Neutronics and photonics of inertial confinement fusion pellets with internal breeding

The neutronics and photonics performance of a pellet with a small DT core spark trigger, surrounded by a large volume of D to enable tritium and He-3 breeding, is examined. The response to a 70% DD and 30% DT composite neutron spectrum is calculated using either W, Be, or Pb as structural materials at core density radius products ranging from 9.42 to 94.2 kg/m². At a core density-product of 94.2, the DT neutron source leads to an excess particle multiplication of 0.43 neutrons per source neutron. The percentage of energy leakage from the pellet in the form of escaped neutrons is 42.3% of the source energy for the DT source, and 28.8% for the DD source. The gamma-ray energy percentage deposited in the pellet is 26.7% for the DT source and 106.6% for the DD source. For the pellet with the composite source, the energy multiplication factor is 1.27. Thus the large DD contribution to the composite neutron source results in the pellet performing many of the functions normally reserved for the blanket such as spectral softening, breeding, and neutron and energy multiplication. The neutron energy leakage is 38.4% of the source energy for the composite source. It is estimated that the neutron energy leakage amounts to 10% of the fusion energy, compared with 70% as neutron energy in a DT pellet. These results are significantly different from those encountered in conventional DT inertial confinement designs, and thus lower tritium inventories, higher power densities, reduced radiation damage, and materials activation of the reactor coolant and structure may be achievable.     

Calculation of response matrix of CaSO4:Dy based neutron dosimeter using Monte Carlo code FLUKA and measurement of 241Am–Be spectra

Response matrix for CaSO₄:Dy based neutron dosimeter was generated using Monte Carlo code FLUKA in the energy range thermal to 20 MeV for a set of eight Bonner spheres of diameter 3–12″ including the bare one. Response of the neutron dosimeter was measured for the above set of spheres for ²⁴¹Am–Be neutron source covered with 2 mm lead. An analytical expression for the response function was devised as a function of sphere mass. Using Frascati Unfolding Iteration Tool (FRUIT) unfolding code, the neutron spectrum of ²⁴¹Am–Be was unfolded and compared with standard IAEA spectrum for the same.     

Measurements of neutron induced activation of concrete at 64.5 MeV

The present paper presents the measurement of neutron induced activations on concrete using the 64.5 MeV quasimonoenergetic neutrons produced at the intense ⁷Li(p, n) neutron source at Cyclotron and Radioisotope Center, Tohoku Univeristy (CYRIC). The data were corrected for the effect of continuous neutrons in the source. The neutron energy, neutron yields and the spectrum of continuous neutrons were confirmed with the neutron time-of-flight method and the neutron activation measurement of the ²⁰⁹Bi(n, Xn) reactions having various threshold energy values. The nuclides produced by thermalized source neutrons are negligible. New data were obtained for concrete activation.     

Calculations of the transport of neutrons and secondary gamma rays through concrete for incident neutrons in the energy range 15 to 75 MeV

Earlier work by Alsmiller et al. considered coupled neutron and secondary-gamma-ray transport through a thick shield of silicon dioxide with 5% water by weight for neutron sources with energies of 50, 100, 200, 300 and 400 MeV. In that work, the approximation was made that gamma rays were produced only by neutron capture. In the present work, coupled neutron and secondary-gamma-ray transport through a thick shield of concrete for neutron sources with energies of 15, 25 and 75 MeV is considered. In this study, gamma-ray production for all interactions involving neutrons with energies up to 15 MeV was included; i.e., the approximation made here is that gamma-ray production can be neglected for interactions by neutrons with energies > 15 MeV.For incident neutron energies of 15, 25, 50, and 75 MeV, results of total and gamma-ray dose equivalents are given as a function of depth into the slab. For the 50- and 75-MeV incident neutron energies, the gamma-ray dose equivalent was found to be no more than 5% of the total dose equivalent at all depths considered ( 1500 g/cm²). For the 15- and 25-MeV incident neutron energies, however, the gamma-ray dose equivalent dominates at greater depths into the slab. A conservative estimate of the effect of including gamma rays produced in interactions with neutrons of energies > 15 MeV indicates that the calculated total dose equivalent would increase by no more than 5%.     

A function to provide neutron spectrum produced from the 9Be + p reaction with protons of energy below 12 MeV

A function to give the total neutron production cross section, angular distribution, and energy spectrum via the ⁹Be + p reaction has been created by fitting experimental data to characterize compact neutron sources with thick Be targets bombarded by protons with energy below 12 MeV. To examine the suitability of the function, calculations of the angle-dependent neutron energy spectra produced in thick Be targets with 4- and 12-MeV protons using the function were compared with corresponding experiments and calculations using the nuclear data libraries of ENDF/B-VII.0 and JENDL4.0/HE. The function was in better agreement with the experiments than the calculations using the libraries except for at backward angles. The ¹¹⁵In(n,n’)^115mIn reaction rates calculated using GEANT4 with source neutrons given by both the function and ENDF/B-VII.0 were compared with that measured at the RIKEN Accelerator-Driven Compact Neutron Source to evaluate the neutron spectrum above 1 MeV. The function slightly overestimated the measurement by 14% and the calculation with ENDF/B-VII.0 underestimated by 35%. It was concluded that the function can be applied in compact neutron source designs.     

Proton simulation of displacement effects induced in metals by 14 mev neutrons

In designing a D-T fusion reactor, one must know the effect of a high flux of 14 MeV neutrons on structural materials. Available laboratory sources of 14 MeV neutrons are not intense enough to expose samples to the expected flux. Bombardment with other particles is one way of simulating the anticipated neutron environment. The energy spectrum of atoms recoiling from collisions with bombarding particles can be calculated from elastic-scattering and nonelastic-reaction data for the incident species. This analysis shows that 16 MeV protons closely simulate the displacement effects caused by 14 MeV neutrons. In niobium the average atom recoiling from a 14 MeV neutron interaction has 65 keV of damage energy. The mean damage energy deposited per cm³ of niobium by a fluence of one 14 MeV neutron per cm² is 14 keV. The equivalent quantity for 16 MeV protons incident on niobium is 33 keV.     

Calculated neutron cross sections for Cu and Nb up to 32 MeV for neutron damage analysis

Cross sections for neutron interaction with Cu and Nb, with emphasis on spectra of light particles from binary reactions, are calculated for neutron energies from 4 to 32 MeV for estimating recoil probability densities for the analysis of damage experiments with a Be (d, n) neutron source. Nuclear model parameters were adjusted to reproduce the available cross-section data around 14 MeV. Helium production cross sections were also calculated for ⁶³Cu for neutrons below 20 MeV, as an illustration of the Hauser-Feshbach method for calculating tertiary reaction cross sections.     

Direct neutron spectrum measurement to validate Zr(n,2n) reaction cross-section at 14 MeV

Lithium zirconate, Li₂ZrO₃, is known as a candidate blanket material in a fusion reactor. Various neutronics benchmark experiments for zirconium have thus been carried out so far. According to the independent benchmark studies by two parties, the neutron spectrum calculations show fairly large overestimation for most evaluated nuclear data libraries. However, the reason has not yet been made clear up to now. The author's group expects it would be due to a problem of evaluation for the ^natZr(n,2n) reaction cross-section, because the cross-section measurement is basically not possible with the foil activation method for zirconium isotopes except for ⁹⁰Zr.In the present study, two neutrons emitted from ^natZr(n,2n) reaction have been measured directly to investigate the reason for the above overestimation. The measurement was done with our own special technique of detecting angle-correlated neutrons by the coincidence detection technique and the pencil-beam DT neutron source of FNS, JAEA. Angle-correlated energy differential cross-sections for ^natZr(n,2n) reaction were successfully measured. The obtained total cross-section above the emitted neutron energy of 800 keV was fairly larger than the one evaluated in JENDL-3.3. The total cross-section of ^natZr(n,2n) reaction was estimated by extrapolating the spectrum down to zero energy taking into account the nuclear temperature. The estimated cross-section value with the nuclear temperature of 1 MeV, which is larger than the one adopted in JENDL-3.3, was in acceptable agreement with JENDL-3.3. It is suggested from the result that the disagreement pointed out in the previous benchmark studies may be due to inappropriate nuclear temperature used in the evaluation. Further investigation of the nuclear temperature employed in the nuclear data evaluation should thus be carried out once again.     

Measurements of activation cross-sections for the F(n, α)N reaction for neutrons with energies between 13 and 15 MeV

In this study, activation cross-sections were measured for the ¹⁹F(n, α)¹⁶N reaction at six different neutron energies from 13.5 and 14.9 MeV. The fast neutrons were produced via the ³H(d,n)⁴He reaction on SAMES T-400 neutron generator. The cyclic activation technique was used. Induced gamma activities were measured by a high-resolution gamma-ray spectrometer with high-purity germanium (HpGe) detector. Measurements were corrected for gamma-ray attenuations, random coincidence (pile-up), dead time and fluctuation of neutron flux. Results were compared with the previous works.     

Cross Section Measurement for 199Hg(n,n′)199m Hg Reaction from 0.78 to 6.3 MeV

Cross sections for the ¹⁹⁹Hg(n, n′)^199m Hg reaction have been measured at 10 energy points from 0.78 to 6.3 MeV by the activation method. Monoenergetic neutrons below 2 MeV were produced by the ⁷Li (p, n)⁷Be reaction and those above 2 MeV were produced by the D(d, n)³He reaction using a 5.5 MV Van de Graaff accelerator. The neutron flux was determined with a proton recoil telescope counter and In-foils. The measured cross sections were expressed by an empirical formula. The fission spectrum averaged cross section calculated with this formula is 238.3 mb for the Watt-type fission spectrum, and is about 14% smaller than that recently measured by three of the authors.     

Feasibility of the application of statistical correlation techniques for the experimental study of the multiplication of 14 MeV neutrons in Be assemblies

The multiplication of 14 MeV neutrons through (n, 2n) reactions in Be and Pb, which are leading candidates for deployment as premultipliers in (D, T) fusion reactor blankets, is under active investigation in several laboratories. In the case of ⁹Be, since the secondary neutrons released in an (n, 2n) reaction could still carry enough energy to cause further generations of (n, 2n) reactions, a systematic study of the multiplicity distribution of the burst of neutrons leaking out of a finite-sized Be assembly, following deposition of a 14 MeV neutron in it, could yield very useful information on the physics of the (n, 2n) multiplicative process. Drawing from the wealth of experience available from previous studies in the field of reactor noise analysis and from the non-destructive assay of Pu content in sealed packages through measurement of the ²⁴⁰Pu spontaneous fission disintegration rate, as developed recently in the field of safeguards, a statistical correlation technique has been proposed for the study of (n, 2n) multiplication in Be (and Pb). The technique also appears to have potential applications for the study of the multiplicity distribution of neutrons produced in spallation targets.     

Development of the high-energy neutron fluence rate standard field in Japan with a peak energy of 45 MeV using the 7Li(p,n)7Be reaction at TIARA

In this study, we developed a 45 MeV neutron fluence rate standard of Japan. Quasi-monoenergetic neutrons with a peak energy of 45 MeV in the neutron standard field were produced by the ⁷Li(p,n)⁷Be reaction using a 50-MeV proton beam from an azimuthally varying field (AVF) cyclotron of the Takasaki Ion Accelerators for Advanced Radiation Application (TIARA). The neutron energy spectrum was measured using an organic liquid scintillation detector and a ⁶Li-glass scintillation detector by the time-of-flight method, and using a Bonner sphere spectrometer by the unfolding method. The absolute neutron fluence was determined using a proton recoil telescope (PRT) composed of the liquid scintillation detector and a Si(Li) detector that was newly developed in the present study. The detection efficiency of the PRT was obtained using the MCNPX code. The peak neutron production cross section for the ⁷Li(p,n)⁷Be reaction was also derived from the neutron fluence in order to confirm the neutron fluence of the TIARA high-energy neutron field. The peak neutron production cross section obtained in the present study was in good agreement with those of previous studies. The characteristics of the 45-MeV neutron field in TIARA were successfully evaluated in order to calibrate high-energy neutron detectors and high-energy neutron dosimeters.     

Fission neutron spectrum of Cf

Average cross sections for the unmoderated spontaneous fission neutron spectrum of ²⁵²Cf have been measured for 10 different threshold reactions by activation method. On the basis of the measured data and of the excitation functions the Maxwellian temperature for ²⁵²Cf neutron spectrum has been deduced and a value of T = 1.41 ± 0.02 MeV was found between 2.5 and 15 MeV. The temperature has been determined also from the average energy which was deduced from the age of epithermal neutrons in water measured by gold foil. The high σ values obtained by the low threshold reactions and the age method confirm the assumption on the excess of neutrons in the low energy region of the spectrum.     

Nuclear reactions induced by deuterons and their applicability to skin tumor treatment through BNCT

In this work the D(d,n)³He and ⁹Be(d,n)¹⁰B reactions have been studied in a low-energy regime as neutron sources for skin tumor treatment in the frame of accelerator-based BNCT (AB-BNCT). The total neutron production and the energy and angular distributions for each reaction at different bombarding energies and for the thick targets considered (TiD₂, Be) have been determined using the available data in the literature. From this information, a feasibility study has been performed by means of MCNP simulations. The thermal, epithermal and fast neutron fluxes and doses at skin tumor positions (loaded with 40 ppm ¹⁰B) which are located on a whole-body human phantom have been simulated for different D₂O moderator depths. The best-case performance shows that a high tumor control probability (TCP) of 99% corresponding to a weighted dose in tumor of 40 Gy can be reached at the tumor position keeping the weighted dose in healthy tissue below 12.5 Gy, by means of the ⁹Be(d,n)¹⁰B reaction at 1.1 MeV for a deuteron current of 20 mA and a 30 cm D₂O moderator in 52 min. The availability of low-energy neutrons in the ⁹Be(d,n)¹⁰B reaction from the population of excited levels between 5.1 to 5.2 MeV in ¹⁰B and the convenience of a thin beryllium target are discussed.As a complement concerning alternatives to the Li(metal) + p reaction, the neutron yield of refractory lithium compounds (LiH, Li₃N and Li₂O) were calculated and compared with a Li metal target.     

Measurement of Gamma-Ray Production Cross Sections of Iron for Incident Neutron Energies between 6 and 33 MeV

Measurement of differential γ-ray production cross sections, i.e. (n, x γ) cross sections, of Fe was made for neutron energies from 6 to 33 MeV. Neutrons used in the experiment were white neutrons produced with (p, n) reactions by 35 MeV protons using a thick Be target. The neutron energy was analyzed by the time-of-flight method and bunched into 3 MeV wide energy bins, for each of which the spectrum of secondary γ-rays produced in an Fe sample was measured by a BGO scintillator at an angle of 144° to the neutron beam direction.

The obtained (n, xγ) cross sections agreed well with other data and the evaluated data file of ENDF/B-IV at neutron energies below 15 MeV where data were existing. The JENDL-3 file overestimated the γ-ray spectra at γ-ray energies of 3 to 7 MeV. The present work newly provided the data in the neutron energy range above 20 MeV. The GNASH calculation made by Young reproduced the measured data fairly well even at these higher energies.     

Measurement of Double-Differential Neutron Emission Spectra from Uranium-238

We have performed the measurement of neutron emission spectra from ²³⁸U using a time-of-flight technique, and deduced the following data; (1) the prompt fission neutron spectra for 2 MeV incident neutrons at two emission angles of 90° and 135°, (2) the double-differential neutron emission cross sections at the incident energies of 1.2, 2.0, 4.2, 6.1 and 14.1 MeV. The emission spectra and the cross sections for scattering process were also deduced by subtracting the fission neutrons from the experimental spectra. The experimental results were compared with other experiments and the evaluations of JENDL-3 and ENDF/B-IV.

From the fission spectrum data ranging from 2 to 12 MeV, we have derived the best fit parameters for the Maxwellian and Watt type distribution functions. The experimental spectra are described with the Maxwellian spectrum with temperature of 1.24–1.26 MeV and are softer than both evaluations.

The spectra and cross sections for inelastic-scattering showed substantial disagreement with the evaluations concerning the discrete levels between 0.5 and 1.2 MeV, and continuum neutrons due to evaporation and pre-equilibrium processes. The secondary neutron angular distributions at 14 MeV incident energy were reproduced fairly well with the systematics.     

Neutron multiplication in 9Be: A stochastic approach

A method for the systematic derivation of the Multiplicity Spectrum (MS) of leakage neutrons from ⁹Be assemblies, starting from the forward Kolmogorov equation (familiar from stochastic neutron kinetics) is proposed. An exact solution to the multigroup Kolmogorov equation has been obtained using the method of Probability Generating Function (PGF). The effect of correlation between the two secondary neutrons in the (n, 2n) process on neutron multiplication has been investigated using this formalism. While the average multiplicity is not sensitive to correlation between the secondary neutrons in the (n, 2n) process, the MS for multiplicities ⩾ 5 is found to be highly sensitive. When correlation is accounted for the probability P(v) of leakage of v neutrons from a ⁹Be assembly differs by orders of magnitude from that without correlation for the higher multiplicities. As the correlation between the two neutrons contains valuable information, experimental study of the MS (especially the higher multiplicity events) can serve as a very powerful tool for the investigation of the (n, 2n) multiplicative process and can help resolve some of the observed discrepancies regarding the multiplication of 14 MeV neutrons in ⁹Be. The feasibility of the experimental determination of MS using a statistical correlation technique has been discussed in detail by Srinivasan (1985).     

Photo-Neutron Production in Heliotron-E

Photo-neutron emission in H₂ and He discharges was observed in the initial ohmic heating experiments of the Heliotron-E. Typical total neutron yield was 10⁹ neutrons per pulse under high level of runaway electrons (≧10 MeV). Neutron flux was localized near the limiters. Energy spectrum of neutrons was continuous up to about 2 MeV. The radioactive nuclides in the limiters and the vacuum chamber irradiated by runaway electrons showed that ⁵⁸Ni(γ, n) ⁵⁷Ni and ⁵³Cr(γ, n)⁵¹Cr reactions had occurred, proving that the photo-disintegration process was the source of neutron flux.     

Measurement of keV-neutron capture cross-sections for 164Dy

The neutron capture cross-sections of ¹⁶⁴Dy were measured in the neutron energy region of 10 to 90 keV using the 3-MV Pelletron accelerator of the Research Laboratory for Nuclear Reactors at the Tokyo Institute of Technology. Pulsed keV neutrons were produced from the ⁷Li(p,n)⁷Be reaction by bombarding a lithium target with the 1.5-ns bunched proton beam from the Pelletron accelerator. The incident neutron spectrum on a capture sample was measured by means of a TOF method with a ⁶Li-glass detector. Capture γ-rays were detected with a large anti-Compton NaI(Tl) spectrometer, employing a TOF method. A pulse-height weighting technique was applied to observed capture γ-ray pulse-height spectra to derive capture yields. The capture cross-sections were obtained by using the standard capture cross-sections of ¹⁹⁷Au. The present results were compared with the previous measurements and the evaluated values of ENDF/B-VI.