Tritium Production Rate Research Articles

The integral experiment on beryllium with D-T neutrons for verification of tritium breeding

A clean benchmark experiment on beryllium was performed with D-T neutrons at the FNS facility of the Japan Atomic Energy Agency. The main objective was to verify the integral data related to the tritium production on lithium isotopes. Tritium production rates, as well as activation reaction rates were measured inside the beryllium assembly that was shaped as a pseudo-cylindrical slab with an area-equivalent diameter of 628 mm and a thickness of 355 mm. Experimental results were analyzed with a three-dimensional Monte Carlo transport code MCNP-4C and FENDL/MC-2.0, JENDL-3.2/3.3 neutron transport libraries. Evaluation of reaction rates was based on the cross section data taken from the JENDL Dosimetry File and ENDF B-VI data libraries. Analysis shows that all calculation combinations (transport and activation cross section libraries) used for evaluation of reaction rates give data that is agreeable with measured values within 10%.

Nuclear analyses for two “look-alike” helium-cooled pebble bed test blanket sub-modules proposed by the US for testing in ITER

The US is proposing two “look-alike” sub-modules, based on helium-cooled pebble bed (HCPB) ceramic breeder, to be tested in the same test blanket module (TBM) that will occupy a quarter of a port in ITER and placed next to the Japanese TBM. The TBM has a toroidal width of 73 cm, a radial depth of 60 cm and a poloidal height of 91 cm. The ceramic breeder is made of Li 4SiO 4 with 75% Li-6 enrichment (60% packing factor) and beryllium is used as the multiplier. The two sub-modules are arranged in two configurations, namely a layered configuration and an edge-on configuration. In the present work, we analyze these two sub-modules using two-dimensional discrete ordinates transport codes in R-θ model that accounts for the presence of the ITER shielding blanket and the surrounding frame of the port. The objectives are: (1) to examine the profiles of heating and tritium production rates in the two sub-modules, both in the radial and toroidal direction, in order to identify locations where neutronics measurements can be best performed with least perturbation from the surroundings, (2) to provide both local and integrated values for nuclear heating rates required for subsequent thermo-mechanics analysis, and (3) to compare the tritium production capabilities of two variants for the HCPB blanket concept, mainly the parallel and the edge-on configurations. We present the main findings from this study in this paper.

Fusion power plant performance analysis using the HERCULES code

The HERCULES code was developed to assist in the safety and environmental analysis tasks of the European Power Plant Conceptual Study (PPCS). Its capabilities, however, extend beyond this kind of assessments. Coupling of the different computational tools and nuclear databases to the same geometry and material compositions, automation of input file generation and post-processing subroutines enable self-consistent calculations and radial–poloidal analyses of several parameters of interest in fusion technology design, optimising time and computer resources. Results include 2D variations of the neutron flux distribution and energy spectra, material activation and related quantities, wall loading, nuclear primary heating, tritium and helium production and radiation damage rates to materials. The capabilities of the neutron transport module of HERCULES are presented here and benchmarked on the PPCS models. Results are consistent with values assumed in the design process or obtained through the more specific PPCS analyses: tritium generation ratios are in the range between 1.12 and 1.36, energy multiplication between 1.16 and 1.39, and average wall loads between 1.90 and 2.54 MWm −2. Issues related to tritium breeding and possible tungsten first wall coating of the near-term models are discussed.

Tritium breeding experiments with blanket mock-ups containing [formula omitted]-enriched lithium titanate and beryllium irradiated with DT neutrons

Experiments have been conducted with breeding blanket mock-ups consisting of layers of candidate materials for JAERI’s pebble bed blanket design: lithium titanate, beryllium and reduced activation ferritic steel F82H. The mock-ups were irradiated with 14 MeV neutrons from the DT neutron source at FNS. The lithium titanate was enriched to 40 and 95% in Li 6 . The local tritium production rate in the 12 mm thick breeding layers was measured by means of pellet detectors of lithium titanate, Li 2CO 3 and liquid scintillation counting techniques. Additional verification of the neutron spectrum was obtained from material activation such as activation foils. The results were analyzed with the Monte Carlo neutron transport code MCNP-4C and several transport cross section libraries such as FENDL/MC-2.0. The calculation/experiment ratio was found to be within the combined error estimate of calculation and experiment but in some cases, the calculations showed an overestimation tendency.

On the nuclear response of the helium-cooled lithium lead test blanket module in ITER

The helium-cooled lithium lead (HCLL) concept has been recently selected as one of the two European reference designs foreseen for the breeding blanket of a demonstration fusion reactor. In particular, within the framework of the research and development activities on this blanket line, an HCLL test blanket module (TBM) has to be designed and manufactured to be implemented in ITER. At the Department of Nuclear Engineering (DIN) of the University of Palermo, a research campaign has been carried out to investigate the nuclear response of HCLL-TBM inside ITER by a numerical approach based on the Monte Carlo method. A realistic 3D heterogeneous model of HCLL-TBM has been set-up and inserted into an ITER 3D semi-heterogeneous one that simulates realistically the reactor lay-out up to the cryostat. A Gaussian-shaped neutron source has been adopted for the calculations. The main features of the HCLL-TBM nuclear response have been determined, paying particular attention to the deposited power and the tritium production rate together with the spatial distribution of their volumetric densities. The radiation damage of the structural material has also been investigated through the evaluation of displacement per atom and helium and hydrogen production rate.

International Benchmark Activity of Tritium Measurement of Blanket Neutronics

Tritium benchmark experiments with D-T neutron are a key issue to verify the tritium production rate (TPR) of the fusion blanket. The most useful method to measure the TPR in the neutron benchmark experiments is the liquid scintillation counting with Li2CO3 pellet. Ten years ago, the method of Li2CO3 pellet has been sufficiently verified the accuracy by means of D-T fast neutron irradiation and it was concluded within 10%. However, on the recent breeding blanket design, tritium is dominantly produced with the thermal neutron made with the scattering of D-T neutron and also the accuracy of the tritium production rate is requested below 10%. Therefore, previous verification is not sufficient for the recent blanket design and it is necessary to carry out the activity of the verification again. The JAERI, ENEA and TUD began to carry out the tritium benchmark experiment to verify the tritium production rate for the recent fusion blanket.

Methods for Tritium Production Rate Measurement in Design-Oriented Blanket Experiments

Direct and indirect methods have been developed for measurements of tritium production rates on different Li isotopes. The methods are mainly intended for design-oriented blanket experiments and are based on the activation of thin diagnostic pellets. In the direct method, the tritium activity bred from Li isotopes can be evaluated separately using two activated Li-containing pellets, measured by liquid scintillation counting. The method allows precise measurements of tritium activity in the pellet at a level of 2 Bq/g. The indirect method is based on the use of model activation reactions, 35Cl(n,α)32P and 31P(n,γ)32P, that possess a similar neutron spectrum sensitivity as direct tritium production reactions on 7Li and 6Li, respectively, and an effective measurement of the activation product of model reactions, 32P, by Cherenkov radiation counting. The reaction rate of model reactions can be calibrated in order to obtain the TPR.

International comparison of measuring techniques of tritium production for fusion neutronics experiments: Status and preliminary results

A neutronics experiment has been launched in the European fusion technology program on a mock-up of a test blanket module with the objective to validate the capability of nuclear data to predict nuclear responses with qualified uncertainties. In the experiment, the mock-up will be irradiated with a 14-MeV neutron source and the tritium production rate will be measured and compared with the calculated value by neutron cross sections from the European Fusion File. Given the difficulty of performing a reliable measurement of tritium production rate in such experiments, an international benchmark comparison on tritium measurement techniques has been carried out in collaboration between ENEA, TUD/VKTA and also JAERI that has already performed such experiments. The main purpose of the comparison is to establish the reference technique of TPR measurements to be used later in the TBM experiment. Three steps of the comparison procedure are presented in the paper as well as the preliminary results. The implications for the TBM experiment are also discussed.

Experimental studies on tungsten-armour impact on nuclear responses of solid breeding blanket

In order to experimentally evaluate the tungsten armour impact on tritium production of the solid breeding blanket being developed by JAERI for tokamak-type DEMO reactors, neutronics integral experiments have been performed using DT neutrons at the Fusion Neutron Source facility of JAERI. Solid breeding blanket mockups relevant to the DEMO blanket have been applied in this study. The mockups are made of a set of layers consisting of 0–25.2 mm thick tungsten, 16 mm thick F82H, 12 mm thick Li2TiO3 and 100–200 mm thick beryllium with a cross-section of 660 × 660 mm in maximum. Pellets of Li2CO3 are embedded in the Li2TiO3 layers to measure the tritium production rate. By installing the 5 mm, 12.6 mm and 25.2 mm thick tungsten armours, the sum of the integrated tritium productions at the pellets are reduced by about 2.1%, 2.5% and 6.1% relative to the case without the armour, respectively. Numerical calculations have been conducted using the Monte Carlo code. In the case of the mockups with the tungsten armour, calculation results for the sum of the integrated tritium productions agree well with the experimental data within 4% and 19% in the experiments without and with a neutron reflector, respectively.

On the Strategy and Requirements for Neutronics Testing in ITER

Neutronics testing is among the several types of fusion technology testing scheduled to be performed in ITER. The three ports assigned for testing will test several blanket concepts proposed by the various parties with test blanket modules (TBM) that utilize different breeders and coolants. Nevertheless, neutronics issues to be resolved in ITER-TBM are generic in nature and are important to each TBM type. Dedicated neutronics tests specifically address the accuracy involved in predicting key neutronics parameters such as tritium production rate, TPR, volumetric heating rate, induced activation and decay heat, and radiation damage to the reactor components. In this paper, we address some strategies for performing the neutronics tests. Tritium self-sufficiency cannot be confirmed by testing in ITER, however, the testing can provide valuable information regarding the main parameters needed to assess the feasibility of achieving tritium self-sufficiency. The paper also addresses the operational requirement (i.e. flux and fluence) as well as the geometrical requirement of the test module (i.e. minimum size) in order to have meaningful and useful tests. Measured neutronics data require high spatial resolution. This necessitates that the measured quantity be as flat as possible in the innermost locations inside the test module. This requirement has been confirmed in the present work based on results from two-dimensional calculations. The US and Japan solid breeder test blanket modules are placed inside half a port in ITER. The R-θ model used accounts for the presence of the ITER shielding blanket and the surrounding frame of the port.

Engineering Scaling Requirements for Solid Breeder Blanket Testing

An engineering scaling process is applied to the solid breeder ITER TBM designs in accordance with the testing objectives of validating the design tools and the database, and evaluating blanket performance under prototypical operating conditions. The goal of scaling is to ensure that changes in structural response and performance caused by changes in size and operating conditions do not reduce the usefulness of the tests. Initially, constitutive equations are applied to lay out the basic operating and design parameters that dominate blanket phenomena. The suitability of these similarity criteria for the TBM design is then confirmed by comparing finite element predictions of prototype and scale model responses. The TBM design also takes into account the need to check the codes and data for future design use. Specifically, predictability of tritium production and nuclear heating rates in a complex geometry, tritium release and permeation characteristics under fusion environments belong to this category. We conclude that this engineering scaling design process has maximized the value of ITER testing.

Heterogeneous breeding blanket experiment with lithium titanate and beryllium

An integral breeding blanket experiment has been done at Fusion Neutronics Source of JAERI 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 Li 2 CO 3 pellet detectors. The tritium production from 7 Li was estimated by means of the reactions 32 S(n, p) 32 P and 35 Cl(n, α ) 32 P which have effective cross-sections similar to 7 Li(n, n α )T but are more sensitive methods. The neutron flux in the assembly was checked with activation foils. The measurement results were analyzed 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.

On the nuclear response of the water-cooled Pb–17Li test blanket module for ITER-FEAT

Within the European Fusion Technology Programme, the Water-Cooled Lithium Lead (WCLL) DEMO breeding blanket line was selected in 1995 as one of the two EU lines to be developed in the next decades, in particular with the aim of manufacturing a Test Blanket Module (TBM) to be tested in ITER-FEAT. The present paper is focused on the study of the WCLL-TBM nuclear response in ITER-FEAT, being specifically oriented to the investigation of the local effects due to the typical C-shaped tubes of the breeder zone, since they could play a pivotal role in the module-relevant thermo–mechanical design. A 3D heterogeneous model of the WCLL-TBM, realistically simulating its new lay out and taking into account 9% Cr martensitic steel as reference structural material, has been set-up. A particular attention has been paid to the simulation of the characteristic ‘C’ shape of the breeder zone double walled tubes, which have been realistically reproduced. The WCLL-TBM model has been inserted into an existing ITER-FEAT 3D semi-heterogeneous model accounting for a proper D–T neutron source. Analyses have been performed by means of MCNP-4C code running on a cluster of four workstations through the implementation of a parallel virtual machine. For each analysis a large number of histories (>10.000.000) have been simulated, obtaining statistical uncertainties on the results lower than 3%. The main features of the WCLL-TBM nuclear response have been determined focusing the attention on power deposition density, material damage through DPA and He and H production rate, daily tritium production and tritium production rate radial distribution in the module. The obtained results are herewith presented and critically discussed.

Neutronics experiments for DEMO blanket at JAERI/FNS

In order to verify the accuracy of the tritium production rate (TPR), neutron irradiation experiments have been performed with a mockup relevant to the fusion DEMO blanket consisting of F82H blocks, Li2TiO3 blocks with a 6Li enrichment of 40% and 95%, and beryllium blocks. Sample pellets of Li2TiO3 were irradiated and the TPR was measured by a liquid scintillation counter. The TPR was also calculated using the Monte Carlo code MCNP-4B with the nuclear data library JENDL-3.2 and ENDF-B/VI. The results agreed with experimental values within the statistical error (10%) of the experiment. Accordingly, it was clarified that the TPR could be evaluated within 10% uncertainty by the calculation code and the nuclear data. In order to estimate the induced activity caused by sequential reactions in cooling water pipes in the DEMO blanket, neutron irradiation experiments have been performed using test specimens simulating the pipes. Sample metals of Fe, W, Ti, Pb, Cu, V and reduced activation ferritic steel F82H were irradiated as typical fusion materials. The effective cross-sections needed to calculate the formation of the radioactive nuclei (56Co, 184Re, 48V, 206Bi, 65Zn and 51Cr) due to sequential reactions were measured. From the experimental results, it was found that the effective cross-sections increased remarkably while coming closer to polyethylene board, which was a substitute for water. As a result of this present study, it has become clear that the sequential reaction rates are important factors in the accurate evaluation of induced activity in fusion reactor design.

Neutronic and photonic analysis of the water-cooled Pb17Li test blanket module for ITER-FEAT

Within the European Fusion Technology Program, the Water-Cooled Lithium Lead (WCLL) DEMO breeding blanket line was selected in 1995 as one of the two EU lines to be developed in the next decade, in particular with the aim of manufacturing a Test Blanket Module (TBM) to be implemented in ITER. This specific goal has been maintained also in ITER-FEAT program even if the general design parameters of the TBMs have reported some changes. This paper is focused on the investigation of the WCLL-TBM nuclear response in ITER-FEAT through detailed 3D-Monte Carlo neutronic and photonic analyses. A 3D heterogeneous model of the most recent design of the WCLL-TBM has been set-up simulating realistically its new lay out and taking into account 9% Cr martensitic steel as structural material. It has been inserted into an existing 3D semi-heterogeneous ITER-FEAT model accounting for a proper D–T neutron source. The analyses have been performed by means of MCNP-4C code running on a cluster of four workstations through the implementation of a parallel virtual machine. The main WCLL-TBM nuclear responses have been determined focusing the attention on power deposition density, material damage through displacement per atom (DPA) and He and H production rate, daily tritium production and tritium production rate radial distribution in the module. Moreover, the impact of using lithium at various Li 6 enrichment on the TBM nuclear response has been investigated. The results obtained are herewith presented and critically discussed.

Neutronics Experiment of 6Li-enriched Breeding Blanket with Li2TiO3/Be/F82H Assembly Using D-T Neutrons

To improve the nuclear performance of fusion breeding blankets, the application of thermal type blankets with 6Li-enriched ceramics and low-activation ferritic steel is suggested. However, appropriate neutronics experiments to investigate the nuclear performance of such blankets have never been carried out. Therefore, we have done blanket neutronics experiments with stratified 95-% enriched 6Li2TiO3, stainless steel F82H and beryllium blocks using JAERI’s Fusion Neutronics Source (FNS) and verified the prediction accuracy of the tritium production rate and the production rate of selected isotopes in F82H in with the Monte Carlo code MCNP-4B and JENDL Fusion File, JENDL-3.2 and ENDF/B-VI data. Calculated tritium and 51Cr production rates were in good agreement with the measured values within the experimental error. However, discrepancies were found for 187W and 56Mn production rates.

Heat deposition, damage, and tritium breeding characteristics in thick liquid wall blanket concepts

The advanced power extraction (APEX) study aims at exploring new and innovative blanket concepts that can efficiently extract power from fusion devices with high neutron wall load. Among the concepts under investigation is the free liquid FW/liquid blanket concept in which a fast flowing liquid FW (∼2–3 cm) is followed by thick flowing blanket (B) of ∼40–50 cm thickness with minimal amount of structure. The liquid FW/B are contained inside the vacuum vessel (VV) with a shielding zone (S) located either behind the VV and outside the vacuum boundary (case A) or placed after the FW/B and inside the VV (case B). In this paper we investigate the nuclear characteristics of this concept in terms of: (1) attenuation capability of the liquid FW/B/S and protection of the VV and magnet against radiation damage; (2) profiles of tritium production rate and tritium breeding ratio (TBR) for several liquid candidates; and (3) profiles of heat deposition rate and power multiplication. The candidate liquid breeders considered are Li, Flibe, Li–Sn, and Li–Pb. Parameters varied are (1) FW/B thickness, L, (2) Li-6 enrichment and (3) thickness of the shield.

Tritium Production Rate and Tritium Breeding Ratio: A Comparative Study of Li2O and Lithium Assemblies with Different Concentrations of 6Li

In view of the blanket design of a futuristic deuterium-tritium fusion reactor, a time-dependent study of 14-MeV neutrons has been carried out in bare lithium and Li2O blanket assemblies with different concentrations of 6Li nuclei. For assemblies of different sizes, time-dependent total neutron fluxes, a tritium production rate (TPR), and a tritium breeding ratio (TBR) up to 40% concentration of 6Li (natural concentration being 7.42 at.%) have been reported. A multigroup diffusion equation and eigenfunction expansion method has been used. This study shows that for any concentration of 6Li, the values of TPR as well as TBR are higher for a Li2O assembly than those obtained for all corresponding (of same size) assemblies of lithium. However, for a given assembly of lithium or Li2O, the TBR values do not show any observable change with 6Li concentration beyond ~40%. Further, for any concentration, the values of TPR and TBR decrease substantially in both types of systems as the side of the cubic assembly is reduced from 1 to 0.5 m.

Water-cooled Pb–17Li test blanket module for ITER: Impact of the structural material grade on the neutronic responses

The Water-Cooled Lithium Lead (WCLL) DEMO blanket is one of the two EU lines to be further developed with the aim of manufacturing by 2010 a Test Blanket Module for ITER (TBM). In this paper results of a 3D-Monte Carlo neutronic analysis of the TBM design are reported. A fully 3D heterogeneous model of the WCLL–TBM has been inserted into an existing ITER model accounting for a proper D–T neutron source. The structural material assumed for the calculations was martensitic 9% Cr steel code named Z 10 CDV Nb 9-1. Results have been compared with those obtained using MANET. The main nuclear responses of the TBM have been determined, such as detailed power deposition density, material damage through DPA and He and H gas production rate, radial distribution of tritium production rate and total tritium production in the module. The impact of using natural Lithium on the TBM system operation has also been evaluated.

Enhancing tritium breeding characteristics of a fusion reactor blanket

In this paper we have reported our results of a detailed time dependent study of neutron spectra, tritium production rate, tritium breeding ratio, slowing down time and average energy following a pulse of 14 MeV neutrons in Li+C assemblies of different sizes. We have solved the time dependent diffusion equation with energy dependent buckling as an eigenvalue problem. Our results show that an initial 14 MeV pulse slows down to the keV region within about 200 ns after its injection. At later times, its energy decreases rather slowly because elastic scattering is the sole mechanism of energy loss. Since most of the tritium producing reactions take place with neutrons having energies in this region, the tritium production rate is higher in a Li+C system than a corresponding natural lithium assembly. Moreover, in Li+C the calculated value of TBR is almost twice the value obtained for a natural lithium assembly of the same size. Also, by varying the concentration of 6Li we find that the value of TBR tends to saturate beyond about 40%.