Blanket Neutronics Research Articles

Gas composition change during operation of a compact discharge fusion neutron source with a closed deuterium supply system

A compact cylindrical fusion neutron source has been developed for various applications, such as a D–T fusion neutron source for blanket neutron transport experiments for blanket development. Currently, the operation of a discharged fusion device is mostly deuterium–deuterium (D–D) fusion, in which D2 fuel is continuously fed to the vacuum chamber and pumped out to keep its pressure constant. Deuterium–tritium (D–T) fusion operation requires a closed supply system for radioactive tritium. So, the fuel supply system has been developed using ZrCo. In this system, however, fuel D2 gas is diluted due to light hydrogen released from the inner wall of the vacuum chamber and the surface of the electrodes, which results in a decrease of neutron production rate (NPR). Therefore, the external gas supply mode was performed to remove light hydrogen from the electrode surface before the operation of the closing system. The results of the gas analysis show that H2 was not increased and the D2 ratio was maintained. As a result, the neutron production rate achieved (NPR) > 105 n/s, and the dependency of the NPR on the voltage was similar to that of the external supply mode. These results indicate that removing light hydrogen from the electrode surface is effective to suppress fuel gas dilution.

Neutronics optimization study on the first wall design of CFETR for TBR enhancement

Tritium breeding capability is one of the most important objectives for fusion reactor design, and the tritium self-sufficiency has been set as a crucial target for the China Fusion Engineering Test Reactor (CFETR). The first wall (FW) is the first component of the blanket facing plasma, which bears high heat flux and high neutron current. Due to the neutrons entering into the blanket pass through the FW and then absorbed by the breeder materials, the FW design directly influences the blanket neutronics and tritium breeding performance. In this study, the helium-cooled solid breeder (HCSB) blanket of CFETR is selected as the reference design, and the impacts of the FW design parameters on tritium breeding performance have been studied comprehensively. A new conceptual mixed armor design and FW neutronics optimization guidelines have been proposed to enhance the tritium breeding ratio (TBR). Based on the mixed armor design and the optimization guidelines, the preliminarily optimized helium-cooled FW design shows about 9% TBR enhancement.

SINBAD – Radiation shielding benchmark experiments

The creation of an international shielding benchmark database was presented in 1988 at the International Reactor Shielding Conference (ICRS7) in Bournemouth, UK.M. Salvatores was among the authors of the proposal and had promoted and contributed to the project since the first initiatives, showed continued interest and encouraged the development of the database. He was Chairman of the Committee of Reactor Physics (NEACRP) for 2 years (1984–1985) and Chair of the Shielding benchmark group (1982–1988). In particular, he chaired two annual meetings in 1984 and 1985, called to initiate the collaborative programme on the analysis of shielding benchmarks for the validation of the JEF data files where the need to organize shielding benchmark was recognized and the presentation at ICRS7 defined the overall project.SINBAD officially started in the early 1990′s as a collaboration between the OECD/NEA Data Bank and RSICC with the goal to preserve the information on the performed radiation shielding benchmark experiments and make these available in a standardised form to the international community. One key point concerned the sensitivity and uncertainty analyses required to define their quality and figures of merit. The database comprises now 102 shielding benchmarks, divided into three categories, covering both low and inter-mediate energy particles applications: fission reactor shielding (48 benchmarks), fusion blanket neutronics (31), and accelerator shielding (23) benchmarks. The database is intended for different users, including nuclear data evaluators, computer code developers, experiment designers and university students. SINBAD is available from RSICC and from the NEA Data Bank. The database was extensively used within the scope of numerous national and international projects, such as PWR Pressure vessel surveillance, fusion programme (ITER reactor studies), different OECD Working Parties on Evaluation Cooperation (WPEC) Subgroups, nuclear data validation, IAEA nuclear data projects, etc.The history of the database and few examples of its use are illustrated, for cross-sections, response functions and covariance matrix validation.

Evaluation of tritium production rate in a blanket mock-up using a compact fusion neutron source

We report a neutronics study of a blanket mock-up using a discharge-type compact fusion neutron source. Deuterium–deuterium fusion neutrons were irradiated to the mock-ups composed of tritium breeder and neutron reflector/moderator. The tritium production rate (TPR) per source neutron was measured by a single-crystal diamond detector with a 6Li-enriched lithium fluoride film convertor after the calibration process. Despite the low neutron yield, energetic alpha and triton particles via 6Li(n, t)α neutron capture as well as 12C via elastic scattering were successfully detected by the SDD with high signal to noise ratios. The TPRs were experimentally evaluated with errors of 8.4%–8.5% at the 1σ level at the positions with high thermal neutron fluxes where the errors were dominantly introduced by uncertainties in the monitoring of the neutron production rate. The calculated to experimental (C/E) values of TPR were evaluated to be 0.91–1.27 (FENDL-2.1) and 0.94–1.28 (FENDL-3.1). As the neutron source can generate 14 MeV neutrons using a mixed gas of deuterium and tritium, this approach provides more opportunities for blanket neutronics experiments.

Evaluation of lithium glass scintillation detector responses for tritium production rate measurement in blanket neutronics experiments using the PHITS code

Lithium (Li) glass scintillation detectors were utilized by the neutron response of a 7Li glass subtracted from the neutron response of a 6Li glass to evaluate the tritium production rate (TPR) in the fusion blanket neutronics experiments, which is a so called Li glass method. TPRs in the Li2O blanket and the Li/V-alloy blanket were measured by a 6Li2O and a 6Li2CO3 pellet method, and by the Li glass method in the benchmark neutronics experiment with D-T neutrons at the Fusion Neutronics facility (FNS) of JAERI. TPRs measured by the Li glass method are lower than TPRs measured by the pellet method. The neutron responses of Li glass detectors have been calculated to study the discrepancies of two different methods by using the PHITS code. In the blanket, the neutron spectra change to harder from the backside to the front side of blanket. The reaction rate of several reactions, such as 6Li(n, n’α)D, 7Li(n, n’α)T, 7Li(n, 2n)6Li, and the elastic scattering of 6Li and 7Li increase with the neutron energy. The low-energy neutron peak was considered only for the calculation of TPR from Li glass, which loses the information of the 6Li(n, α)T reaction from high-energy neutron range. In addition, the reaction rate of this peak evaluated by the Li glass method is lower than real TPR of this peak for 6Li glass caused by weakened effect from other reactions. Therefore, the ratios of this peak reaction rate from the Li glass method against the total TPR of 6Li glass were evaluated to be a correction factor to correct Li glass method. After the correction, TPRs measured by Li glass method agree well with TPRs measured by pellet method.

РАСЧЕТНО-ЭКСПЕРИМЕНТАЛЬНОЕ ОПРЕДЕЛЕНИЕ ПРОСТРАНСТВЕННО-ЭНЕРГЕТИЧЕСКОГО РАСПРЕДЕЛЕНИЯ НЕЙТРОНОВ В БОКОВОМ ЭКРАНЕ РЕАКТОРА БОР-60. ЧАСТЬ 2

The BOR-60 reactor is actively used for conducting irradiation experiments in support of nuclear reactors both under operation, construction and design. Most of these experiments are conducted in the reactor core using various types of irradiation rigs. However, in recent years there has been keen interest in conducting reactor tests in the BOR-60 blanket. This is associated with several factors: the availability of blanket cells, weak influence on the reactor neutronic characteristics and safety, wide range of irradiation conditions (temperature, neutron flux density, damage doses, neutron spectrum), and cost of irradiation. The experience in BOR-60 blanket computational and experimental studies showed an ill agree-ment between the calculated and experimental neutronic characteristics. In addition to no orientation of the samples under study relative to the core center, there is a significant radial non-uniformity in the neutronic characteristics distribution in the BOR-60 blanket that limits its application for experimental studies. Therefore, detailed computational and experimental studies of the spatial-energy distribution of neutrons in the BOR-60 steel blanket is an important and urgent task.

Parameter study on helium cooled ceramic breeder blanket neutronics with CFETR integration design platform

The development of CFETR integration design platform is in progress to accelerate the CFETR design in consideration of both core physics and engineering practicability. Within the engineering framework of CFETR integration design platform, the neutronics interface module has been implemented. This work presents the parameter study on helium cooled ceramic breeder blanket neutronics with the neutronics interface module of the CFETR integration design platform. The parameterized 3-D neutronic model of CFETR was set in the first place. Afterwards, the workflow of neutronic analyses was established. The concerned geometry and composition parameters of breeder blanket were selected as input and the tritium breeding ratio was selected as target. And, then, the workflow was abstracted as the mathematical model in the form of the so called response surface based on sufficient MCNP simulations. With the workflow or response surface model, the estimation and optimization of target parameter could be achieved. Finally, a set of local optimal parameters was proposed for CFETR helium cooled ceramic breeder blanket.

Lithium Conducting Solid Electrolytes for Electrolysis of Lithium Tritide and Enrichment of Lithium-6

Lithium stable, lithium conducting electrolytes may improve methods to attain high tritium breeding yields required by many current D-T fusion energy production designs. Two ways such an electrolyte could aid is by reducing one unit of operation from the current baseline method from the process of tritium extraction from the breeding blanket, and by enrichment of lithium-6 from natural lithium to allow for use of a lithium-6 enriched blanket in combination with blanket neutron multipliers to attain a higher breeding per fusion neutron. Lithium tritide (LiT) is generated as a by-product of a tritium reaction with lithium cooling blanket material in fusion reactors for tritium breeding and power production. Recovery of tritium from the blanket materials is important to prevent environmental release and to recover valuable material. The Maroni process is the leading method to recover LiT from the lithium blanket using a molten salt liquid-liquid extraction method. The Maroni process can introduce salt components into the fusion reactor if separation is not complete and that can cause corrosion and activation of the salt components. SRNL has developed a direct LiT electrolysis method in the lithium blanket material using solid lithium ion conductors and has shown the evolution of hydrogen from the Li/LiH material during electrolysis. This method has been demonstrated at SRNL and results will be discussed. The U.S. has historically stockpiled lithium-6 and lithium-7 for use in nuclear applications. These stockpiles of lithium isotopes are being depleted and the U.S. closed production facilities in the 1980's. The longest operating plants in the U.S. used the column exchange (COLEX) process, which utilized an Hg-Li amalgam that preferentially forms with lithium-6. The operation of these plants resulted in significant environmental contamination with mercury (Hg). SRNL is now conducting research to use the same type of electrolyte as used for LiT electrolysis to enrich lithium in lithium-6. This research is to develop a high temperature method for lithium-6 enrichment that includes conduction of lithium-6 through a solid state electrolyte and the formation of a molten Li-Bi alloy. Both the solid state conduction of Li and the formation of the Li-Bi alloy can potentially lead to enrichment of lithium-6 based on current research findings and thermodynamic modeling. Combining the conduction and alloying steps will make it possible to create a smaller and more efficient separation process. The use of bismuth (Bi) is much more environmentally friendly than Hg and has a similar energetic change, which assists the separation process. Results to date will be discussed.

Physical investigation for neutron consumption and multiplication in fusion–fission hybrid test blanket module

Inelastic scattering of high energy fusion neutrons does affect the performance of fusion blanket based on the choice of different materials. It will also affect the behavior of source neutrons in a subcritical fusion fission hybrid blanket and consequently the transmutation and tritium breeding performance. A fusion fission hybrid test blanket module (HTBM) is designed which is presumed to be tested in a large sized tokamak and plasma neutron source is similar to ITER. In this preliminary design of HTBM the neutron source and loss factors are computed for the detailed neutronic performance analysis. The neutronic analysis of hybrid blanket module is performed for five different TRU fuel types: TRU-Zr, TRU-Mo, TRU-Oxide, TRU-Carbide and TRU-Nitride. In this module design, it is aimed to burn and transmute the TRU nuclides from high-level radioactive waste of PWR spent fuel. The effect of TiC reflector on transmutation and tritium breeding performance of HTBM is also quantified. MCNPX is used for neutronic computations. Neutron spectrum, capture to fission ratio and waste transmutation ratio of each fuel type are compared to evaluate their waste transmutation performance. Tritium breeding ratio is also compared for two coolant options: Li and LiPb eutectic.

Investigation on the Possibility of Tritium Self-Sufficiency for CFETR Using a PWR Water-Cooled Blanket

The neutron wall load ( \(P_{\mathrm{n}})\) of Chinese fusion engineering testing reactor (CFETR) will be less than 1 MW/m2. To meet the net tritium breeding ratio (TBR) of the reactor, a new water-cooled blanket concept is considered. The blanket neutronics schemes are performed to explore the local TBR issues in the \(P_{\mathrm{n}}\) range of 1–5 MW/m2, which aims at the effective design of the blanket concept considering the tritium self-sufficiency. As a result, the calculation results are compared with the local TBR values and the material fraction changes. It is found that the local TBR has the high value at low \(P_{\mathrm{n}}\) while the blanket size in radial direction is determined. It is mainly because of the total breeding area increasing due to the pipe pitch increasing in the model. This leads to the possibility for CFETR using a simplified blanket interior. In addition, to match the pressurized water reactor (PWR) water-cooled condition, a reduced size of blanket module in toroidal direction is achievable. It can be concluded that a PWR water-cooled blanket has more benefits to CFETR engineering implementation in the future.

20 Years of SINBAD (Shielding Integral Benchmark Archive and Database)

The SINBAD project started in the early 1990’s as a collaboration between the Organization for Economic Cooperation and Development’s Nuclear Energy Agency Data Bank (OECD/NEADB) and the Radiation Safety Information Computational Center (RSICC) at the Oak Ridge National Laboratory (ORNL) with the goal to preserve the information on the performed radiation shielding benchmarks and make them available in a standardised form to the international community. The SINBAD database comprises now 100 shielding benchmarks, covering fission reactor shielding, fusion blanket neutronics, and accelerator shielding. A thorough revision of the SINBAD benchmark experiments was undertaken recently in order to verify the completeness and consistency of the benchmark information, in particular concerning the evaluation of the experimental sources of uncertainty. New improved computer code models were prepared. This review process is expected to provide users with an easier choice and help them make better use of the experimental information. The OECD NEA Working Party on Scientific Issues of Reactor Systems (WPRS) Expert Group on Radiation Transport and Shielding (EGRTS) was started in 2011 with the mandate among others, to monitor, steer and support the continued development of the SINBAD database. SINBAD is available from RSICC and from the NEA Data Bank.

Numerical simulation of a blanket cooling system for fusion reactor based on PWR conditions

The simulations of a blanket cooling system were presented to address the choice of cooling channel geometry and coolant input data which are related to blanket engineering implementation. This work was performed using computer aided design (CAD) and computational fluid dynamics (CFD) technology. Simulations were carried out for the blanket module with a size of 0.6m×0.45m in toroidal plane, and the nuclear heat was applied on the cooling system at Pn (neutron wall load) of 5MW/m2. The structure factors and input data of hydraulics were investigated to explore the optimal parameters to match the PWR condition. It was found that the inlet velocity of first wall (FW) channel should be within the range of 2.48–3.34m/s. As a result, the temperature rise (TR) of the coolant in the FW channel would be 24–25K. This leads to the remaining space for TR within the range of 15K in the piping circuits. It also indicated that the FW plays an important role in TR (reaches 60% of the whole cooling system) due to its high level of Pn and heat flux in the zones. It was predicted that the nuclear heat inside blanket module could be removed completely by the piping circuits with an acceptable pipe bore and the related input data. Finally, a possible design range of cooling parameters was proposed in view of engineering feasibility and blanket neutronics design.

Resourceability on nuclear fuel cycle by transmutation approach

A resourceability on nuclear fuel cycle by transmutation of fission products in the spent fuel of nuclear reactors is discussed in this paper to investigate the feasibility of “creation and utilization” of Apres ORIENT from Adv.-ORIENT cycle, in which chemical “separation and utilization” of nuclear rare metals (platinum group metals, Mo, Tc, rare earth, etc.) has been proposed since FY2006. Apres ORIENT research program was newly initiated in FY2011 for nuclear transmutation of fission products into stable or short-lived highly-valuable elements. In the resourceability of rare earth metals from fission products, non-radioactive Nd and Dy can be created from Pr and Tb, respectively, by transmutation. Especially, the Dy creation has a relatively high feasibility of about 10–20 %/y in creation rate. A proper moderation of neutrons in blanket of fast reactors may be required to provide a high creation rate of La from Ba. In light platinum group metals, non-radioactive Ru can be created from Tc by transmutation, of which creation rate is about 4–5 %/y in blanket of fast reactors. Pd created from Rh is almost non-radioactively depending on the isotope fraction of 107Pd. Rh creation from Ru is not feasible under the neutron irradiation of typical nuclear reactors.

Development of the Water Cooled Ceramic Breeder Test Blanket Module in Japan

The development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) is being performed as one of the most important steps toward DEMO blanket in Japan. For the TBM testing and evaluation toward DEMO blanket, the module fabrication technology development by a candidate structural material, reduced activation martensitic/ferritic steel, F82H, is one of the most critical items from the viewpoint of realization of TBM testing in ITER. In Japan, fabrication of a real scale first wall, side walls, a breeder pebble bed box and assembling of the first wall and side walls have succeeded. Recently, the real scale partial mockup of the back wall was fabricated. The fabrication procedure of the back wall, whose thickness is up to 90mm, was confirmed toward the fabrication of the real scale back wall by F82H. Important key technologies are almost clarified for the fabrication of the real scale TBM module mockup. From the view point of testing and evaluation, development of the technology of the blanket tritium recovery, development of advanced breeder and multiplier pebbles and the development of the blanket neutronics measurement technology are also performed. Also, tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been started as the verification test of tritium production performance. This paper overviews the recent achievements of the development of the WCCB TBM in Japan.

Re-Evaluation of the Neutronics Experiments for the ITER Shielding Blanket Using ENDF/B-VII, JEFF-3.1, and JENDL-3.3

MCNP calculations for a benchmark representative of the fusion blanket neutronics shielding in SINBAD (Shielding Integral Benchmark Archive Database) were performed by using the four nuclear data libraries, ENDF/B-VII, JEFF-3.1, JENDL-3.3, and FENDL-2.1. Neutron and gamma flux spectra were calculated at two positions in a mock-up of the ITER inboard shield system. The results were compared with each other and also compared with measured data from the neutronics experiments for the ITER shielding blanket. For neutron spectra calculations, it is noted that the MCNPX calculations using all libraries agree well with experiments at positions A and B. For gamma spectra calculations, an overall good agreement can be stated and the tendency of a slight underestimation with penetration depth is observed.

Investigation on the TPR prediction accuracy in blanket neutronics experiments with reflector at JAEA/FNS

In the neutronics experiment for the ITER test blanket module with a 6Li-enriched Li 2TiO 3 layer and a beryllium layer conducted at the FNS facility of Japan Atomic Energy Agency, the calculated tritium production rate (TPR) was approximately 10% larger than the measured value only when a neutron source reflector composed of SS316 was attached. On the other hand, the influence of the reflector on the TPR prediction accuracy was not seen in the recent blanket experiment with a natural Li 2TiO 3 layer, beryllium layers and the reflector. We investigated the former experiment in detail, and found an unphysical tendency in the measured TPR distribution. In order to clarify whether the deterioration of the TPR prediction accuracy originates from the reflector or not, we have conducted the same experiment as the previous experiment again. In the present experiment, the measured TPR distribution inside the 6Li-enriched Li 2TiO 3 layer agreed well with the calculated value within an estimated experimental error of 6%. We conclude that the overestimation of TPR observed in the previous experiment would be due to some experimental faults and that the TPR prediction accuracy is good even in the case with the reflector.

A development of user-friendly graphical interface for a blanket simulator

A web-based user-friendly graphical interface (GUI) system, named GUMBIS (Graphical User-friendly Monte-Carlo-Application Blanket-Design Interface System), was developed to cut down the efforts of the researchers and practitioners who study tokamak blanket designs with the Monte Carlo MCNP/MCNPX codes. GUMBIS was also aimed at supporting them to use the codes for their study without having through understanding on the complex menus and commands of the codes. Developed on the web-based environment, GUMBIS provides task sharing capability on a network. GUMBIS, applicable for both blanket design and neutronics analysis, could facilitate not only advanced blanket R&D but also the education and training of the researchers in the R&D.

Measurement of TPR distribution in natural Li 2TiO 3/Be assembly with DT neutrons

In the previous blanket neutronics experiments conducted at the FNS facility of Japan Atomic Energy Agency, the following disagreements between experiments and analyses have been pointed out: (1) in the experiment with a 6Li-enriched Li 2TiO 3 layer and a beryllium layer, the calculation was found to overestimate the tritium production rate (TPR) by approximately 10% when a neutron source reflector composed of SS316 was attached. (2) In the experiment with natural Li 2O pebbles sandwiched by beryllium layers, TPR near the rear beryllium layer was overestimated by up to 10%. In order to identify the cause of the above problems clearly, a new blanket neutronics experiment with DT neutrons using natural Li 2TiO 3 layer and beryllium layers was conducted at FNS. TPR distributions inside the natural Li 2TiO 3 layer in assemblies with and without the source reflector were measured with Li 2CO 3 pellets. The measured TPR agreed well with the calculations within 6% in both assemblies. The influence of the reflector and beryllium on TPR prediction accuracy was not seen in the present experiment.

The potential presence and minimisation of plutonium within the irradiated beryllium in fusion power plants

Beryllium is expected to be one of the materials used in DEMO-type fusion power plants as blanket neutron multiplier for solid breeders and also in ITER as protective layer of plasma facing components. Among the aspects to be considered when dealing with Be a relevant one is its impurities content. One of the impurity reported by manufacturers is uranium, and that could represent a main source of concern, because its irradiation under fusion neutrons will lead to the formation of actinides, such as plutonium. The paper presents a parametric assessment utilizing for beryllium material composition a range of possible U impurity content between 0 and 100 ppm wt. First Wall neutron wall load ranging from 0.5 to 2.2 MW/m 2 and neutron fluence from 0.5 to 10.5 MW year/m 2 are considered (for ITER and DEMO-type reactor, respectively) to estimate the Pu production. In all those analyses the irradiation scenarios considered are based on a deuterium–tritium plasma. The evaluation have been performed using the EASY-2007 activation code package, with the EAF2007 activation library. An estimate of the Pu production at the end of the ITER fusion machine preliminary operation with deuterium–deuterium plasma is also presented. A preliminary analysis of the regulatory constraints on the maximum Pu production and indications of possible way for its minimisation is presented and discussed in the paper.

Tokamak Reactor System Analysis Coupled with a Self-Consistent Shielding Calculation for the Development of a Fusion Reactor Concept

A system analysis has been performed to develop the concepts for a fusion reactor and to identify the design parameters by using the tokamak system analysis code at KAERI (Korea Atomic Energy Research Institute). The system code elucidates the device parameters which satisfy the plasma physics and engineering constraints by taking into account a wide range of plasma physics and technology effects, simultaneously. The calculation of 1-D neutronic system code was coupled with this tokamak system code to optimize the reactor parameters. The numerical simulation for blanket neutronics was performed with MCNP5 code to calculate the tritium breeding ratios and neutron multiplications, which were the input parameter of system code. With the coupled system analysis and one-dimensional neutronic calculation, we assessed various types of DEMO blanket concepts with the requirements for the DEMO selected as to demonstrate the tritium self-sufficiency, to generate a net electricity amount, and for a steady-state operation.