Fission Product Nuclides Research Articles

Deterministic Assessment for Environmental Impact of Yucca Mountain Repository Measured by Radiotoxicity

Environmental impact of the Yucca Mountain Repository (YMR) has been quantitatively and deterministically evaluated in terms of the radiotoxicity of transuranic (TRU) and fission-product radionuclides existing in the environment after released from failed packages. Inventory abstraction has been made based on the data published in Final Environmental Impact Statement by US Department of Energy (DOE). Mathematical model and computation code have been developed based on analytical solutions. Environmental impact from the commercial spent nuclear fuel (CSNF) packages is about 90% of the total impact including the contribution from defense waste (DW) packages. Impacts due to isotopes of Cm, Am, Pu and Np, and their decay daughters are dominant, compared with those from fission-product nuclides. Numerical results show that reduction of the TRU nuclides by a factor of 100 makes the impact from CSNF smaller than that from DW.

Environmental Impact of Yucca Mountain Repository in the Case of Canister Failure

Mathematical models and a computation code have been developed for total release of transuranic (TRU) and fission product radionuclides from waste packages in the Yucca Mountain Repository (YMR) into the surrounding geosphere in the case of simultaneous package failure. The total amount of these radionuclides in the geosphere, which is called the environmental impact in this paper, has been expressed in terms of radiotoxicity. Inventory abstraction has been made, based on the data provided in the Final Environmental Impact Statement published by the U.S. Department of Energy. Various types of waste packages in the YMR have been abstracted into commercial spent nuclear fuel (CSNF) and defense waste. For defense waste, co-disposal and naval spent fuel have been abstracted separately. Numerical results show that within the total environmental impact, contribution from the defense waste packages is about 10%, which is close to the fraction of the repository capacity allocated for defense waste. Impacts due to isotopes of TRU and their decay daughters are dominant, compared with those from fission product nuclides. If the mass of TRU nuclides to be disposed of in the repository were reduced by a factor of 100, the impact from CSNF would become smaller than that from defense waste.

Reconsideration of the Theoretical Supplementation of Decay Data in Fission Product Decay Heat Summation Calculations

In the early stage of the development of the fission product (FP) decay data library for decay-heat summation calculations a serious disagreement between calculations and sample-irradiation measurements was experienced world wide. This problem was essentially circumvented by introduction of the nuclear model calculations. By applying the recent results using the total absorption gamma-ray spectrometer to the decay heat calculations we found that TAGS results seem to be free from the pandemonium problem and, in this respect, it provides a solid basis of the summation calculations without the supplementation of the nuclear model calculations. Among the typical FP nuclides which deserve the future TAGS experiment there are 98Nb, 100Nb, 105Mo and 102Tc and others from the present survey.

Dissolution of Ce, Zr and La-containing magnetites and nickel ferrite in citric acid-EDTA-gallic acid formulation

It has been shown by us earlier that gallic acid (GA) can be used as a reductant in dilute chemical decontaminant formulations containing Ethylene Diamine Tetra Acetic acid (EDTA) as chelant. The results on the dissolution of magnetite in such a formulation were promising. Moreover, the superior radiation stability of gallic acid vis-à-vis other reductants such as ascorbic acid (AA) or oxalic acid (OA) is another plus point for this formulation. Besides having an inherent stability against radiation degradation, it is able to protect even EDTA against radiation-induced decomposition to a great extent unlike the case of ascorbic acid. In an extension of that work, dissolution experiments have now been carried out on nickel ferrite and magnetites containing fission product nuclides such as La, Ce and Zr. This is to simulate the presence of fission product oxides in magnetite resulting from a possible phase of operation with leaky fuel. The rate constants have been determined using both the inverse cubic rate law and also a general kinetic law model. Both models predicted the same trend in the behaviour of dissolution. They showed linear behaviour in their rate plots up to 80–90% of dissolution. Typically using the ICR model in the case of nickel ferrite, although there is an initial induction period, the rate constants for the dissolution were determined from the ICR model to be 0.016 and 0.0036 min −1 at 353 and 333 K, respectively. The presence of Ce in particular either alone or in combination with Zr/La at a level of about 0.5 at.% equivalent each in magnetite is seen to increase the surface area of the oxide. The rate constants (typical ICR model) for the dissolution at 353 K in a 11:44:4 mM citric acid-EDTA-gallic acid (CEG) formulation taken with magnetite and Ce, Zr and La-containing magnetites equivalent to yielding 22 mM Fe upon complete dissolution are as follows: 0.051 min −1 (magnetite), 0.071 min −1 (Ce 2O 3 containing magnetite), 0.063 min −1 (Ce 2O 3, ZrO 2 containing magnetite), 0.1 min −1 (Ce 2O 3, ZrO 2, La 2O 3 containing magnetite). The presence of Ce, Zr and La at ≈0.5 at.% level each has not resulted in any turbidity in solution at the end of magnetite dissolution suggesting a chemical dissolution of the oxides of rare earth elements and Zr in the formulation. Simple magnetite (worth 0.4 mM Fe upon complete dissolution) taken in a 1.4:1.4:1.7 CEG formulation yielded a rate constant of 0.076 min −1. It is concluded that the presence of rare earths in magnetite at low amounts has in fact improved the dissolution characteristics of magnetite in CEG formulation. A rate constant value of 0.1 min −1 was obtained for Ce,Zr,La-containing magnetite in 11:44:4 mM CEG as against a reported value of 0.031 min −1 in 11:44:4 mM CEA fomulation. Thus, CEG formulation appears to be better than CEA formulation for the dissolution of magnetite. Apart from better kinetics of dissolution, the major advantage of CEG formulation over CEA formulation seems to be its stability towards radiation-induced decomposition, a consideration important when chemical decontamination is planned under a short shutdown period (a few days) of the reactor.

CENDL-3 – Chinese Evaluated Nuclear Data Library, Version 3

CENDL-3 includes 214 nuclides for general purpose. Among them, the data of 169 nuclides were newly evaluated or reevaluated. Comparing to CENDL-2.1 and other evaluated libraries, the data were updated and improved. For structure materials, the data were given for both natural element and their isotopes, and are consistent between them. For fission product nuclides, the data were completely newly evaluated. For light nuclides, the data of double differential cross section have been made much improved. The data are being tested for some thermal, fast reactor assembles and some leak spectrum experiments. Some problems have been found and are being improved.

Development of JENDL FP Decay Data File 2000

JENDL FP Decay Data File 2000 has been developed as a special purpose file of JENDL. The file contains various decay data of fission product nuclides needed for the analyses of FP accumulation, decay heat, aggregate beta- and gamma-ray spectra and so on in a nuclear reactor. The number of unstable nuclides including isomeric states is 1087.

Determination of Gamma-Ray Exposure Rate from Short-Lived Fission Products under Criticality Accident Conditions

For the assessment of γ-ray doses from short-lived fission products (FPs) under criticality accident conditions, γ-ray exposure rates varying with time were experimentally determined in the Transient Experiment Critical Facility (TRACY). The data were obtained by reactivity insertion in the range of 1.50 to 2.93$. It was clarified from the experiments that the contribution of γ-ray from short-lived FPs to total exposure during the experiments was evaluated to be 15 to 17%. Hence, the contribution cannot be neglected for the assessment of γ-ray doses under criticality accident conditions. Computational analyses also indicated that γ-ray exposure rates from short-lived FPs calculated with the Monte Carlo code, MCNP4B, and photon sources based on the latest FP decay data, the JENDL FP Decay Data File 2000, well agreed with the experimental results. The exposure rates were, however, extremely underestimated when the photon sources were obtained by the ORIGEN2 code. The underestimation is due to lack of energy-dependent photon emission data for major short-lived FP nuclides in the photon database attached to the ORIGEN2 code. It was also confirmed that the underestimation arose in 1,000 s or less of time lapse after an initial power burst.

Determination of Gamma-Ray Exposure Rate from Short-Lived Fission Products under Criticality Accident Conditions.

For the assessment of γ-ray doses from short-lived fission products (FPs) under criticality accident conditions, γ-ray exposure rates varying with time were experimentally determined in the Transient Experiment Critical Facility (TRACY). The data were obtained by reactivity insertion in the range of 1.50 to 2.93$. It was clarified from the experiments that the contribution of γ-ray from short-lived FPs to total exposure during the experiments was evaluated to be 15 to 17%. Hence, the contribution cannot be neglected for the assessment of γ-ray doses under criticality accident conditions. Computational analyses also indicated that γ-ray exposure rates from short-lived FPs calculated with the Monte Carlo code, MCNP4B, and photon sources based on the latest FP decay data, the JENDL FP Decay Data File 2000, well agreed with the experimental results. The exposure rates were, however, extremely underestimated when the photon sources were obtained by the ORIGEN2 code. The underestimation is due to lack of energy-dependent photon emission data for major short-lived FP nuclides in the photon database attached to the ORIGEN2 code. It was also confirmed that the underestimation arose in 1,000 s or less of time lapse after an initial power burst.

New Method for Calculating Aggregate Fission Product Decay Heat with Full Use of Macroscopic-Measurement Data

We propose a new “hybrid” method for calculating the aggregate decay heat from fission product nuclides after a fission burst. The decay heat from a given fissioning system is expressed as a linear combination of macroscopic-measurement data for other fissioning systems with a small residual term. This method is based on the linearity of the decay heat to the fission yield. The coefficients in the linear combination are obtained from fitting the fission yield of the given fissioning system with a linear combination of fission yields of other fissioning systems. To demonstrate usefulness of this method, it is applied to examining the consistency among measured decay heat powers of five fast and three thermal neutron induced fissions. The hybrid-method calculations agree well with the measurements and usual summation calculations at cooling times before 4,000 s, except for a γ component measurement of the 235U thermal fission at about 2,000 s. These results indicate the consistency and reliability of the decay heat evaluation for these systems with the above exception. Furthermore, they also imply usefulness of the present method in predicting the decay heat of other fissioning systems, for which no measurements have been performed so far.

Estimation of Beta- and Gamma-Ray Spectra for JENDL FP Decay Data File

Beta- and gamma-ray spectra of fission product nuclides were estimated by a theoretical model for the nuclides having no or incompletely measured spectral data. The estimation was performed with Gross Theory of Beta Decay. The estimated spectra were so stored in JENDL FP Decay Data File 2000 as to keep the consistency between the average decay energy value derived from the spectrum and that used for decay heat analysis. The JENDL spectra of FP nuclides and the aggregate beta- and gamma-ray spectra of fission product nuclides calculated using the JENDL spectra showed good agreement with measured data.

Numerical solution of the chain model of fission product nuclides. Effect on the reactivity of the reactor

Runge–Kutta algorithm of fourth order devised for the numerical solution of ordinary differential equations of fuel isotopic composition and fission products accumulation systems is described. Several hundreds fission product isotopes are included in partially depleted fuel as a result of direct yield from nuclear fission reaction, radioactive decay of fission products and/or neutron capture. The algorithm of the suggested method was coded and compared with both the analytical method and the international ORIGEN2 code system. The accuracy and speed of the calculations as obtained by implementing the method in a PCs with Linux system are found to be acceptable and faster than their resemblance. The effect of the calculated fission product concentrations on the reactivity of the reactor has been included.

Re-evaluation of Nuclear Data of Fission Product Nuclides for JENDL-3.2

Re-evaluation was performed for the neutron-induced reaction data of fission product nuclides compiled in the Japanese Evaluated Nuclear Data Library, JENDL-3.1. The nuclides whose data were re-evaluated are 63 nuclides from 75As to 155Eu. Main quantities revised are the resonance parameters, capture cross sections and inelastic scattering cross sections. As the results of re-evaluation, the thermal neutron capture cross sections, the resonance integrals and the capture cross sections in the keV region were largely improved. The evaluated data cover the neutron energy region from 10∼5 eV to 20MeV. The present results have been adopted to JENDL-3.2.

Estimation of Beta- and Gamma-Ray Spectra for JENDL FP Decay Data File.

Beta- and gamma-ray spectra of fission product nuclides were estimated by a theoretical model for the nuclides having no or incompletely measured spectral data. The estimation was performed with Gross Theory of Beta Decay. The estimated spectra were so stored in JENDL FP Decay Data File 2000 as to keep the consistency between the average decay energy value derived from the spectrum and that used for decay heat analysis. The JENDL spectra of FP nuclides and the aggregate beta- and gamma-ray spectra of fission product nuclides calculated using the JENDL spectra showed good agreement with measured data.

Sustained nuclear energy without weapons or reprocessing using accelerator-driven systems

Accelerator-driven thermal-spectrum molten-salt nuclear technology can greatly simplify nuclear energy technology by eliminating reprocessing and greatly enhancing once-through burn-up. In effect the accelerator may be employed as a substitute for frequent reprocessing and recycle. The accelerator makes possible reduction in plutonium and minor actinides from current LWRs by a factor of more than ten without reprocessing while converting the plutonium remnant to a non-weapons-useful isotopic composition. The accelerator also enhances the once-through energy production from fertile material by a factor of ten without reprocessing compared to once-through LWR technology. This technology would eliminate the need to deploy plutonium production indefinitely, and reprocessing and recycle for at least several hundred years. The energy production technology proposed here operates primarily on the Th-U cycle with a minor contribution from the U-Pu cycle to eliminate the weapons-usefulness of 233 U. There are two key innovations in addition to the accelerator. One is the use of liquid fuel flowing once through a pool of material undergoing fission thereby allowing high burn-up concurrently with continuous removal of fission product without reprocessing. The second is the unanticipated low capture cross section of fission product nuclides which substantially enhances the neutron economy in this type of system. The supplement of neutrons from the accelerator, the reduced fission product neutron capture, and the continuously flowing fuel are the enablers for the performance described here. This technology allows an essentially complete decoupling of nuclear energy from nuclear weapons.

Fission Product Model for BWR Analysis with Improved Accuracy in High Burnup

A new fission product (FP) chain model has been studied to be used in a BWR lattice calculation. In attempting to establish the model, two requirements, i.e. the accuracy in predicting burnup reactivity and the easiness in practical application, are simultaneously considered. The resultant FP model consists of 81 explicit FP nuclides and two lumped pseudo nuclides having the absorption cross sections independent of burnup history and fuel composition. For the verification, extensive numerical tests covering over a wide range of operational conditions and fuel compositions have been carried out. The results indicate that the estimated errors in burnup reactivity are within 0.1Δk for exposures up to 100GWd/t. It is concluded that the present model can offer a high degree of accuracy for FP representation in BWR lattice calculation.

Fission Product Model for BWR Analysis with Improved Accuracy in High Burnup.

A new fission product (FP) chain model has been studied to be used in a BWR lattice calculation. In attempting to establish the model, two requirements, i.e. the accuracy in predicting burnup reactivity and the easiness in practical application, are simultaneously considered. The resultant FP model consists of 81 explicit FP nuclides and two lumped pseudo nuclides having the absorption cross sections independent of burnup history and fuel composition. For the verification, extensive numerical tests covering over a wide range of operational conditions and fuel compositions have been carried out. The results indicate that the estimated errors in burnup reactivity are within 0.1Δk for exposures up to 100GWd/t. It is concluded that the present model can offer a high degree of accuracy for FP representation in BWR lattice calculation.

Measurements and Analyses of Reactivity Effect of Fission Product Nuclides in Epithermal Energy Range

Experimental data usable for evaluating cross sections of main fission product elements (Rh, Cs, Nd, Sm, Eu and Gd) in the epithermal energy range were measured. A cadmium-covered vessel containing a pure water or an aqueous solution of a fission product element was inserted at the center of TCA (Tank-type Critical Assembly) core. Reactivity effects were obtained by the difference in the critical water levels between a pure water and an aqueous solution in the vessel. The measured reactivity was more than 1 φ and it was greater than the experimental uncertainties. Since the adjoint thermal flux below the cadmium-cutoff energy are largely depressed in the vessel, the reactivity effects in epithermal energy range could be measured. The analyses for the experiments were performed using the SRAC code system and neutron transport calculation code TWOTRAN. The exact Perturbation theory was applied to calculate the reactivity effects of fission product elements. The calculated reactivity effects using JENDL-3.2 and ENDF/B-IV cross-section libraries were compared against the measured values. The analyses using JENDL-3.2 gave reasonable results for these measurements.

Generation and validation of ORIGEN-S libraries for depletion and transmutation calculations based on JEF2.2 and EAF3 basic data

The data libraries for light elements, actinides and fission products of the ORIGEN-S code for depletion and transmutation calculations in the SCALE4.1 computer code system have been updated with respect to cross-section data, radioactive-decay data and fission-product yield data using JEF2.2 as the basic data source and EAF3 as an additional source. This required the fission-product library to be extended with 201 new fission-product nuclides or isomeric states. The effect of the update of different quantities involved is evaluated with a burn-up benchmark. When ORIGEN-S is used as a stand-alone code, i.e. without regular update of cross-sections of the major nuclides due to changes in the neutron spectrum during burn-up, the results show appreciable differences in actinide and fission-product densities due to the cross-section update. The effects of updates of decay data and fission-product yields are generally small, but with noticeable exceptions. The update of fission and capture reaction energies gives a small but systematic change in actinide and fission-product concentration. The new ORIGEN-S libraries have also been converted for use with the SCALE4.2 package.

Benchmark Model of Critical Experiment at TCA for Integral Evaluation of Fission Product Nuclide Cross Sections

Benchmark models of the critical experiments which measured reactivity effects of main fission product elements (Rh, Cs, Nd, Sm, Eu, and Gd) and of an important burnable poison element, erbium, were presented. These elements were dissolved in a vessel which was inserted into the central region of a fuel rod array of TCA (Tank-Type Critical Assembly). The atomic number densities of aqueous solutions in the vessel were calculated. Using these number densities, criticality calculations with MCNP 4A, TWOTRAN code, and JENDL-3.2 cross section library were performed for these experiments.

Measurements and Analyses of Reactivity Effect of Fission Product Nuclides in Epithermal Energy Range.

Measurements and Analyses of Reactivity Effect of Fission Product Nuclides in Epithermal Energy Range.