00085 Heterogeneous breeding blanket experiment withlithium titanate and beryllium

05 Nuclear fuels (scientific, technical) 06/00078 Design of an intelligent fuzzy logic controller for a nuclear research reactor

06•00082 Feynman-oz measurements on the fast critical zero-power reactor MASURCA

Adda, F. et al. Progress in Nuclear Energy, 2005, 46, (3 4), 328 347. The main objective of this paper is to design an intelligent controller system based on the concepts of fuzzy logic. This latter will be used to control the power of a nuclear reactor. The principle of this controller is based on rules established from experiments used with a classical controller and from the knowledge and the expertise of the operators of the reactor. This intelligent controller could be used in parallel with the actual system, which is semiautomatic, as a decision aided system to assist the operators in the control room.

Kloosterman, J. L. and Rugama, Y. Progress in Nuclear Energy, 2005, 46,(2), 111 125. Within the framework of the MUSE program, Feynman-cr measurements were performed on the critical fast reactor MASURCA. The pres and cons of two measurement methods are described. Because of the low count rate, measurements recording the time interval between neutron pulses have preference compared with measurements that record the neutron counts in adjacent time intervals. The latter method requires much more disk space while less information is stored. Independent of the measurement method, it turned out that each detected neutron gave a TTL pulse and an echo pulse with a delay of about 120 ns. A correction to the variance-to-mean ratio has been derived, and the experimental setup was tuned to avoid these echoes in later measurements. The measurement method that records the time interval between neutron counts can be used to artificially increase the detector dead time. In principle, such measurement data can be used to measure the dependence of the correlated part of the variance-to-mean ratio on dead time. However, in this case, it was only possible to extract the dependence of the uncorrelated part on dead time. A good agreement was found between the measurements and theory. The value for the prompt neutron decay constant as measured with two detectors in the reflector region of the core reach about 5780 ± 264 s t, which agrees quite well with the expected values for the delayed neutron fraction (~ ~ 0.33%) and the neutron generation time (~0.55 ~ts). Repeated measurements with more sensitive detectors at the same positions gave consistently a lower value for the one detector and a higher value for the other. An explanation for this is not yet found.

06/00079 Erosion, transport, and tritium codeposition analysis of a beryllium wall tokamak Brooks, J. N. el al. Fusicm Engineering and DesL¢,n, 2005, 72, (4), 363 375. The authors analysed beryllium first wall sputtering erosion, sputtered material transport, and T/Be codeposition for a typical next-generation tokamak design - the fusion ignition research experiment (FIRE). The results should be broadly applicable to any future tokamak with a beryllium first wall. Starting with a fluid code scrapeoff layer attached plasma solution, plasma D o neutral fluxes to the wall and divertor were obtained from the DEGAS2 neutral transport code. The D + ion flux to the wall was computed using both a diffusive term and a simple convective transport model. Sputtering coefficients for the beryllium wall were given by the VFTRIM-3D binary-collision code. Transport of beryllium to the divertor, plasma, and back to the wall was calculated with the WBC+ code, which tracks sputtered atom ionization and subsequent ion transport along the SOL magnetic field lines. Then, using results from a study of Be/W mixing/sputtering on the divertor, and using REDEP/WBC impurity transport code results, the divertor surface response was estimated. Finally, tritium codeposition rates were computed in Be growth regions on the wall and divertor for D - T plasma shots using surface temperature dependent D-T/Be rates and with different assumed oxygen contents. Key results were: (1) peak wall net erosion rates varied from about 0.3 n m s ~ for diffusion-only transport to 3 n m s for diffusion plus convection (2) T/Be codeposition rates varied from about 0.1 to 10.0 mg T s ~ depending on the model, and (3) core plasma contamination from wall-sputtered beryllium was low in all cases (<0.02%). Thus, based on the erosion and codeposition results, the performance of a beryllium first wall is very dependent on the plasma response, and varies from acceptable to unacceptable.

06•00080 Evaluation and optimization of General Atomics' GT-MHR reactor cavity cooling system using an axiomatic design approach Thielman, J. et al. Nuclear Engineering and Design, 2005, 235, (13), 1389 1402. The development of the Generation IV (Gen-IV) nuclear reactors has presented social, technical, and economical challenges to nuclear engineering design and research. To develop a robust, reliable nuclear reactor system with minimal environmental impact and cost, modularity has been gradually accepted as a key concept in designing highquality nuclear reactor systems. While the establishment and reliability of a nuclear power plant is largely facilitated by the instalment of standardized base units, the realization of modularity at the subsystem/sub-unit level in a base unit is still highly heuristic, and lacks consistent, quantifiable measures. In this work, an axiomatic design approach is developed to evaluate and optimize the reactor cavity cooling system (RCCS) of General Atomics' Gas Turbine-Modular Helium Reactor (GT-MHR) nuclear reactor, for the purpose of constructing a quantitative tool that is applicable to Gen-IV systems. According to Suh's axiomatic design theory, modularity is consistently represented by functional independence through the design process. Both qualitative and quantitative measures are developed here to evaluate the modularity of the current RCCS design. Optimization techniques are also used to improve the modularity at both conceptual and parametric level. The preliminary results of this study have demonstrated that the axiomatic design approach has great potential in enhancing modular design, and generating more robust, safer, and less expensive nuclear reactor sub-units.

06•00081 Experiences and applications of PEANO for online monitoring in power plants Fantoni, P. F. Progress in Nuclear Energy, 2005, 46, (3~4), 206 225. This paper describes the motivations, ideas, work and lesson learned that resulted in the development of the software product called PEANO, for signal validation of process sensors in a dynamic system. The focus of this report is more on the process that started in 1994 and brought to the realization of the current system in 2003 (PEANO ver. 4.1).

06/00083 Flashing-induced density wave oscillations in a natural circulation BWR-mechanism of instability and stability map Furuya, M. et al. Nuclear Engineering and Design, 2005, 235, (14), 1557 1569. Experiments were conducted to investigate two-phase flow instabilities due to flashing in a boiling natural circulation loop with a chimney at low pressure. The SIRIUS-N facility was designed to have nondimensional values nearly equal to those of typical natural circulation boiling water reactor (BWR). The observed instability is suggested to be flashing-induced density wave oscillations, since the oscillation period correlated well with the passing time of single-phase liquid in the chimney section regardless of system pressure, heat flux, and inlet subcooling. Stability maps were obtained in reference to the inlet subcooling and the heat flux at the system pressures of 0.1, 0.2, 0.35, and 0.5 MPa. The flow became stable below a certain heat flux regardless of the channel inlet subcooling. The stable region enlarged with increasing system pressure. Thus, the stability margin becomes larger in a startup process of a reactor by pressurizing the reactor sufficiently before heating according to the stability map.

06/00084 Helium gas permeability of SiC/SiC composite used for in-vessel components of nuclear fusion reactor Hino, T. eZ al. Fusion Engineering and Design, 2005, 73, (1), 51 56. SiC/SiC composite is a candidate material of blanket in fusion reactors since silicon carbide is a low activation material and the temperature of helium coolant can be taken high, 1100 K. Helium gas permeability of SiC/SiC composite, however, has been little measured so far for this purpose. For SiC/SiC composites made by several methods, the helium gas permeability was measured using a vacuum system consisting of high and low-pressure chambers, and its application for the blanket was discussed. The permeability of SiC/SiC composite made by the method called as nano-powder infiltration and transient eutectoid (NITE) process was observed to be very low, 10-11 m2/s. A change of the permeability owing to heat cycles was also measured. The increase of the permeability was observed to be not large. The leak flux of helium through the SiC/SiC coolant pipe and first wall of blanket module was analysed based upon the density balance of helium. It is shown that the leak rate into fusion plasma can be taken lower the helium production rate due to fusion reactions if the pumping is attached to the blanket module. Then, the fuel dilution due to the helium leak may be avoided even when the SiC/SiC blanket modules are employed.

06/00085 Heterogeneous breeding blanket experiment with lithium titanate and beryllium Klix, A. el al. Fusim~ Engineering and DesL~n, 2005, 72, (4), 327 337. An integral breeding blanket experiment has been done at Fusion Neutronics Source of J A E R I with an assembly of three layers of lithium titanate, beryllium, and low activation ferritic steel F82H. The experimental assembly and the neutron source were enclosed in a SS316 stainless steel can and irradiated with D - T neutrons. The tritium production rate was measured with small Li2CO3 pellet detectors. The tritium production from 7Li was estimated by means of the reactions -32 S(n, p)-32 P and-35 Cl(n, n~,)-32 P which have effective cross-sections similar to 7Li(n, n~)T but are more sensitive methods. The neutron flux

Fuel and Energy Abstracts

January 2006

13

05 Nuclear fuels (scientific, technical) in the assembly was checked with activation foils. The mea s ure me nt results were analysed with the three-dimensional Monte Carlo code MCNP-4C and three different beryllium cross-section evaluations taken from FENDL/MC-2.0, FENDL/E-2.0, and EFF-3.03. The library FENDL/MC-2.0 was also used for the other materials in the assembly. All three beryllium libraries model the fast flux in the assembly well. Differences were seen for indium foils. An overestimation was observed for the tritium production.

06•00086 Hybrid soliton nuclear reactors: a model and simulation (encapsulated long living accelerator driven system) Gaveau, B. et al. Nuclear Engineering and Design, 2005, 235, (15), 1665 1674. The purpose of this study was to explore the potential of Hybrid Soliton Reactors (Rdacteur Hybride h Soliton, RHYS) for producing energy. In this case an encapsulated long-living fission reactor is driven by a proton accelerator, who produces neutrons on a target. In a first part of the paper the mathematical approach of such a sub-critical reactor is given, as an extension of the 'Soliton Reactor' recently proposed by other studies. In the second part results of simulations are given and the possibilities to control such a system explored.

06•00087 Identification and localization of absorbers of variable strength in nuclear reactors Demazi~re, C. and Andhill, G. Annals o/Nuclear Energy, 2005, 32, (8), 812 842. This paper investigates the possibility of localizing a noise source of the type 'absorber of variable strength' (or reactor oscillator) from as few as five neutron detectors evenly distributed throughout the core of a commercial nuclear reactor. The novelty of this investigation lies with the fact that the calculations are performed for a realistic 2-D heterogeneous reactor in the 2-group diffusion approximation, via the prior determination of the corresponding reactor transfer function. It is first demonstrated that the response of such a reactor to a localized perturbation deviates significantly from point-kinetics. The space-dependence of the induced neutron noise thus carries enough information about the location of the noise source, which makes it possible to determine its position from a few detector readings. The identification of the type of noise source is easily performed from the in-phase behaviour of the induced neutron noise. Different unfolding techniques are finally tested. All these techniques rely on the use of the reactor transfer function. One of these techniques is based on the comparison between the actual measured neutron noise and the neutron noise calculated for every possible location of the noise source. This technique is very reliable and almost insensitive to the contamination of the detector signals by background noise, but also extremely CPU consuming. Another technique, based on the piecewise inversion of the reactor transfer function and requiring little CPU effort, was developed. Although this technique is much less reliable when background noise is present, this technique is useful to indicate a region of the reactor where a noise source is likely to be located.

analyses based on first-order perturbation theory calculations have been performed using the deterministic code E R A N O S (Version 2.0) in conjunction with its adjusted nuclear data library ERALIB-1. It is found that the M4SC2 configuration, independent of the external source, is quite representative of the different XADSs for actinide capture reactions at the centre of the fuel zone, relative to 2 3 9 p u fission at the same location. For the case of a threshold fission reaction, such as that in 239U, the sensitivity to the external source is significantly higher. With respect to the corresponding spectral index, M4SC2 with the D(d,n)He 3 source remains quite representative of the He- and Nacooled XADSs. For the system with Pb/Bi coolant, on the other hand, effects of uncertainties associated with the data for these two nuclides and their low content in the MUSE configuration result in significantly lower associated representativity factors. A better overall representativity of the Pb/Bi-cooled XADS is expected to be achieved by the new MUSE_Na/Pb configuration.

06•00090 Improved core design of the high temperature supercritical-pressure light water reactor Yamaji, A. et al. Annals el'Nuclear Energy, 2005, 32, (7), 651 670. A new coolant flow scheme has been devised to raise the average coolant core outlet t e mpe ra t ure of the High Temperature Supercritical Pressure Light Water Reactor (S C LWR H). A new equilibrium core is designed with this flow scheme to show the feasibility of an SCLWR-H core with an average coolant core outlet temperature of 530°C. In previous studies, the average coolant core outlet temperature was limited by the relatively low t e mpe ra t ure outlet coolant from the core periphery. In order to achieve an average coolant core outlet t e mpe ra t ure of 500°C, each fuel assembly had to be horizontally divided into four sub-assemblies by coolant flow separation plates, and coolant flow rate had to be adjusted for each sub-assembly by an inlet orifice. However, the difficulty of raising the outlet coolant temperature from the core periphery remained. In this study, a new coolant flow scheme is devised, in which the fuel assemblies loaded on the core periphery are cooled by a descending flow. The new flow scheme has eliminated the need for raising the outlet coolant t e mpe ratu r e from the core periphery and removed the coolant flow separation plates from the fuel assemblies.

06•00091 Investigation of efficient 1311 production from natural uranium at Tehran research reactor Khalafi, H. et al. Annals o/Nuclear Energy, 2005, 32, (7), 729 740. Iodine-131, which has a half-life of 8.05 days, is the one of the most widely used radionuclides in medical diagnosis and treats some diseases of thyroid gland. Optimization of 1- 1I production in Tehran research reactor (TR R ) was studied by two different methods. Primarily, standard nuclear codes such as O R I G E N , WIMS and C I T A T I O N were applied and then analytical solutions technique was followed. Calculated results and experimental works in the bench scale

06•00088 Immobilization of radioactive evaporator concentrate in mortar matrix

irradiation in the unique production line.

Plecas, I. B. and Dimovic, S. Progress in Nuclear Energy, 2005, 46, (2), 151 157. Traditional methods of processing evaporator concentrates from NPP are evaporation and cementation. These methods allow the transformation of a liquid radioactive waste into an inert form, suitable for a final disposal. To assess the safety for disposal of radioactive mortar-waste composition, the leaching of - C s from immobilized radioactive evaporator concentrate into a surrounding fluid has been studied. Leaching tests were carried out in accordance with a method recommended by IAEA. Determination of retardation factors, KF, and coefficients of distribution, kd, using a simplified mathematical model for analysing the migration of radionuclides, has been developed. The experiment achieved the lowest leaching values after 60 days in samples. Results presented in this paper are examples of results obtained in a 20-year mortar and concrete testing project, which will influence the design of the engineered trenches system for a future central Serbian radioactive waste disposal centre.

06•00092 Measurement of (n, n') reaction cross-sections of 7gBr, 9°Zr, 197Au and 2°7pb with pulsed d-D neutrons

06/00089 Importance of the MUSE experiments for emerging ADS concepts from the nuclear data viewpoint Plaschy, M. et al. Anna/s o/Nuclear Energy, 2005, 32, (8), 843 856. The current investigation, conducted in the general framework of the MUSE program ('MUltiplication avec une Source Externe'), considers the representativity of a specific configuration of its fourth phase (M4SC2), which is driven by an external D(d,n)He 3 or T(d,n)He 4 neutron source, with respect to current concepts of eXperimental Accelerator Driven Systems (XADSs) with gas (He), Na and Pb/Bi coolants. The study has been carried out from the nuclear data viewpoint, with the external source being accounted for in an appropriate manner. In this context, data sensitivity/uncertainty

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January 2006

Shimizu, T. et al. Annals of Nuclear Energy, 2005, 32, (9), 949 963. Activation cross-sections for the (n, n') reaction were m eas u r ed by means of the activation method at neutron energies of 3.1 and 2.54 MeV using a p u l s e d neutron beam. The target nuclei were 79 Br, 90 Zr, 197 Au, and 20TrPb whose half-lives were between 0.8 and 8 s. The value of the 9°Zr(n, n f) 9°mZr reaction was obtained for the first time. In order to confirm the pulsed neutron beam measuring method, the cross-section data of 79 Br and 197 Au were compared with previous data obtained using a pneumatic sample transport system. The results of this comparison were in agreement within the range of experimental error. The d-D neutrons were generated by bombarding a deuterated titanium target with a 350-keV d+-beam at the 80 ° beam line of the Fusion Neutronics Source (FNS) at the Japan Atomic Energy Research Institute. In order to obtain reliable activation cross-sections, careful attention was paid to correct the efficiency for a volume source, and the self-absorption of gamma rays in irradiated samples. The systematics of the (n, n') reaction at a neutron energy of 3.0 MeV, which can predict cross-section of (n, n') reaction with an accuracy of 50%, was proposed for the first time on the basis of the data.

06/00093 Modeling of flashing-induced instabilities in the start-up phase of natural-circulation BWRs using the twophase flow code FLOCAL Manera, A. et al. Nuclear Engineering and Design, 2005, 235, (14), 1517 1535. This paper reports on the modelling and simulation of flashing-induced instabilities in natural-circulation systems, with special emphasis on natural-circulation boiling water reactors (BWRs). For the modelling