Progress in “COE-INES”

In the year 2002 and 2003, the Japanese Ministry of Education, Culture, Sports, Science and Technology started the “ The 21st Century Center of Excellence (COE) Program”, which is planned to continue for 5 years. A program proposed by Tokyo Institute of Technology “Innovative Nuclear Energy Systems for Sustainable Development of the World” simply called as COE-INES was selected as only one program in nuclear engineering field. The program consists of four main activities: research, education, society and internationalism. The research will be performed on the innovative nuclear energy systems, which include innovative nuclear reactors and innovative fuel cycles. Both free thinking and overall vision are taken on the research, and stressed on education also. In the education, COE-INES Captainship Program is promoted by integrating research with education, and we will foster creative researchers and engineers. Society is also a very important issue for nuclear energy. We try to coevolve nuclear energy with society and to strive towards the fulfillment of SR as well as to research innovative nuclear energy systems. We believe these ideas are occupied by many scientists in various countries. Then we are promoting the international collaboration for research and education on innovative nuclear energy systems.     

Turbulent diffusion of heat between connected flow passages Part 2: Predictions using the “κ-” turbulence model

Mixing by turbulent diffusion and secondary flow between the parallel subchannels of ducts simulating smooth, bare rod bundles has been investigated both experimentally and analytically. This paper, part 1, outlines the problem and reports experimental results; part 2 deals with the computer prediction of turbulent flow in rod bundles.The experimental work was performed on a long wind tunnel which was designed to achieve an axially unchanging, fully developed temperature profile at the tunnel exit. This is believed to be the first substantial investigation of inter-subchannel mixing using this technique. Detailed measurements were made over a range of Reynolds numbers and in three configurations simulating pitch to diameter ratios of 1.833, 1.375 and 1.1.The results of the experimental work confirm the major findings of previous investigations, in particular that the inter-subchannel mixing rates are considerably higher than predicted by simple diffusion theory, and are relatively insensitive to variations in the gap width between the rods. Effective diffusivities through the gap appear to be strongly anisotropic and there is no evidence of secondary flows.     

Sodium cooled small fast long-life reactor “4S”

CRIEPI and Toshiba Corp. have been exploring to realize a small-sized nuclear reactor for the needs of dispersed energy source and multi-purpose reactor. A conceptual design of 4S (Super-Safe, Small and Simple) reactor is proposed to meet the following design requirements: (1) All temperature feedback reactivity coefficients including whole core sodium void reactivity are negative; (2) The core integrity is secured against all anticipated transient without reactor scram; (3) No emergency power nor active mitigating system is required; (4) The reactivity core lifetime is more than 10 years. The 4S reactor is a metallic fueled sodium cooled fast reactor. A target of an electrical output is 10–50 MW. A remarkable feature of 4S is that its reactivity is not controlled by neutron absorber rods but by neutron reflectors to cope with a long core lifetime and a negative coolant void reactivity.

This study includes a design consideration of 4S. Design discussions are mainly focused on various core designs to meet above requirements. A tall core active height is adopted to gain long core lifetime. An averaged fuel burn-up is tried to be increased up to 100 GWd/ton from a point of economic view. The reference 4S designs are 10 MWe (30 years core lifetime) and 50 MWe (10 years core lifetime).     

A new ray tracing code “ELECTRA”

A ray tracing code has been developed on the basis of the charge density method. This code can solve Laplace's equation and can integrate Lorentz's equation of motion in the two, axially symmetric three and three dimensional spaces. The calculation procedure used in the code is described. As an example of a three-dimensional field calculation, the distribution of the fringing field of an electrostatic quadrupole lens is derived, and the definite integrals which are necessary to calculate third-order aberrations are given. The features of the code are summarized.     

“CANDLE” burnup regime after LWR regime

CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) burnup strategy can derive many merits. From safety point of view, the change of excess reactivity along burnup is theoretically zero, and the core characteristics, such as power feedback coefficients and power peaking factor, are not changed along burnup. Application of this burnup strategy to neutron rich fast reactors makes excellent performances. Only natural or depleted uranium is required for the replacing fuels. About 40% of natural or depleted uranium undergoes fission without the conventional reprocessing and enrichment.

If the LWR produced energy of X Joules, the CANDLE reactor can produce about 50X Joules from the depleted uranium left at the enrichment facility for the LWR fuel. If we can say LWRs have produced energy sufficient for full 20 years, we can produce the energy for 1000 years by using the CANDLE reactors with depleted uranium. We need not mine any uranium ore, and do not need reprocessing facility. The burnup of spent fuel becomes 10 times. Therefore, the spent fuel amount per produced energy is also reduced to one-tenth.

The details of the scenario of CANDLE burnup regime after LWR regime will be presented at the symposium.     

Addendum to “an evaluation of alternative nuclear strategies”

This paper presents an assessment by Battelle-Columbus of the technology associated with several reactor concepts which may be considered “advanced” beyond existing LWR's in terms of improved natural resource utilization. The concepts chosen for evaluation and intercomparison are the HTGR, GCFR, MSBR and LWBR. Numerous conclusions may be reached from the study and some interesting trends can be observed. Of greatest significance is the fact that the strategies associated with alternative reactors/fuel cycles will not produce dramatic decreases in short-term fissile demand. A second major conclusion is that all of these advanced systems are considered capable of meeting applicable environmental requirements. A third conclusion is that there is no apparent technical reason for deletion of development efforts on any of these reactors, providing that a commercial interest, complete with significant commitment, is existent.     

Radioactivity confinement capability of “PARC” concept for the future reprocessing

Aiming at TRU waste arising reduction and economical competitiveness for the future reprocessing, is proposed an advanced process concept which is named PARC (Partitioning Conundrum Key) process. Enhancement of confinement capability for long-lived nuclides in a simplified Purex process is the primary subject of this R&D project. Technologies for long-lived nuclide recovery are under development, focused on ¹⁴C and ¹²⁹I in head end, ²³⁷Np and ⁹⁹Tc in extraction, and ²⁴¹Am the daughter of ²⁴¹Pu in effluents. Those nuclides focused here are mobile in the environment and highly concerned as potential hazardous among the long-lived nuclides in spent fuels. New functions in PARC process concept are designed to mitigate the environmental impacts of reprocessing wastes and also to improve economy of reprocessing in the future.     

On “lower bound crack resistance curves” as material law for the safety assessment of components

Experimental fracture-mechanics investigations were carried out on large scale specimens. The specimen geometries and the crack depth ratio, a / W, in a parameter field were varied. Three materials of different toughness were chosen for the specimens. Their load—deformation behaviour, crack resistance curves, stretched zones Δa; and crack inititiation values J_i were determined and compared with the results from CT25 specimens. Numerical finite-element calculations were made to determine the state of stress in the specimens and the size of the plastic zones.     

Structural shock tests of prototype FBR “MONJU” scale models

This paper presents some results of experiments which simulate the structural dynamic response of a LMFBR primary coolant boundary to a hypothetical core disruptive accident (HCDA) based on scale models and high explosives. It was noted that high explosives are no longer a good simulant of the HCDA. However, the main purpose of the program, which included this experiment, is not to experimentally predict the dynamic response of the reactor structure at the HCDA, but to validate computer codes, which describe the pressure wave propagation and damage process in the reactor structures, using data obtained from these model experiments. The experiments were undertaken using many 1/15 scale simple models of the reactor vessels and internal structures, as well as 1/15 and 1/7.5 scale complex models of the interim design of prototype LMFBR ‘MONJU’. Simple model experiments involved a series of shock tests using pentolite to investigate the configuration effects of the vessel restraining section, the dipped-plate effect and the core barrel effect, respectively.     

Conceptual design of multi-purpose heat reactor “Nuclear Heat Generator”

Small heat reactors can apply to on site demand such as district heat and air conditioning, industrial process heat, greenhouse, and seawater desalination in urban and rural areas. The purpose of this paper is to design conceptually a multi-purpose reactor named “Nuclear Heat Generator (NHG)” which could be installed in energy consuming area. The reactor of 1MWt output is designed without any needs for fuel exchange and decommissioning on site. This cassette typed reactor vessel with sealing is transported to specified fuel fabrication shop every 3 to 4 years in order to exchange used fuels. Steam generators are involved in the self-pressurized integrated reactor with natural circulation. Generated steam pressure from heating reactor is 0.88 MPa (saturated) which is so less than that of current water reactors. Under low steam pressure it is considerably easy to make design of containment vessel and safety device. For economic competition overcoming scale demerit it will be necessary for the cassette type reactor to optimize its system design for the multi-production effect as well as modular construction and recycling system.     

Attachment concept for the “multi module segments” in a pulsed DEMO reactor

The pulsed thermonuclear demonstration reactor (DEMO) features challenging operational conditions such as high neutron fluxes, high temperatures, and significant thermo-mechanical stresses. These conditions do not require only a selection of advanced structural materials, but also the development of reliable means to assemble the in-vessel components together; allowing thermal expansions, disassembly, and maintenance in attractive scenarios. Over the course of DEMO lifetime, the materials are subjected to embrittlement by neutron irradiation, swelling, considerable thermo-mechanical fatigue and creep. Traditional joining methods may be rarely used in the harsh fusion environment to assemble different components. In addition any proposed layout should cope with the limited space available inside the vacuum vessel (VV).The objective of this study is to review the proposed attachment systems (developed within the latest European DEMO Conceptual Study) for the vertical segmentation concept called “multi module segments” (MMS). In order to find some place to house the attachments the blanket is cut respecting the Tritium Breeding Ratio limit for tritium self sufficiency. The conditions, neutronic and thermal, in which the attachments are supposed to operate, are calculated. The effects of pulsed operations have also been taken into account. The design of the attachments with the available structural materials with and without an active cooling system is analysed and a new concept for plug/unplug attachments is also suggested.