R.T. Perry

Operator declaration verification technique for spent fuel at reprocessing facilities

Abstract A verification technique for use at reprocessing facilities, which integrates existing technologies to strengthen safeguards through the use of environmental monitoring, has been developed at Los Alamos National Laboratory. This technique involves the measurement of isotopic ratios of stable noble fission gases from on-stack emissions during reprocessing of spent fuel using high-precision mass spectrometry. These results are then compared to a database of calculated isotopic ratios using a data analysis method to determine specific fuel parameters (e.g., burnup, fuel type, reactor type, etc.). These inferred parameters can be used to verify operator declarations. The integrated system (mass spectrometry, reactor modeling, and data analysis) has been validated using on-stack measurements during reprocessing of fuel from a US production reactor. These measurements led to an inferred burnup that matched the declared burnup to within 3.9%, suggesting that the current system is sufficient for most safeguards applications. Partial system validation using gas samples from literature measurements of power reactor fuel has been reported elsewhere. This has shown that the technique developed here may have some difficulty distinguishing pressurized water reactor (PWR) from boiling water reactor (BWR) fuel; however, it consistently can distinguish light water reactor (either PWR or BWR) fuels from other reactor fuel types. Future validations will include advanced power reactor fuels (such as breeder reactor fuels) and research reactor fuels as samples become available.

Comparisons of Calculated and Measured 237Np, 241Am, and 243Am Concentrations as a Function of the 240Pu/239Pu Isotopic Ratio in Spent Fuel

The Los Alamos National Laboratory (LANL) has developed a system for determining 237 Np, 241 Am, and 243 Am concentrations in spent fuel from measurements of the 240 Pu/ 239 Pu isotopic ratio using calculations performed with the HELIOS lattice-physics code. Benchmark calculations for several pressurized water reactors (PWRs) were performed and compared to measured values from the literature for fuels with burnups ranging from 0 to 50000 MWd/tonne U. A direct correlation can be found between the 240 Pu/ 239 Pu isotopic ratio and the higher-actinide concentrations for each fuel type. Comparisons of calculated with measured values suggests that the LANL technique would yield 237 Np and 241 Am concentrations within±5% and 243 Am concentrations within ±15% for PWRs. Expanding this system for all reprocessing applications will require more measured data (especially for boiling water reactors and VVER-type reactors), but the existing results show a marked improvement over the previous ORIGEN calculations. Also, a better determination of the 243 Am concentrations may support a greater confidence in the calculated results or suggest an alteration to the existing nuclear data. The present state ofthese neutronics calculations suggests that the technology exists to reduce the need for direct measurement of the 237 Np, 241 Am, and 243 Am concentrations in spent fuel.

Noble‐gas atmospheric monitoring for international safeguards at reprocessing facilities

Environmental monitoring of nuclear activities promises to play a large role in the improvements in international safeguards under the International Atomic Energy Agencys Programme 93+2. Monitoring of stable noble‐gas (Kr, Xe) isotopic abundances at reprocessing plant stacks appears to be able to yield information on the burnup and type of the fuel being processed. To estimate the size of these signals, model calculations of the production of stable Kr and Xe nuclides in reactor fuel and the subsequent dilution of these nuclides in the plant stack are carried out for two case studies: reprocessing of PWR fuel with a burnup of 35 GWd/tU, and reprocessing of CANDU fuel with a burnup of 1 GWd/tU. For each case, a maximum‐likelihood analysis is used to determine the fuel burnup and type from the isotopic data.

Benchmarking of the Los Alamos neutron production rate code SOURCES-3A

Abstract The neutron production rate code SOURCES-3A was benchmarked using various experimental measurements from the literature. These experiments included thick-target yield measurements from Li, Be, B, C, O, F, Mg, Al, Si, UO 2 , and UC. Several integrated experiments (PuBe 13 and UO 2 F 2 homogeneous problems, Po–B 4 C and Po–Be interface problems, and Al 2 O 3 and SiO 2 beam problems) were also modeled, testing all the geometry characteristics of the SOURCES-3A code. The SOURCES-3A calculations were compared with the experimental results and good agreement was found in all cases. These benchmarks have shown that SOURCES-3A spectra and magnitude calculations are accurate to within ±18% for even the most complex problems.

Assessment of a MOX Fuel Assembly Design for a BWR Mixed Reload

Reprocessing benefits are still being debated from the standpoint of economy. However, it is a clear option to reduce the amount of depleted fuel assemblies and a reduction of the reactor plutonium inventories. Several mixed-oxide (MOX) fuel concepts have been considered as an option for mixed-fuel reload assemblies in boiling water reactors in the past. In this work, a new MOX fuel assembly design is proposed. The design is based on the use of a proportional fissile ratio between equivalent fissile plutonium (239Pu + 241Pu) and fissile uranium (235U). This is referred to as the PUF ratio. Furthermore, the moderation ratio will be increased in the assembly as a way to reduce the possible impact of using MOX fuel on the reactivity control systems. The design and performance of the MOX fuel assembly and the mixed core are presented and discussed. The new design, for the cases considered, can increase the MOX batch reload up to 52 MOX assemblies, in comparison with the 24 assemblies from a design that does not increase the moderation ratio. The use of the combined PUF ratio and increased moderation ratio for the MOX assembly allows for a reduction in the average enrichment of fissile plutonium to 4.68 wt%, instead of the 6.75 wt% necessary without increasing the moderation ratio. Both MOX designs produce the same amount of energy during the proposed cycle length and satisfy the same thermal limits. Some comparisons are performed between the core with this MOX fuel assembly and the core that uses only standard uranium assemblies.

Comparisons of HELIOS, ORIGEN2, and Monteburns Calculated 241 Am and 243 Am Concentrations to Measured Values for PWR, BWR, and VVER Spent Fuel

A study was performed at Los Alamos National Laboratory to explore the accuracy of several reactor analysis codes in calculating 241 Am and 243Am concentrations in light water reactor spent fuel. Calculated higher-actinide concentrations were compared to measured values from the literature for three reactor fuels. The fuel samples were taken from the Mihama Unit 3 pressurized water reactor, the Garigliano boiling water reactor, and a VVER-440. The 241Am and 243Am concentrations were calculated using the HELIOS-1.4 lattice-physics code, the ORIGEN2 burnup code, and a linked MCNP/ORIGEN2 code named Monteburns 3.01. Comparisons were made between the calculated and measured values. It was determined that all codes performed consistently well for the Mihama Unit 3 measurements (within ±5% for 241Am and ±20% for 243Am) and the Garigliano measurements (within ±12% for 241 Am and ±20% for 243Am). It was determined that the ORIGEN2 pressurized water reactor libraries are insufficient for the VVER-440 measurements. The HELIOS and MONTEBURNS codes both demonstrated good ability to calculate these isotopes for VVER-440 fuel (±10% for 241Am and ±12% for 243Am). The accuracies of these codes and the associated radiochemical measurements of these higher-actinide isotopes may be insufficient for safeguards and fuel management purposes; thus, development of new methods and modification to existing data libraries may be necessary in order to enable cost-effective safeguarding of these higher-actinide materials.

Calculated Actinide and Fission Product Concentration Ratios for Gaseous Effluent Monitoring Using Monteburns 3.01

Techniques have been developed at Los Alamos National Laboratory for accurately calculating certain spent-fuel isotope concentration ratios for pressurized water reactor assemblies using a linked MCNP/ORIGEN2 code named Monteburns 3.01, without resorting to an assembly or full-core calculation. The effects of various fuel parameters such as the number of radial fuel regions per pin, burnup step size, reactor power, reactivity control mechanisms, and axial profiles have been studied. The significance of each factor was determined. A method was also proposed for calculating spent-fuel inventories as a function of burnup for a wide range of reactors and fuel types. It was determined that accurate calculations can be obtained using a three-dimensional, modified pin cell with seven radial fuel regions and two (flat-flux) axial fuel regions calculated with 2000 MWd/tonne U burnup steps for burnups ranging from 0 to 50 000 MWd/tonne U. The calculational technique was benchmarked to measured values from the Calvert Cliffs Unit 1 reactor, and good agreement from the point of view of calibrating a monitoring instrument was found for most cases.

Derivation and Implementation of the Three-Region Problem in the SOURCES Code System

The SOURCES code system was designed to calculate neutron production rates and spectra in materials due to the decay of radionuclides [specifically from (α,n) reactions, spontaneous fission, and delayed neutron emission]. The current version (SOURCES-3 A) is capable of calculating (α,n) source rates and spectra for three types of problems: homogeneous materials, interface problems, and beam problems. Recent interest in (α,n) sources has prompted the development of a fourth scenario: the three-region problem. To allow SOURCES to confront this problem, the relevant equations defining the α-particle source rates and spectra at each interface and the neutron source rates and spectra per unit area of interface were derived. These equations (in discretized form) were added as a new subroutine to the SOURCES code system (dubbed SOURCES-4A). The new code system was tested by analyzing the results for a simple three-region problem in two limits: with an optically thin ‘intermediate region’ and with an optically thick ‘intermediate region.’ To further validate the code system, SOURCES-4A will be experimentally benchmarked as measured data becomes available.

Neutron sources and spectra from cold fusion

Deterministic methods are used to calculate the neutron and photon sources and spectra that would develop if fusion reactions were occurring in cold fusion experimental devices. The results from the calculations give the neutron and gamma spectra resulting from a 2.45-MeV and a 14.1 MeV neutron source. The neutron source strength from certain (gamma,n) and (alpha,n) reactions are also determined.