A structural evaluation of the tungsten isotopes via thermal neutron capture

Total radiative thermal neutron-capture $\gamma$-ray cross sections for the $^{182,183,184,186}$W isotopes were measured using guided neutron beams from the Budapest Research Reactor to induce prompt and delayed $\gamma$ rays from elemental and isotopically-enriched tungsten targets. These cross sections were determined from the sum of measured $\gamma$-ray cross sections feeding the ground state from low-lying levels below a cutoff energy, E$_{\rm crit}$, where the level scheme is completely known, and continuum $\gamma$ rays from levels above E$_{\rm crit}$, calculated using the Monte Carlo statistical-decay code DICEBOX. The new cross sections determined in this work for the tungsten nuclides are: $\sigma_{0}(^{182}{\rm W}) = 20.5(14)$ b and $\sigma_{11/2^{+}}(^{183}{\rm W}^{m}, 5.2 {\rm s}) = 0.177(18)$ b; $\sigma_{0}(^{183}{\rm W}) = 9.37(38)$ b and $\sigma_{5^{-}}(^{184}{\rm W}^{m}, 8.33 \mu{\rm s}) = 0.0247(55)$ b; $\sigma_{0}(^{184}{\rm W}) = 1.43(10)$ b and $\sigma_{11/2^{+}}(^{185}{\rm W}^{m}, 1.67 {\rm min}) = 0.0062(16)$ b; and, $\sigma_{0}(^{186}{\rm W}) = 33.33(62)$ b and $\sigma_{9/2^{+}}(^{187}{\rm W}^{m}, 1.38 \mu{\rm s}) = 0.400(16)$ b. These results are consistent with earlier measurements in the literature. The $^{186}$W cross section was also independently confirmed from an activation measurement, following the decay of $^{187}$W, yielding values for $\sigma_{0}(^{186}{\rm W})$ that are consistent with our prompt $\gamma$-ray measurement. The cross-section measurements were found to be insensitive to choice of level density or photon strength model, and only weakly dependent on E$_{\rm crit}$. Total radiative-capture widths calculated with DICEBOX showed much greater model dependence, however, the recommended values could be reproduced with selected model choices. The decay schemes for all tungsten isotopes were improved in these analyses.Comment: 25 pages, 15 figures, 15 table

Investigation of \u3csup\u3e186\u3c/sup\u3eRe via radiative thermal-neutron capture on \u3csup\u3e185\u3c/sup\u3eRe

Partial -ray production cross sections and the total radiative thermal-neutron capture cross section for the 185Re(n,)186Re reaction were measured using the Prompt Gamma Activation Analysis facility at the Budapest Research Reactor with an enriched 185Re target. The 186Re cross sections were standardized using well-known 35Cl(n,)36Cl cross sections from irradiation of a stoichiometric natReCl3 target. The resulting cross sections for transitions feeding the 186Re ground state from low-lying levels below a cutoff energy of Ec=746keV were combined with a modeled probability of ground-state feeding from levels above Ec to arrive at a total cross section of σ0=111(6)b for radiative thermal-neutron capture on 185Re. A comparison of modeled discrete-level populations with measured transition intensities led to proposed revisions for seven tentative spin-parity assignments in the adopted level scheme for 186Re. Additionally, 102 primary rays were measured, including 50 previously unknown. A neutron-separation energy of Sn=6179.59(5)keV was determined from a global least-squares fit of the measured -ray energies to the known 186Re decay scheme. The total capture cross section and separation energy results are comparable to earlier measurements of these values

Developments in Capture- γ Libraries for Nonproliferation Applications

The neutron-capture reaction is fundamental for identifying and analyzing the γ-ray spectrum from an unknown assembly because it provides unambiguous information on the neutron-absorbing isotopes. Nondestructive-assay applications may exploit this phenomenon passively, for example, in the presence of spontaneous-fission neutrons, or actively where an external neutron source is used as a probe. There are known gaps in the Evaluated Nuclear Data File libraries corresponding to neutron-capture γ-ray data that otherwise limit transport-modeling applications. In this work, we describe how new thermal neutron-capture data are being used to improve information in the neutron-data libraries for isotopes relevant to nonproliferation applications. We address this problem by providing new experimentally-deduced partial and total neutron-capture reaction cross sections and then evaluate these data by comparison with statistical-model calculations

Radiative-capture cross sections for the La139(n,γ) reaction using thermal neutrons and structural properties of La140

A set of prompt partial γ-ray production cross sections from thermal neutron capture were measured for the 139La (n,γ) reaction using a guided beam of subthermal (thermal and cold) neutrons incident on a nat La2O3 target at the Prompt Gamma Activation Analysis facility of the Budapest Research Reactor. Absolute 140La cross sections were determined relative to the well-known comparator 35Cl(n,γ) cross sections from the irradiation of a stoichiometric nat LaCl3 sample. The total cross section for radiative thermal neutron capture on 139La from the sum of experimentally measured cross sections observed to directly feed the 140 La ground state was determined to be σ0 = 8.58(50) b. To assess completeness of the decay scheme and as a consistency check, the measured cross sections for transitions feeding the ground state from levels below a critical energy of Ec = 285 keV were combined with a modeled contribution accounting for ground-state feeding from the quasi continuum to arrive at a total cross section of σ0 = 9.36(74) b. In addition, a neutron-separation energy of Sn = 5161.001(21) keV was determined from a least-squares fit of the measured primary γ-ray energies to the low-lying levels of the 140La decay scheme. Furthermore, several nuclear structure improvements are proposed for the decay scheme. The measured cross-section and separation-energy results are comparable to earlier measurements of these quantities

Pulsed ti: sapphire laser power amplifier

We have demonstrated an all solid state Ti:Sapphire laser system consisting of a power oscillator and single pass amplifier. The electrical-to-optical efficiency far exceeds that of the current CW systems. The pump lasers have lower capital and operating costs than the Argon-ion laser. In the future, we plan to scale the output power to higher levels by adding a fourth pump laser and improving the output power of the current pump lasers. Modeling results suggest that a large increase in efficiency can be realized by improving the beam quality of the pump lasers, even at the cost of reduced output power. We will explore this option by adding apertures to the cavity and/or reducing the rod diameter along with optimizing the resonator design. Other improvements in efficiency which will be investigated include double passing the amplifier for better extraction. To complete this work, the laser system will be converted into a two-stage amplifier. A narrow band, lower power oscillator currently under development will be injected in to the amplifier to study the extraction and efficiency characteristics of the amplifier throughout the tuning range of Ti:Sapphire. Detailed beam quality measurements will also be made. Other work will include doubling the narrow band output for materials processing applications

A COMPARISON BETWEEN PROPOSED SMALL MODULAR REACTORS AND EXISTING POWER REACTORS WITH REGARD TO SPENT FUEL NUCLEAR MATERIAL ATTRACTIVENESS A COMPARISON BETWEEN PROPOSED SMALL MODULAR REACTORS AND EXISTING POWER REACTORS WITH REGARD TO SPENT FUEL NUCLEAR M

ABSTRACT The nuclear material attractiveness of used fuel from proposed small modular reactors is evaluated relative to used fuel from the existing fleet of power reactors. Irradiated fuels at several burn-ups and cooling times are considered. The methodology for evaluating the materials attractiveness is based on previously used metrics and binning approaches and is consistent with the "attractiveness levels" that are normally reserved for nuclear materials in DOE nuclear facilities. Commercial power reactor fuels are unattractive at charge but may become attractive after discharge and age, depending upon the degree of burn-up, the fuel composition, and the reactor type. Some used Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) fuels in the US are over 40 years in age and their radiation dose rates continue to decline, calling into question the "self protecting" nature of these older used fuels. This study examines the attractiveness of used fuel assemblies from typical BWR 7x7, BWR 8x8, PWR 17x17, PWR-MOX 17x17, and VVER-440 reactors. A new generation of small modular reactor (SMR) designs promises a number of benefits relative to the existing fleet of commercial power reactors, including portability, viable initial investment level, scalability due to modularity, and improved security. The somewhat shorter length (and hence lighter weight) of SMR fuel assemblies along with the potential for greater decentralization are additional factors that need to be considered. Like commercial power reactors fuels, the two candidate SMR fuels are unattractive at charge, but may become attractive after discharge and age, depending upon the degree of burn-up, the fuel composition, and the reactor type. For all practical purposes the attractiveness of the used commercial power reactor fuels and used fuels from the two SMRs under consideration in the US are identical. The differences between the existing power reactors and the two proposed SMRs largely comes down to differences in fuel assembly size and facility characteristics. This study is consistent with previous studies that demonstrate the importance of ensuring tha