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.     

Experimental study of neutron multiplication in BeO by the shift-register coincidence technique

The multiplication of 14 MeV neutrons by Be, which is used as a neutron multiplier in the design of fusion reactor blankets, is of topical interest as T-breeding depends on this. Earlier investigations have shown that the experimental multiplication of 14 MeV neutrons in Be and BeO is 20–25% lower than the calculated values. Recently, a different method has been proposed to measure the leakage multiplication based on a statistical correlation technique. This note explores the applicability of such a technique to determine experimentally the leakage multiplication of Am-Be source neutrons of average energy 4.5 MeV in a BeO assembly. The experimental method uses a high-level neutron coincidence counter (HLNCC) equipped with a shift-register coincidence logic unit. The deduced value of neutron leakage multiplication for 5 and 10 cm thick BeO in near rectangular geometry is 1.12 ± 0.03 and 1.15 ± 0.03, respectively.     

Evaluation of reactivity bonus due to (n, 2n) multiplication in Be/BeO-reflected 233U uranyl nitrate solution systems

Beryllium experiences (n, 2n) multiplication reaction for neutrons having energy above the threshold of ≈ 1.8 MeV. In the case of beryllium-reflected/moderated reactor assemblies these (n, 2n) neutrons contribute additional reactivity. This paper evaluates the reactivity contribution due to (n, 2n) reactions in the Be and BeO reflector of reflected ²³³U uranyl nitrate solution systems in spherical geometry in the light of the latest ⁹Be cross-section data. The results shows that the (n, 2n) reactions in a given thickness of reflector are directly proportional to the core leakage and thus very much a function of core radius. In the core radius range of 4–15 cm the reactivity bonus due to (n, 2n) reactions in a 30-cm thick Be reflector varies from 6.2 to 3.0%, which corresponds to a critical mass saving of ≈18–9%. The corresponding reactivity contributions for BeO reflected systems are ≈25% lower.     

Sensitivity studies of the neutron multiplicity spectrum in the spallation of Pb targets

The number of neutrons produced per incident proton in the spallation of Pb targets is of direct relevance to the design of accelerator breeders. The nuclear cascade initiated by high-energy protons in spallation targets is usually described by an intranuclear cascade evaporation (INCE) model. Even though this model describes various average nuclear properties of spallation targets fairly well, differential quantities such as energy spectra, angular spectra etc. are not reproduced within the limits of experimental uncertainty. One of the reasons for this is the uncertainty in the magnitude of the parameters involved in the model, notably the level density parameter B₀ whose magnitude is quoted by different workers to be in the range of 8–20 MeV. The accuracy of B₀ could be improved if we could experimentally determine a quantity which is much more sensitive to B₀ than the average neutron yield. In this paper we discuss one such quantity, namely the neutron multiplicity spectrum (MS). We compute the MS due to the spallation of Pb targets of different sizes at proton energies of 1.5, 1.0 and 0.59 GeV using the Monte Carlo code HETC. It is noticed that for the 1.5 GeV proton case the probability P(v) for leakage of v neutrons for v in the range of 60–65, changes by about 70% when B₀ is varied from 8 to 20 MeV. The corresponding change in the average neutron yield is <20%. It is therefore suggested that an accurate measurement of the MS can serve as a useful tool to narrow down the range of uncertainty in the B₀ parameter.     

Investigation of the infinite medium integral neutronic data for incident fusion source neutrons

In the present work, the physical behavior of integral data in infinite medium has been evaluated for incident fusion neutrons with the help to the 3-D Monte Carlo code. In a fusion reactor blanket with finite dimension, the integral quantities will be more or less different from the infinitive medium results, depending on the neutron leakage fraction. Design studies foresee the reduction of the neutron leakage out of the blanket as possible in order to prevent the nuclear heating in super conducting fusion magnets and to keep all neutrons primarily in the coolant. The most important materials in fusion technology, namely tritium, beryllium, lead, thorium, and uranium have been investigated in infinitive medium. The main purpose of this work is to calculate the integral tritium breeding ratio, ²³³U breeding rate, ²³⁹Pu breeding rate, heat release, neutron multiplication ratio through (n,x) and fission (when applicable) reactions in those mixtures which are composed when first UO₂ and ThO₂ are mixed with natural lithium (Nat.Li) or ⁶Li for a volume fraction from 0 to 100%. Then the variable UO₂-Nat.Li (UO₂ mixed with Nat.Li) and UO₂-⁶Li (UO₂ mixed with ⁶Li) compositions will be mixed with Beryllium (Be) and Lead (Pb) for a volume fraction from 0 to 100%. However, the variable TO₂-Nat.Li (ThO₂ mixed with Nat.Li) and ThO₂-⁶Li (ThO₂ mixed with ⁶Li) compositions will be mixed with Be and Pb for a volume fraction mentioned above.     

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.     

Experimental prompt fission neutron “sawtooth” data described by the “point by point” model

The experimental data concerning the prompt fission neutron number as function of the fission fragment mass (currently named “sawtooth” data) when they exist, allow a more refined verification of the “point by point” model and in the same time the validation of the methods used to determine the model parameters corresponding to the fission fragment pairs covering the entire fission fragment range.The prompt neutron multiplicities of each fission fragment pair provided by the “point by point” model are compared with the experimental data concerning the sum of neutrons emitted by the light and heavy fragments forming a pair obtained from experimental sawtooth data.The available fission fragment pair multiplicity and sawtooth experimental data for the ^233,235,238U(n, f) and ²³⁷Np(n, f) reactions as well as for the spontaneous fission of ²⁵²Cf are well described by the “point by point” model results, proving again that this treatment is a powerful tool for prompt fission neutron multiplicity and spectrum evaluation purposes.     

Preliminary spectrum shifter design for intermediate energy nuclear data benchmark experiments with DT neutrons

Integral benchmark experiments with DT neutrons are not always sufficient for nuclear data benchmarking in the MeV region, below 10 MeV. A neutron spectrum shifter, which will be placed between a sample and a DT neutron source, is effective to moderate DT neutrons incident to the sample. In order to estimate effects of the spectrum shifter, the ratio of the contribution of 14 MeV neutrons in the leakage neutron and gamma-ray spectra was calculated with MCNP-4C for an experimental configuration at FNS of JAEA, Japan. The calculations were carried out for a Li₂TiO₃ sample with a Be, D₂O, or ⁷LiD spectrum shifter. It was found out that the Be shifter was superior to others and the Be shifter was effective to decrease the contribution of 14 MeV neutrons especially for secondary gamma-ray spectrum measurements.     

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.     

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.     

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.     

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.     

Electron paramagnetic resonance dose response studies for neutron irradiated human teeth

The dosimetric response of neutron irradiated human tooth enamel has been investigated using electron paramagnetic resonance (EPR) dosimetry. Continuous energy fast neutrons of mean energy less than 450 keV were produced from the McMaster University 3 MV K.N. Van de Graaff accelerator employing a thick lithium target via ⁷Li(p,n)⁷Be interaction. Prior to its use for various experiments, the gamma dose contamination of the neutron beams was determined at the selected proton beam energies using the tissue-equivalent proportional counter (TEPC). The neutron sensitivity (/Gy-100 mg) of human tooth enamel remained constant for various mean neutron energies ranging from 167 to 450 keV. Similarly, the EPR signal intensity remained independent of the neutron dose rate variation from 0.5 to 2.4 Gy/h.     

Measurements of Double-Differential Neutron Emission Cross Sections of 6Li, 7Li and 9Be for 11.5 MeV and 18.0 MeV Neutrons

Double-differential neutron emission cross sections (DDXs) of ⁶Li, ⁷Li and ⁹Be were measured for 18.0 MeV and 11.5 MeV incident neutrons produced by the T(d, n) and ¹⁵N(d, n) reactions respectively, using the Tohoku University Dynamitron time-of-flight (TOF) spectrometer. The data were obtained at 13 laboratory angles, and angular-differential cross sections (ADXs) of elastic and inelastic scattering neutrons were derived from the DDXs. For 11.5 MeV neutrons, we obtained the neutron emission spectra over the secondary neutron energies by newly employing the double TOF method as well as the conventional one. In the measurements at 18.0 MeV, we achieved better energy resolution than in our previous studies by using a neutron detector that has a larger solid angle and a thinner tritium target. The experimental results of DDXs and ADXs were compared with our previous results and the evaluated data given in JENDL-3.2, JENDL Fusion File and ENDF/B-VI. It is found that the JENDL data reproduce the experimental ones very well.     

Analysis of (n, 2n) multiplication in lead

Lead is being considered as a possible amplifier of neutrons for fusion blankets. A simple one-group model of neutron multiplications in Pb is presented. Given the 14 MeV neutron cross section on Pb, the model predicts the multiplication. Given measured multiplications, the model enables the determination of the (n, 2n) and transport cross sections. Required for the model are: P—the collision probability for source neutrons in the Pb body—and W—an average collision probability for non-virgin, non-degraded neutrons. In simple geometries, such as a source in the center of a spherical shell, P and an approximate W can be expressed analytically in terms of shell dimensions and the Pb transport cross section. The model was applied to Takahashi's measured multiplications in Pb shells in order to understand the apparent very high multiplicative power of Pb. The results of the analysis are not consistent with basic energy-balance and cross-section magnitude constraints in neutron interaction theory.     

<Superscript>239</Superscript>Pu prompt fission neutron spectra

Calculations of ²³⁹Pu(n, F) prompt fission neutron spectra have been performed for neutron energy up to 20 MeV. The exclusive spectra of pre-fission neutron reactions (n, xnf) were calculated on the basis of the Hauser-Feshbach model simultaneously with the cross sections of (n, F) and (n, 2n) reactions. The spectra of neutrons emitted by fission fragments were approximated by a sum of two Watt distributions. The components of the prompt fission neutron spectra due to pre-fission neutrons are manifested in the prompt fission neutron spectra and the average neutron energy. A correlation is established between this effect in the contribution of emissive fission (n, xnf) in the fission cross-section of ²³⁹Pu(n, F) and ²³⁵U(n, F). It is shown that the ²³⁹Pu(n, F) prompt fission neutron spectra used in applied calculations do not correspond to the experimental differential data and the systematic regularities in the spectra and their average energy found for the most carefully studied nuclei ^235,238U and ²³²Th. __________ Translated from Atomnaya énergiya, Vol. 103, No. 2, pp. 119–124, August, 2007.     

Adsorption of Radiostrontium on “Calcium Phosphate-Anion Exchange Resin”

Making use of monoenergetic neutrons produced by a 2 MeV Van de Graafif accelerator, the cross section for the ¹⁰³Rh (n.n′) ^103mRh reaction was measured in the energy range from 180 keV to 4.6 MeV. To obtain the relative activity of ^103mRh, the 20 keV X-rays emitted by the nuclide were measured with a thin NaI(Tl) scintillator with Be window. For absolute measurement, a standard ¹⁰³Ru source was made from ruthenium trichloride irradiated with thermal neutrons. The self-shielding effect to the 20keV X-rays was eliminated by the realization of a thin layer of Ru obtained by electrodeposition.

The fission-spectrum-averaged cross section for this reaction determined by the present method is 558±32 mb, and the effective cross section 960±55 mb.

The reaction was applied to the measurement of the fast neutron flux in the experimental facilities of the Kyoto University Reactor. The present report also gives a discussion on the advantage of using the ¹⁰³Rh(n.n′) ^103mRh reaction as fast neutron radiation damage monitor.     

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.     

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.     

Neutron dose transmission measurements for several new concrete samples including colemanite

We have prepared four baryte and four concrete samples having respectively 0%, 5%, 10% and 15% colemanite concentrations. Neutron dose transmission measurements have been done by using a source of mono-energetic neutrons (E_eff = 4.5 MeV–²⁴¹Am–Be). It has been shown that when colemanite percentages of the samples increase, neutron dose transmission values for the samples decrease. It is thus possible to enhance the neutron shielding property of baryte and ordinary concrete by adding different proportions of colemanite.