Demonstration Reactor Research Articles

COMPARISON OF THE DECAY HEAT REMOVAL SYSTEMS IN THE KALIMER-600 AND DSFR

A sodium-cooled demonstration fast reactor with the KALIMER-600 as a reference plant is under design by KAERI. The safety grade decay heat removal system (DHRS), which is important to mitigate design basis accidents, was changed in the reactor design. A loss of heat sink and a vessel leak in design basis accidents were simulated using the MARS-LMR system transient analysis code on two plant systems. In the analyses, the DHRS of KALIMER-600 had a weakness due to elevation of the overflow path for the DHRS operation, while it was proved that the DHRS of the demonstration reactor had superior heat transfer characteristics due to the simplified heat transfer mechanism.

Evaluation methodology for advance heat exchanger concepts using analytical hierarchy process

Abstract This study describes how the major alternatives and criteria being developed for the heat exchangers for next generation nuclear reactors are evaluated using the analytical hierarchy process (AHP). This evaluation was conducted as an aid in developing and selecting heat exchangers for integrating power production and process heat applications with next generation nuclear reactors. The basic setup for selecting the most appropriate heat exchanger option was established with evaluation goals, alternatives, and criteria. The two potential candidates explored in this study were shell-and-tube (helical coiled) and printed circuit heat exchangers. Based on study results, the shell-and-tube (helical coiled) heat exchanger is recommended for a demonstration reactor in the near term, mainly because of its reliability.

Overview of R&D activities on tritium processing and handling technology in JAEA

In JAEA, the tritium processing and handling technologies have been studied at TPL (Tritium Process Laboratory). The main R&D activities are: the tritium processing technology for the blanket recovery systems; the basic tritium behavior in confinement materials; and detritiation and decontamination. The R&D activities on tritium processing and handling technologies for a demonstration reactor (DEMO) are also planned to be carried out in the broader approach (BA) program by JAEA with Japanese universities. The ceramic proton conductor has been studied as a possible tritium processing method for the blanket system. The BIXS method has also been studied as a monitoring of tritium in the blanket system. The hydrogen transfer behavior from water to metal has been studied as a function of temperature. As for the behavior of high concentration tritium water, it was observed that the formation of the oxidized layer was prevented by the presence of tritium in water (0.23GBq/cc). A new hydrophobic catalyst has been developed for the conversion of tritium to water. The catalyst could convert tritium to water at room temperature. A new Nafion membrane has also been developed by gamma ray irradiation to get the strong durability for tritium.

Comparative Study of Cost Models for Tokamak DEMO Fusion Reactors

Cost evaluation analysis of the tokamak-type demonstration reactor DEMO using the PEC (physics-engineering-cost) system code is underway to establish a cost evaluation model for the DEMO reactor design. As a reference case, a DEMO reactor with reference to the SSTR (steady state tokamak reactor) was designed using PEC code. The calculated total capital cost was in the same order of that proposed previously in cost evaluation studies for the SSTR. Design parameter scanning analysis and multi regression analysis illustrated the effect of parameters on the total capital cost. The capital cost was predicted to be inside the range of several thousands of M$s in this study.

Conceptual core design study for a high-flux LFR demonstrator

A preconceptual core design study for a pool-type Lead-cooled Fast Reactor (LFR) demonstrator (DEMO) has been developed in the frame of an I-NERI between the Italian National Agency for the New Technologies, Energy and Sustainable Economic Development (ENEA) and Argonne National Laboratory (ANL), based on the European Lead-cooled System (ELSY) reference concept. A demonstration reactor is expected to prove the viability of technology to be implemented in the first-of-a-kind industrial power plant. DEMO specifications as a nuclear power facility demonstrating ELSY main features and performance, besides validating design methodology and tools, have been defined. Suitable design parameters have been set to meet the foremost objective of reaching a high fast neutron flux while respecting all technological constraints. Preliminary thermal-hydraulic analyses have been carried out to verify safety limits were not exceeded.

Parametric analyses of partial sub-assembly blockages of the 50 MWth gas cooled experimental technology and demonstrator reactor (ETDR)

This paper presents a parametric study of thermal hydraulic and structural mechanic analyses of accidental blockages of the hottest sub-assembly of the 50 MWth gas-cooled fast reactor, ETDR. The blockage ratios were 60% and 100% of the sub-assembly flow cross-section located at the first row of grid spacers. Temperature profiles in the fuel and cladding were calculated as a function of time using computational fluid dynamics. The results were incorporated into finite elements analyses to evaluate thermal and mechanical stresses and strains in the cladding and fuel. 2D simulations with generalized plane strains were used in the structural analyses applying the maximum power density in the pellet as calculated by CFD. The thermal analyses showed that a 60% partial blockage increases the maximum cladding temperature by 130 °C within a time period of 50 s, whereas a full blockage will lead to clad melting (1320 °C) in about 8 s after accident initiation. In the finite element analyses several, mainly conservative, assumptions have been made to incorporate the phenomena occurring in the pellet–cladding interaction. The results of the finite element analyses represent a first study of the pellet–cladding mechanical interface under given transients, exploring the development of a methodology to be used for future analyses. The model has been verified to predict realistic contact stresses in line with analytical solutions during partial or full blockages. However, more elaborate 2D and 3D contact models, with creep and irradiation creep material models for both fuel and clad, together with parametric studies on friction coefficients and number of fuel fragments are foreseen for future work on failure criteria.

The impact of cobalt aluminate formation on the deactivation of cobalt-based Fischer–Tropsch synthesis catalysts

It has been reported that cobalt aluminate formation is a cause of deactivation during Fischer–Tropsch synthesis (FTS), as it forms at the expense of active cobalt and is irreducible during FTS. To study this quantitatively, wax-coated Co/Pt/Al 2O 3 catalyst samples were removed periodically from an extended demonstration reactor FTS run operated at commercially relevant conditions and analysed with X-ray Absorption Near Edge Spectroscopy (XANES). With XANES, wax protected spent samples could be analysed in a “pseudo in-situ” mode. Under commercially relevant FTS conditions the catalyst undergoes reduction and minimal amounts of cobalt aluminate were found. It is proposed that the cobalt aluminate is formed from the residual CoO present in the catalyst after reduction. Additionally, the formation of aluminate was investigated with XANES and X-ray photoelectron spectroscopy (XPS) and TPR-MS on catalysts taken from laboratory continuous stirred tank reactor (CSTR) runs with varying water partial pressure (1–10 bar). Even at high water partial pressures ( P H 2 O = 10 bar , P H 2 O / P H 2 = 2.2 ) only around 10% cobalt aluminate is formed while the metallic fraction of cobalt still increased compared to the fresh catalyst. The work shows that cobalt aluminate formation during FTS at realistic conditions is not a major deactivation mechanism.

Status and key issues of reduced activation ferritic/martensitic steels as the structural material for a DEMO blanket

The status and key issues of reduced activation ferritic/martensitic (RAFM) steels R&D are reviewed as the primary candidate structural material for fusion energy demonstration reactor blankets. This includes manufacturing technology, the as-fabricated and irradiates material database and joining technologies. The review indicated that the manufacturing technology, joining technology and database accumulation including irradiation data are ready for initial design activity, and also identifies various issues that remain to be solved for engineering design activity and qualification of the material for international fusion material irradiation facility (IFMIF) irradiation experiments that will validate the data base.

ELSY—European LFR Activities

The European Lead Fast Reactor has been developed in the frame of the European lead system (ELSY) project funded by the Sixth Framework Programme of EURATOM. The project, coordinated by Ansaldo Nucleare, involved a wide consortium of European organizations. The ELSY reference design is a 600MWe pool-type reactor cooled by pure lead. The project demonstrates the possibility of designing a competitive and safe fast critical reactor using simple engineered technical features, whilst fully complying with the Generation IV goals. The paper focuses on the main aspects of the proposed design for the European lead fast reactor highlighting the innovation of this reactor concept and overall objectives. Special attention has been dedicated to safety starting from the first step of the design development taking into account other important aspects, such as the investment protection, the compactness of the primary system as well as sustainability. The main safety features of the proposed innovative decay heat removal (DHR) systems are presented. From the beginning of 2010, and for a duration of three years, the European Commission (EC) is financing the new project Lead European Advanced Demonstration Reactor (LEADER) as part of the 7th Framework Program. This paper highlights the main objectives of the LEADER project.

ELSY—European LFR Activities

The European Lead Fast Reactor has been developed in the frame of the European lead system (ELSY) project funded by the Sixth Framework Programme of EURATOM. The project, coordinated by Ansaldo Nucleare, involved a wide consortium of European organizations. The ELSY reference design is a 600MWe pool-type reactor cooled by pure lead. The project demonstrates the possibility of designing a competitive and safe fast critical reactor using simple engineered technical features, whilst fully complying with the Generation IV goals. The paper focuses on the main aspects of the proposed design for the European lead fast reactor highlighting the innovation of this reactor concept and overall objectives. Special attention has been dedicated to safety starting from the first step of the design development taking into account other important aspects, such as the investment protection, the compactness of the primary system as well as sustainability. The main safety features of the proposed innovative decay heat removal (DHR) systems are presented. From the beginning of 2010, and for a duration of three years, the European Commission (EC) is financing the new project Lead European Advanced Demonstration Reactor (LEADER) as part of the 7th Framework Program. This paper highlights the main objectives of the LEADER project.

Evaluation of Extrapolation for Creep-Fatigue Life by Hysteresis Energy

The creep-fatigue life has been evaluated by the hysteresis energy in 316FR stainless steel with low carbon and medium nitrogen, which is a candidate for structural material in Fast Breeder Reactor (FBR) plant with the design life of 60 years. The creep-fatigue is a main damage mode to prevent. The hysteresis energy rate is considered as the parameter to predict the life time. It is clear that the relationship between this parameter and the time to failure can be approximately expressed by the power-law function. The function depends on the ratio of plastic strain to total strain. Total fracture energy for creep-fatigue loading intends to be independent of the ratio of plastic strain to total strain in long-term test condition. The value is related to grain boundary strength for creep-fatigue loading because fracture mode in long-term test condition is intergranular fracture. The life could be predicted by the function in the case of no significant change of fracture energy. Coarse precipitation, for example sigma phase, might be considered as a factor to change the fracture energy. It is important to predict the precipitation formation. The result of life prediction by the hysteresis energy rate is compared with that of the time fraction rule based on “Demonstration Reactor Design Standard (Draft)”. The predicted lives by both methods for long-term region are comparable and independent of the ratio of plastic strain to total strain.

Deuterium ion irradiation for tritium breeding material and evaluation for tritium inventories in test blanket module and blanket of a demonstration reactor

The tritium produced in a tritium breeder of blanket has to be recovered under the temperature distribution to reduce the tritium inventory. Lithium titanate pebbles were irradiated by deuterium ions with different temperatures and ion fluences. The deuterium retained in the pebbles desorbed in forms of HD, D2, HDO and D2O. The amount of retained deuterium decreased for the temperature higher than 473K, and became to zero for the temperature higher than 773K. If the temperature range is taken from 573 to 1173K in the solid breeder TBM, the tritium inventory is lower than 1 gram. However, in the demonstration reactor, the total tritium inventory of blankets becomes a few kilograms. For the reduction of the tritium inventory, the region with the low temperature region has to be significantly reduced.

Cover Chem. Eng. Technol. 12/2010

AbstractHeterogenous Flow in a Bubble Column Demonstration Reactor (2010 © Evonik Oxeno GmbH)

Towards Diagnostics for a Fusion Reactor

The requirements for measurements on modern tokamak fusion plasmas are outlined, and the techniques and systems used to make the measurements, usually referred to as “diagnostics”, are introduced. The basics of three particular diagnostics - magnetics, neutron systems, and a laser based optical system - are outlined as examples of modern diagnostic systems, and the implementation of these diagnostics on a current tokamak (JET) are described. The next major step in magnetic confinement fusion (MCF) is the construction and operation of the International Thermonuclear Experimental Reactor (ITER), which is a joint project of China, Europe, Japan, India, Korea, the Russian Federation, and the U.S. Construction has begun in Cadarache, France. It is expected that ITER will operate at the 500 MW level. Because of the harsh environment in the vacuum vessel where many diagnostic components are located, the development of diagnostics for ITER is a major challenge - arguably the most difficult challenge ever undertaken in the field of MCF diagnostics. The main elements in the diagnostic step are outlined using the three chosen techniques as examples. Finally, the step beyond ITER to a demonstration reactor, DEMO, that is expected to produce several GWs of fusion power is considered and the impact on diagnostics outlined. It is shown that the applicability and development steps needed for the individual diagnostic techniques will differ. The challenges for DEMO diagnostics are substantial and a dedicated effort is required to find and develop new techniques, and special techniques appropriate to the DEMO environment. It is argued that the limitations and difficulties in diagnostics should be a consideration in the optimization and designs of candidate DEMOs.

Interfacial properties of HIP joints between beryllium and reduced activation ferritic/martensitic steel

ITER test blanket modules are the most important components to validate energy production and fuel breeding for future fusion demonstration reactors. Reduced activation ferritic/martensitic steel is recognized as one of the promising structural materials for the breeding blanket systems. Beryllium is a primary candidate plasma facing materials for ITER blanket. In this work, the interfacial properties of Be/reduced activation ferritic/martensitic steel (RAF/Ms) joints were investigated for the first wall of an ITER test blanket module (TBM). The joints were produced by the solid-state hot isostatic pressing (HIP) method. Chromium (Cr) was used as a diffusion barrier with a thickness of 1 μm or 10 μm, formed by plasma vapor deposition on the Be surface. The HIPping was conducted at 1023 K and 1233 K with 160 MPa of static pressure. The temperatures are standard normalizing and tempering temperatures of F82H. EPMA showed the Cr layer effectively worked as a diffusion barrier at 1023 K. However, for the F82H/Be interface which underwent HIP at 1233 K followed by tempering a Be rich layer was formed. Bend tests revealed that a thin Cr layer and low temperature HIP is preferable. The joint with a thick Cr layer suffer from brittleness of Cr itself.

Nuclear design analyses of the helium cooled lithium lead blanket for a fusion power demonstration reactor

This work presents nuclear design analyses for a demonstration power reactor (DEMO) employing the HCLL (helium cooled lithium lead) blanket for the tritium breeding and the power production. The related neutronic calculations were performed with the Monte Carlo code MCNP using a dedicated 3D torus sector model of DEMO. The model has been generated by the geometry conversion tool McCad from available CAD models developed in the frame of the European DEMO study for a device with integrated HCLL blanket modules. The nuclear design analyses address the tritium breeding capability, the shielding performance and the nuclear power production taking into account various engineering design parameters. The HCLL type DEMO was shown to provide tritium self-sufficiency with a sufficient high safety margin. The shielding performance was proven to be sufficient to allow the operation of the DEMO power plant over the full-anticipated lifetime.

Recent activities on tritium technologies of BA DEMO-R&D program in JAEA

The R&D for tritium technologies to a demonstration reactor (DEMO) are planned to be carried out in the Broader Approach (BA) program in Japan by JAEA with Japanese universities: (1) tritium analysis technology; (2) basic tritium safety research; and (3) tritium durability test. The EU joins the discussions and assessment of the R&D results. As the recent activities on the tritium technologies of the BA program, a multi-purpose RI facility, where the above R&D subjects will be carried out, has been designed in detail. A preliminary safety study has also been carried out for the amount of tritium released to the environment and for the radiation dose of workers. The facility is now under construction at Rokkasho in Aomori. The main subjects of the R&D of tritium analysis are the technologies for real-time analysis for hydrogen isotopes, gas, liquid and solid (such as microGC and calorimeter). The materials of interest include F82H, SiC, ZrCo, solid and liquid advanced breeder and multipliers. In the tritium durability tests, organic materials and metals planned to be used in a DEMO plant are studied for the radiation and the corrosion damage. A series of preliminary studies for the above subjects has been started.

Preliminary core characterization of a Generation IV lead fast reactor DEMO: Goals, design rationales and options

An Italian effort has been initiated under a cooperation between ENEA and CIRTEN for the investigation of a concept for a small size Generation IV Lead-cooled Fast Reactor (LFR) demonstration project (DEMO) which could provide significant support to the European Lead-cooled System (ELSY), by validating lead technology and the overall system behaviour. A demonstration reactor is expected to prove the viability of technology for use in a future commercial power plant, construction and operation, with the purpose of assessing the general strategy to use, to the largest extent. Thus, the first step towards the DEMO conceptual design has been the specification of the issues of ELSY – assumed as reference LFR – to investigate/validate, the objectives to reach, tools and methodology. As a result, a set of general requirements that must be met have been first identified and further scope analyses have been carried out in order to determine suitable core configurations. Finally, static neutronics analyses have been performed using the ERANOS2.1 deterministic code in conjunction with the JEFF3.1 nuclear data library to accomplish a preliminary core characterization; the mutual compatibility of both objectives and technological constraints has been confirmed a posteriori by thermal–hydraulic estimations. Results have shown that a 250 MWth lead-cooled MOX-fueled DEMO may be a viable concept: in the case of 480 °C reference coolant outlet temperature, 2 m s −1 average coolant velocity and 42 cm active core extension, high neutron flux values (around 2.5 times the ELSY ones) have been obtained (peak flux: 6.3 × 10 15 cm −2 s −1 with corresponding fast fraction of approximately 15.8%, i.e., some 1 × 10 15 cm −2 s −1) thanks to the significant power densities reached.

A Prospect on the Demonstration of Electric Power Production from Fusion Energy

This paper describes a prospect toward electric power production by the Fusion energy. In the first part of the paper, a principle of TOKAMAK system which is the successful magnetic-confinement-systems for fusion reactors are shown, and then the ITER project based on TOKAMAK and the present status of ITER is reviewed. In the remainder of the paper, a roadmap for fusion energy and conceptual designs of Demonstration reactors are briefly described.

Development of Safety Assessment Code for Decommissioning of Nuclear Facilities (DecDose)

A safety assessment code, DecDose, for decommissioning of nuclear facilities has been developed, based on the experiences of the decommissioning project of Japan Power Demonstration Reactor (JPDR) at Japan Atomic Energy Research Institute (currently JAEA). DecDose evaluates the annual exposure dose of the public and workers according to the progress of decommissioning, and also evaluates the public dose at accidental situations including fire and explosion. As for the public, both the internal and the external doses are calculated by considering inhalation, ingestion, direct radiation from radioactive aerosols and radioactive depositions, and skyshine radiation from waste containers. For external dose for workers, the dose rate from contaminated components and structures to be dismantled is calculated. Internal dose for workers is calculated by considering dismantling conditions, e.g. cutting speed, cutting length of the components and exhaust velocity. Estimation models for dose rate and staying time were verified by comparison with the actual external dose of workers which were acquired during JPDR decommissioning project. DecDose code is expected to contribute the safety assessment for decommissioning of nuclear facilities.