Mahmoud Z. Youssef

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.

Uncertainties in Prediction of Tritium Breeding in Candidate Blanket Designs Due to Present Uncertainties in Nuclear Data Base

Estimates of the uncertainty ..delta../sub D/ in predicting the achievable tritium breeding ration (TBR) due to the uncertainties in nuclear data base are presented for several fusion blanket concepts. Specifically, the impact of the current uncertainties in measuring basic nuclear data on the calculated TBR is analyzed and discussed for four leading blanket designs that utilize different breeding materials, namely, Li/sub 2/O, 17Li-83Pb, LiAlO/sub 2/, and flibe. The impact on the TBR values of various evaluations for beryllium, which is employed as a multiplier in the latter two blankets, has been studied. Estimates for ..delta../sub D/ in other blanket concepts have also been assessed. Moreover, estimates have been made, based on previous studies, for the contribution to ..delta../sub D/ introduced by using neutron cross-section libraries that have different group structure and weighting spectra. Based on statistically incorporating the present cross-section uncertainties and their correlation in the analysis, the range of the uncertainty in TBR was found to be between 2 and 6% in all the concepts considered. The nonstatistical treatment for cross-section errors tends to give larger values for ..delta../sub D/. The uncertainty in TBR introduced by misrepresenting the secondary energy-angle distribution of the /sup 9/Be(n,2n) cross-section ranges from approx.morexa0» =4% in the flibe to approx. =2% in the LiAlO/sub 2/ blanket. Uncertainty up to approx. =15% can be encountered in the TBR evaluation in thin blankets with natural /sup 6/Li enrichment if broad-group cross-section libraries are used. However, this uncertainty can be reduced upon using an appropriate weighting spetrum representative of the one found in these blankets type.«xa0less

Lattice Calculations and Three-Dimensional Effects in a Laser Fusion-Fission Reactor

Three-dimensional neutronics models of the SOLASE-H fusion-fission reactor have been analyzed by Monte Carlo. In this design, light water reactor (LWR) fertile ThO/sub 2/ fuel bundles are enriched in the fissile isotope /sup 233/U and then shipped for burning in the LWRs. A concept where the fertile fuel bundles constitute a lattice configuration with the moderator-multiplier material is investigated. Parametric lattice calculations as a function of the neutron moderator-multiplier to fuel volume ratio (v/sub m//v/sub f/) in the lattice show that it is possible in such a concept to enhance the fissile nuclei production density in the fertile fuel, compared to cases where a lattice configuration is not used.

Post-analysis for the line source phase IIIA experiments of the USDOE/JAERI collaborative program on fusion neutronics

Abstract Experimental simulation to a line source has been realized at FNS, JAERI, within the USDOE/JAERI collaborative Program on Fusion Neutronics. This simulation achieved by cyclic movement of an annular Li2O test assembly relative to a stationary point source was a step forward in better simulation of the energy and angular distributions of the incident neutron source found in Tokamak plasmas. Thus, in comparison to other experiments previously performed with a stationary point source in the program, the uncertainties (that are system-dependent) in calculating important neutronics parameters, such as tritium production rate, will be more representative of those anticipated in a fusion reactor. The rectangular annular assembly used is 1.3 m × 1.3 m and 2 m long with a square cavity of 0.42 m × 0.42 m cross-section where the stimulated line source is located axially at the center. There is a 1.5 cm - thick S.S. first wall followed by a 20 cm-thick Li2O zone and a 20 cm - thick Li2CO3 zone. The ends of the rectangular assembly were left open. Calculations were performed for many measured items that include tritium production rate from 6Li(T6), 7Li(T7), in-system spectrum measurements, and various activation measurements [e.g. 58Ni(n, 2n), 58Ni(n, p), 93Nb(n, 2n), 90Zr(n, 2n), 27Al(n, α), 115In(n, n′), and 197Au(n, γ)]. These measurements were performed in three radial drawers inside the Li2O and Li2CO3 zones. Flux mapping with foil activation measurements were also performed in the axial direction (z = − 100 cm to z = 100 cm) at the front surface of the assembly in the cavity with the annular blanket in place and comparison was made to the bare line-source case (without annular blanket). The US has used the DOT5.1 code along with RUFF in the determistic calculations while MCNP was used in the Monte Carlo analysis. END F/B-V data was applied in this case. The corresponding codes/data used by JAERI are DOT3.5 along with FNSUNCL, MORSE-DD, and JENDL-3 cross-section data file. In this paper, the calculated-to-measured values, C/E, for the above-mentioned measured items will be given, as obtained individually by the US and JAERI. It will be shown that even with mechanically simulating a line-source, the present methods and codes can well predict the measured items under consideration without a particular difficulty in modeling. The C/E values for T6 and T7 are closer to unity (10%) than those obtained in the previous point source experiments, while the reaction rates are within 10–15% of the measured values.

The Nuclear Analysis of an Annular Li2O Blanket System Surrounding an Artificially Simulated 14-MeV Line Source and Comparison of Calculations to Measurements

Experimental simulation to a line source has been realized at the Japan Atomic Energy Research Institute (JAERI) Fusion Neutronics Source within the U.S. Department of Energy/JAERI collaborative program on fusion neutronics. This simulation, achieved by cyclic movement of an annular Li 2 O test assembly relative to a stationary point source, was a step forward in better simulation of the energy and angular distributions of the incident neutron source found in tokamak plasmas. Thus, compared with other experiments previously performed with a stationary point source, the uncertainties (that are system dependent) in calculating important neutronics parameters, such as tritium production rate (TPR), will be more representative of those anticipated in a fusion reactor. The rectangular annular assembly used is 1.3 x 1.3 m and 2. 04 m long with a square cavity of 0.42 x 0.42 m cross section where the simulated line source (2 m long) is located axially at the center. To characterize the incident neutron source, flux mapping with foil activation measurements was performed in the axial direction (Z=-100 cm to Z = 100 cm) at the front surface of the assembly in the cavity with the annular blanket in place, and comparison was made to the bare line-source case (without annular blanket). Three phases of experiments were performed. In Phase-IIIA, a 1.5-cm-thick stainless steel first wall was used. An additional 2.45-cm-thick carbon layer was added in Phase-IIIB, and a large opening (42.55 x 37.6 cm) was made at one side at the center of the annular assembly in Phase-IIIC. Calculations were performed independently by the United States and JAERI for many measured items that included TPR from 6 Li(T 6 ), 7 Li(T 7 ), in-system spectrum measurements, and various activation measurements. In this paper, the calculated-to-measured values for the aforementioned measured items are given, as obtained separately by the United States and JAERI. In addition, the mean value of the prediction uncertainties of the local and line-integrated TPR and the associated standard deviations are given based on the calculational and experimental results obtained in all the experiments.

The prediction capability for tritium production and other reaction rates in various systems configurations for a series of the USDOE/JAERI collaborative fusion blanket experiments

Abstract Seventeen integral fusion experiments have been performed so far within the USDOE/JAERI Collaborative Program on Fusion Neutronics. The main objective of these experiments is to verify the state-of-the-art neutron transport codes and nuclear data in predicting tritium production rates, in system neutron spectra, activation reaction rates, nuclear heating, and γ decay heating in a Li 2 O test assembly. In performing these experiments, the incident neutron source condition and the experimental geometrical arrangements for the test assembly were altered to study the impact of system changes on the prediction capability for the key neutronics parameters, particularly tritium production rate both locally and globally within predesignated zones in the breeding material. The test assembly itself was ehanged from a simple, one-material zone to a more prototypical blanket that included a stainless-steel first wall, neutron multiplier (beryllium) and coolant channels. The experiments proceeded through phase I and IIIA. In the latter phase, a line source was simulated by cyclic movement of the annular test assembly relative to the stationary point source that is located axially at the center of the inner cavity. In the latter phase, a better simulation has been achieved to the secondary energy, and angular distributions of the incident neutron source found in Tokamak plasmas. In this paper, the results obtained by the USA, quantified in terms of the calculated-to-experimental values (C/Es) for the key neutronics parameters, are discussed for all the experiments performed so far. The change in the trends of these C/E values as one moves from one phase to another is considered by statistically treating these C/E values to arrive at a mean value for the prediction uncertainty in each experiment and an average mean value to all the experiments. This was carried out for tritium production rate, in-system spectra, and other reaction rates.

Tritium and fissile fuel exchange between hybrids, fission power reactors, and tritium production reactors

A mathematical model extending work by Gordon and Harms is developed to describe the fissile fuel and tritium flows in a fusion-fission system consisting of a fusion hybrid reactor, a tritium production reactor, and several fission power reactors. The hybrid reactor plays the role of a fuel factory, providing the fission reactors and the tritium production reactor with their fissile fuel needs. The tritium production reactor ( a fission reactor) is devoted primarily to producing tritium for subsequent use in the hybrid. Different combinations of these systems are found by shifting the tritium breeding function among the various parts. At steady state, the total thermal power in fission reactors per unit of fusion power depends only on the total conversion ratio of the fission reactors and the hybrid. An economic analysis is required to determine which combination of systems will produce electricity at the lowest costs.

Neutronics performance of high-temperature refractory alloy helium-cooled blankets for fusion application

Abstract Among the concepts considered in the advanced power extraction (APEX) study is the helium-cooled refractory metal FW and blanket concept. Refractory metals exhibit high operating temperature and can offer good capability for withstanding high power density operation that is the focus of the APEX study. In this paper, we assess the impact of using various refractory metals on the nuclear heating profiles across the blanket and power multiplication (PM) and on the tritium breeding profiles and tritium breeding ratio (TBR). The refractory metals considered with liquid lithium breeder are W, TZM, and Nb–1Zr. The impact of Li-6 enrichment on these profiles and on TBR and PM is also assessed. Comparison of these nuclear characteristics is also made to other liquid breeder (Flibe and Li–Sn). Because the moderation power of these breeders to neutron energy varies among them, the damage to the structure is different with various structure/breeder combinations. The damage parameters (DPA, helium and hydrogen production) at key locations are also compared with the corresponding values in the thick liquid FW/blanket concept; an innovative design concept under consideration within the APEX study.

COMPONENT LIFETIME COMPARISON AND WASTE VOLUME IN CLiFF Sn/Flibe and Sn/LiPb BLANKETS

Abstract The thin Convective Liquid Flow First Wall (CLiFF) concept is one of the liquid FW concepts investigated in the Advanced Power Extraction (APEX) study for high power density application. Liquid tin has been suggested as the 2 cm thick front flowing liquid layer because of its low vapor pressure. Two choices were selected for the conventional blanket that follows the thin liquid wall (LW), namely: (1) LiPb/SiC blanket, and (2) Flibe/SiC blanket. Lithium is enriched to 90% Li-6 in the first blanket option and to 25% Li-6 in the second option (with 10 cm-thick beryllium front zone). Due to the superior attenuation characteristics of Flibe over LiPb, this impacted the lifetime of the SiC structure used in both options. In this paper, we assessed the lifetime of the SiC structure in the FW/Blanket and the shield in both blanket options. The end-of-life limit of 200 dpa is assumed (corresponding to ∼3% burn-up). The frequency of replacement of each component is estimated based on 30-years plant lifetime. Comparison is made for the waste volume of replaced components in each option. It is shown that the shield can last the plant lifetime in the Flibe blanket while part of the shield in the LiPb blanket will require replacement. The frequency of replacing the FW/Blanket with the LiPb blanket option is twice as much as in the Flibe blanket option. This is translated to a total volume of disposed structure at plant end-of-life from the entire FW/B/shield system that is larger by ∼60%.

Verification of ITER shielding capability and FENDL data benchmarking through analysis of bulk shielding experiment on large SS316:water assembly bombarded with 14 MeV neutrons

Abstract The recently developed FENDL-1 database, both in multigroup form (FENDL/MG-1.0) and continuous energy form (FENDL/MC-1.0) has been tested through analyzing a fusion integral experiment performed at the FNS facility, Japan, on a large bulk shielding assembly made of multilayers of SS316 and water. The assembly is a replica that simulates ITER shielding blanket and is bombarded by a 14xa0MeV source placed at 30xa0cm from the cylindrical assembly and housed inside a SS316 cylindrical can. This activity is undertaken as part of co-operation with JAERI on executing the required neutronics R&D tasks for ITER shield design. The objectives are (a) benchmarking FENDL-1 data and identifying any flaws that may exist in this newly developed database, and (b) examining the range of discrepancy between the calculated nuclear parameters inside the assembly and the measured ones in terms of the ratio of the calculated-to-experimental (C/E) data. Both differential and integral experimental data were analyzed along the central axis of the ∼ 120xa0cm D × 140xa0cm L assembly. The analyses with the multigroup data, MG also included library derived from ENDF/B-VI data base for comparison purposes. The MCNP Monte Carlo (MC) code was used with the FENDL/MC-1 data. The largest range of discrepancy between calculated and measured responses (reaction rates, neutron spectra, gamma ray heating, etc.) was found to be ∼ 20–30% even though in most cases this discrepancy falls within the experimental errors.