibility of mitochondrial genetic instability in this cancer. We amplified both HVRs by PCR and DNA samples obtained from 38 esophageal tumors and matched normal tissues, and then sequenced them. Comparing the sequences of tumors to those of normal tissues, we found 14 somatic mutations in 13 patients (34.2%). Eleven mutations were at the C consecutive stretch from position 303 to 309 of MITOMAP in the mitochondria databank (http://www.mitomap.org/), 1 at position 215 in HVR-II and 2 at positions 16,304 and 16,324 in HVR-I. There were 41 types of germ line variations in HVR-I including 2 not so far recorded in the mtDNA databank and 17 in HVR-II including 1 not yet recorded. We also determined nuclear genome instability of these 38 specimens by analyzing 3 independent microsatellite sequences. While 4 specimens showed a single microsatellite change, which is tumor specific, we did not find any co-relation between a somatic mtDNA mutation and microsatellite instability of nuclear genome DNA. These results suggest that mtDNA mutations might show a genetic instability in esophageal cancer independently from a nuclear genome instability.

A new formula of flux tilt in two-energy-group is defined and a higher-mode and higher-order formulation of flux tilt in two-energy-group is successfully derived on the basis of the modified explicit higher order perturbation (MHP) method, and a relationship between the flux tilt in two-energy-group and the first-mode eigenvalue separation is obtained. Then, it was found that the flux tilt in two-energy-group is proportional to the perturbed reactivity and inversely proportional to the first-mode eigenvalue separation. To show the validity of the relationship, numerical simulations are performed with two-energy-group for a decoupling core at the Kyoto University Critical Assembly (KUCA). Through the numerical simulations for inferring the eigenvalue separation from the perturbed reactivity and the degree of flux tilt induced by a perturbation, it was shown that the validity of the relationship is confirmed and that the flux tilt is independent of the neutron energy. Moreover, by considering the higher-mode and the higher-order terms in the new formula, it was found that the effects of the higher-mode and the higher-order components can be clearly observed in the decoupling core.

High-temperature effects on Photosystem II and plasma membranes, temperature dependence of growth, and acclimation to the growth temperature were studied in a mesophilic cyanobacterium, Synechocystis sp. PCC6803. The following results were obtained. (1) Small but distinct temperature acclimation of the PSII reaction center activity was shown for the first time when the activity was measured at inhibitory high temperatures. However, the reaction center activity showed no apparent acclimation when it was measured at growth temperatures after heat stress. (2) Oxygen-evolving activity and the permeability of plasma membranes showed higher resistance to heat when PCC6803 cells were grown at higher temperatures. (3) Acclimation of photosynthesis to the growth temperature seemed to occur so as to maintain photosynthesis activity not at a maximum level but in a certain range at the growth temperatures. (4) Neither sensitivity to high-temperature-induced dissociation of phycobilisomes from the PSII reaction center complexes nor degradation of phycocyanin were altered by changes in the environmental temperature. (5) A close relationship between the viability of cells and the structural changes of plasma membranes (but not the inactivation of photosynthesis) was observed. The denaturation process of PSII complexes and the relationship between the temperature dependence of the growth of Synechocystis PCC6803 cells and that of the photosynthetic activity are also discussed.

By introducing two perturbed systems from a homogeneous core the higher order perturbation calculation of a heterogeneous core was performed to obtain the neutron flux and reactivity change with the higher mode eigenfunctions in the homogeneous core system which is considered to be an unperturbed system. The main feature of this perturbation calculation is to express analytically the perturbed flux with the trigonometric functions for the purpose of saving a computational time to obtain numerically the higher mode eigenfunctions in the unperturbed system beforehand the perturbation calculation. Then the two higher order perturbation calculations were executed by the Modified Explicit Higher Order Perturbation (MHP) method with two energy groups in a two-dimensional x - y geometry for the two perturbed systems. The objective reactivity was obtained from the difference between the two calculated results of perturbation. For the heterogeneous bare system it was demonstrated that the neutron flux and reactivity change can be accurately evaluated by this perturbation calculation with large numbers of expansion modes and perturbation orders and that the computational time can be reduced economically. However for the perturbation causing the big difference in group constants from the unperturbed system including the change from a bare core with the reflector it was found that the higher order perturbation calculation encounters the difficulty which is closely related to that pointed out so far for the application of explicit higher order perturbation theory itself.

Development of critically safety monitoring is an important issue in nuclear fuel reprocessing plants. In this paper two subcriticality estimation methods are compared. One method is based on conventional artificial neural network method. But its performance was no good in the sense that the accuracy of estimated subcriticality was the same order as the results estimated by the conventional Modified Neutron Source Multiplication method. The other method considered is referred to as the “Pattern matching” method. It can be used as the critical safety monitor for non-uniform fuel solution systems in reprocessing plants. Numerical experiments were carried out to examine the method above mentioned; the estimated subcriticalities agreed well with the true values. The “Pattern matching” method can also be applied to the fuel concentration estimation problem. The estimated fuel distribution agreed reasonably with the true fuel distribution in all cases. It is concluded that the “Pattern matching” method is available critical safety monitor for non-uniform fuel systems in reprocessing plants.

Effects of high-temperature treatments on a thermophilic cyanobacterium, Synechococcus vulcanus, were studied, and the following results were obtained. (1) Oxygen evolution and the PSII photochemical reaction were the most sensitive sites and started to be inactivated at temperatures slightly higher than the cultivating temperature. (2) The decrease in the fluorescence Fv value reflected the inactivation of the charge separation reaction of PSII as well as that of the oxygen evolution reaction. (3) The dark fluorescence level, Fo, showed an increase at around 70 degrees C, which was partially reversed by further incubation at 50 degrees C. This increase reflected the inactivation of PSII reaction centers and probably dissociation of phycobilisomes from the PSII reaction center complexes. (4) At higher temperatures, phycobiliproteins disassembled and denatured in a pH-dependent manner, causing a large Fo decrease. (5) Cell membranes became leaky to low-molecular-weight substances at around 72 degrees C. (6) Inhibition of growth of the cells was recognized when the cells were pretreated at temperatures higher than 72 degrees C. Reversibility of the high-temperature effects and relationship between viability of the cells and the degradation of the cell membranes are discussed.

A PWR core analysis was performed by using the Modified Explicit Higher-Order Perturbation (MHP) method. A new perturbation calculation code with two energy groups in a two-dimensional x-y geometry based on the MHP method TWOPEC was developed to evaluate the change both in the neutron flux distribution and the reactivity caused by a perturbation. The results were compared with those by the direct eigenvalue calculations to examine the capability of the MHP method. It was demonstrated that both the neutron flux distribution and the reactivity can be accurately evaluated by the MHP method for a string and local perturbation caused by the control rod insertion when one uses large numbers of expansion modes and perturbation orders. The MHP method was applied to the pseudo burnup calculations that regarded the change in cross sections caused by the fuel burnup as the perturbation. It was found that one can also obtain satisfactory results not only for the burnup reactivity but also for the control rod worth at each burnup step.

Increases in the chlorophyll fluorescence Fo (dark level fluorescence) during heat treatments were studied in various higher plants. Besides the dissociation of light-harvesting chlorophyll a/b protein complexes from the reaction center complex of PS II and inactivation of PS II, dark reduction of QA via plastoquinone (PQ) seemed to be related to the Fo increase at high temperatures. In potato leaves or green tobacco cultured cells, a part of the Fo increase was quenched by light, reflecting light-induced oxidation of QA- which had been reduced in the dark at high temperatures. Appearance of the Fo increase due to QA reduction depended on the plant species, and the mechanisms for this are proposed. The reductants seemed to be already present and formed by very brief illumination of the leaves at high temperatures. A ndhB-less mutant of tobacco showed that complex I type NAD(P)H dehydrogenase is not involved in the heat-induced reduction of QA. Quite strong inhibition of the QA reduction by diphenyleneiodonium suggests that a flavoenzyme is one of the electron mediator to PQ from the reductant in the stroma. Reversibility of the heat-induced QA reduction suggests that an enzyme(s) involved is activated at high temperatures and mostly returns to an inactive form at room temperature (25 °C).

The Feynman- and Rossi-alpha formulas are calculated for subcritical systems driven by a multiple emission source, i.e., one that emits several neutrons on each source emission event. The prime exa...

The spatial dependence of thermal and fast neutron flux was measured axially in the core of a 1 MW research reactor. The measurements were made by a thin optical fibre detector with a neutron sensitive ZnS(Ag) scintillation tip. For thermal neutrons 6 Li was used, whereas for fast neutrons 232 Th was used as neutron converter. The spatial dependence was measured by moving the fibre axially with a uniform speed. The measurement takes a few minutes, compared to up to 10 h with the conventional wire activation method. Comparison with traditional measurements shows a good agreement.

The Explicit High-Order Perturbation method is among various perturbation methods and its main feature is to expand neutron flux of a perturbed system in a series of the λ-mode eigenfunctions in an unperturbed system. Although various perturbation methods using λ-mode eigenfunctions have been proposed in the past, it was difficult to calculate accurately the perturbed flux and the reactivity with multi energy groups. To overcome this problem we derived a new perturbation formulation unlike the Flux Reconstructed Explicitly by Neutronic Calculated Harmonics method based on the Explicit Higher-Order Perturbation Method and applied it to a simplified core system. To show the validity of this new perturbation method numerical calculations with two energy groups in an x–y geometry were carried out and it was found that the perturbed flux and the reactivity could be accurately obtained by this new perturbation method.

As the part of the international benchmark experiment of the effective delayed neutron fraction held in the Fast Critical Assembly of the Japan Atomic Energy Research Institute, we measured the effective beta for the FCA XIX-1 (Uranium) core, XIX-2 (Plutonium/Natural Uranium) core, and XIX-3(Plutonium) core by using a modified Bennett method. This method was proposed by these authors to suppress scatter of covariance data. Final results of effective beta were 782 ± 16 pcm for the U-core, 368 ± 6 pcm for the Pu/NU-core, and 256 ± 4 pcm for the Pu-core. The ratio of calculated to measured effective beta values were 0.990, 1.022, and 0.980 for U, Pu/NU, and Pu cores, respectively.

The conventional variance-to-mean method (Feynman-α method) seriously suffers the divergency of the variance under such a transient condition as a reactor power drift. Strictly speaking, then, the use of the Feynman-α is restricted to a steady state. To apply the method to more practical uses, it is desirable to overcome this kind of difficulty. For this purpose, we propose an usage of higher-order difference filter technique to reduce the effect of the reactor power drift, and derive several new formulae taking account of the filtering. The capability of the formulae proposed was demonstrated through experiments in the Kyoto University Critical Assembly. The experimental results indicate that the divergency of the variance can be effectively suppressed by the filtering technique, and that the higher-order filter becomes necessary with increasing variation rate in power.

Eigenvalue separation is an important index of power reactor stability in the view of neutronics. In recent years, a long, slender Position Sensitive Proportional Counter has been developed. By using this proportional counter, the spatial neutron flux distribution of a nuclear reactor can be measured rapidly. Using these characteristics, we intend to develop the experimental method evaluating eigenvalue separation as an index for reactor stability. A method based on spatial integrals of neutron flux measured at a subcritical system with an external source is proposed. To verify this method, experiments were carried out at Kyoto University Critical Assembly. It was confirmed that eigenvalue separation can be evaluated within a relative error of 10%.

Activities of oxygen evolution, fluorescence Fv (a variable part of chlorophyll fluorescence) values, and amounts of the 33 kDa protein remaining bound to the thylakoids in intact spinach chloroplasts were measured during and after high-temperature treatment. The following results were obtained. (1) Both the Fv value and the flash-induced oxygen evolution measured by an oxygen electrode were decreased at high temperatures, but they showed partial recovery when the samples were cooled down and incubated at 25°C for 5 min after high-temperature treatment. (2) Oxygen evolution was more sensitive to high temperatures than the Fv value, and the decrease in the Fv/Fm ratio at high temperatures rather corresponded to that in the oxygen evolution measured at 25°C after high-temperature treatment. (3) Photoinactivation of PS II was very rapid at high temperatures, and this seems to be a cause of the difference between the Fv values and the oxygen-evolving activities at high temperatures. (4) At around 40°C, the manganese-stabilizing 33 kDa protein of PS II was supposed to be released from the PS II core complexes during heat treatment and to rebind to the complexes when the samples were cooled down to 25°C. (5) At higher temperatures, the charge separation reaction of PS II was inactivated, and the PS II complexes became less fluorescent, which was recovered partially at 25°C. (6) Increases in the Fv value due to a large decrease in the electron flow from QA to QB became prominent after high-temperature treatment at around 50°C. This was the main cause of the discrepancy between the Fv values and the oxygen-evolving activities measured at 25°C. Relationship between the process of heat inactivation of PS II reaction center complexes and the fluorescence levels is discussed.

We intend to develop a subcritically monitor based on the Feynman-α method, which detects a change in the prompt neutron decay constant caused by a change in the reactivity of a system containing nuclear fuel materials. For the practical use of this method, quick data processing is necessary to facilitate a real time measurement and it needs to be applicable to a wide range of count rates in the neutron detector. The Feynman-α method, based on pulse counting, has a serious problem relating to the counting statistics, namely the count loss effect due to the dead time of counter itself and the following counting circuit. To eliminate the count loss effect, especially in a high count rate region, a quick correction technique should be established to ensure accurate measurement. As a candidate for this, we have developed the data-synthesis technique which processes synthesized data generated by adding multiple time-series data acquired by independent neutron detectors. The present paper provides the derivation of a formula for this data-synthesis technique and the result of a numerical examination on its applicability.