Enriched Uranyl Nitrate Solution Research Articles

Analysis of fission product gas pressure and radioactivity in SAMOP reactor experimental facility

The fission-product gas pressure and radioactivity analysis of a subcritical assembly for 99Mo production (SAMOP) experimental facility have been done. SAMOP reactor is fueled with low enriched uranyl nitrate solution UO2(NO3)2 of 300 g U/L, the reactor core is in the form an annular tube surrounded by a ring of fuel tubes. The SAMOP system is designed to be operated at 100 to 120 hours periodic operation per batch. The analysis method is done by using ORIGEN2 computer code, at the condition of the maximum fuel temperature of 54 °C, 120 hours reactor operating time, and neutron flux varied from 1010 to 1012 n/cm2s. The result show that the inventory of the gaseous fission products is dominantly consisted of the xenon, radium, helium, and tritium isotopes with the total activity of 1.3775x1013 Bq (372.29 Ci) and total gas volume of 0.0986 cm3 or giving a total gas pressure of 5.78 kPa. The isotopes of xenon and krypton are the major contributors to the total gas pressure. The analysis result shows that the gas pressure of the fission products is very small, therefore the pressure monitoring in the SAMOP reactor core is not necessary.

Source Term Analysis of SAMOP Reactor Experimental Facility

The Subcritical Assembly for Molybdenum-99 Production (SAMOP) system by using low enriched uranyl nitrate solution as reactor fuel is being developed and designed in the Centre for Accelerator Science & Technology, Yogyakarta. The source term analysis of the SAMOP experimental facility is done in conjunction with the radiation shielding design. The method used is by using MCNPX software for fuel burn-up calculations. The burnup calculation is done both in the reactor operation without 99Mo extraction and with 100% 99Mo extraction per batch (6 day + 1 day reactor off) along 1 month operation (5 batch). The calculation was done for two conditions i.e. without and with 99Mo extraction. The calculation result shows that the radioactivity inside the SAMOP reactor core reach 523 Ci for the condition without 99Mo extraction. The effective dose at a distance of 50 cm from the outer surface of the SAMOP reactor coolant tank is 1600 mSv/h. When the SAMOP reactor is operated then followed by 99Mo extraction, the overall radioactivity of the reactor core is slightly lower than the above value. The calculation results also shows that the highest dose from volatile nuclide is caused by 132Te isotope i.e. 92.12 mSv/h.

Analisis Keselamatan Kritikalitas Teras Silindris Stacy dengan Perhitungan Transport Monte Carlo

A set of experiment has been done at STACY facility and many fundamental parameters of uranyl nitrate solution have been found out. Criticality is one of main parameters in predicting neutronic characteristic of STACY experiment beside solution level reactivity, void reactivity, kinetic parameter and temperature reactivity which dominates transient phenomenon in abnormal condition. Criticality experiment performed at STACY core uses 9.97% 235U -enriched uranyl nitrate solution with 80-cm-diameter cylindrical and 150-cm-height tank. Eight critical configurations in unrelected and water-reflected conditions were selected in this paper for criticality safety calculation with Monte Carlo transport code MCNPX. For all configurations, MCNPX calculations show good consistency with the trend of producing underestimated k eff. Calculation biases with experimental data ( k eff = 1) for water-reflected configurations, i.e. 0.01-0.18%, were slightly better than those of unreflected configurations (0.14-0.41%). MCNPX calculation results which are better than the prediction of MCNP-4C concludes that MCNPX is more eligible to be applied to criticality safety analysis of uranyl nitrate solution in commercial nuclear fuel cycle facility.

Optimization study and neutronic design calculations of LEU fuelled homogeneous aqueous solution nuclear reactors for the production of short lived fission product isotopes

Low enriched uranium (LEU) based solution reactors possess the potential to meet increasing demand of 99Mo and other short lived fission product radioisotopes being used in medical field. In the current study, optimization and neutronic design calculations have been carried out for an LEU homogeneous aqueous solution reactor with uranyl nitrate as a solution fuel. Lattice calculations and core modeling was performed employing available standard nuclear reactor codes WIMSD and CITATION. Calculation procedure was verified using experimental published data and simulated criticality results were compared with the published data. After verification of the calculation methodology, optimization and design calculations were performed for four different uranium enrichments (5%, 10%, 15%, and 19.99%) of the uranyl nitrate solution. Keeping in view the restraints on peak power density of solution reactors, annular geometry is suited as compared to the cylindrical geometry. Among the four considered enrichments in the analysis, 19.99% enriched uranyl nitrate solution with annular geometry presents optimum design parameters for the homogeneous aqueous solution reactor.

Reactivity Effect Measurement of Neutron Interaction between Two Slab Cores Containing 10% Enriched Uranyl Nitrate Solution without Neutron Isolator

The reactivity effect of neutron interaction between two identical units containing low enriched (10% 235U enrichment) uranyl nitrate solution without neutron isolator was measured in the STACY. The unit has 350 mm of thickness and 690 mm of width and distance between those two units was adjustable from 0 to 1,450 mm. Condition of the solution was about 290gU/l in uranium concentration, about 0.8 N in free nitric acidity, 24–27°C in temperature and about 1.4g/cm3 in solution density. The reactivity effect was estimated from variation of critical solution level from 495 to 763 mm depending on the core distance. The reactivity effect was also evaluated by the solid angle method and a computational method using the continuous energy Monte Carlo code MCNP-4C and the nuclear data library JENDL 3.2. Those estimations were compared and the distance to isolate neutron interaction was overestimated by the solid angle method. The computational method considering neutron reflection from surrounding structures reproduced well the isolation distance given by the experiment.

Analysis of dose distribution for heavily exposed workers in the first criticality accident in Japan.

The first criticality accident in Japan occurred in a uranium processing plant in Tokai-mura on September 30, 1999. The accident, which occurred while a large amount of enriched uranyl nitrate solution was being loaded into a tank, led to a chain reaction that continued for 20 h. Two workers who were pouring the uranium solution into the tank at the time were heterogeneously exposed to neutrons and gamma rays produced by nuclear fission. Analysis of dose distributions was essential for the understanding of the clinical course observed in the skin and organs of these workers. We developed a numerical simulation system, which consists of mathematical human models and Monte Carlo radiation transport programs, for analyzing dose distributions in various postures and applied the system to the dose analysis for the two workers. This analysis revealed the extreme heterogeneity of the doses from neutrons and gamma rays in the skin and body, which depended on the positions and postures of the workers. The detailed dose analysis presented here using color maps is indispensable for an understanding of the biological effects of high-dose exposure to a mixed field of neutrons and gamma rays as well as for the development of emergency treatments for victims of radiation exposure.

Critical Experiments on 10% Enriched Uranyl Nitrate Solution Using an 80-cm-Diameter Cylindrical Core

The third series of the critical experiments on 10% enriched uranyl nitrate solution has been performed at the Static Experiment Critical Facility (STACY) of the Japan Atomic Energy Research Institute. Water-reflected and unreflected 80-cm-diam cylindrical cores were used to obtain the systematic data of critical solution height and differential reactivity for various uranium concentrations from 190 to 240 g/l. The numerically evaluated extrapolation length of neutron flux distribution was in good agreement with the experimental result. The effective neutron multiplication factor keff for each core configuration and the effect of uncertainties on keff were also numerically evaluated with both the detailed experimental configuration of critical cores and a benchmark model provided for the validation of nuclear calculation codes. The MCNP 4B was used for the evaluation calculations with JENDL-3.2 cross-section library, and the value of the keff of the benchmark model was reproduced within the difference of 0.05% Δkeff for the water-reflector cores and 0.17% Δkeff for the unreflected cores.

Reactivity Effect Measurement of Neutron Interaction between Two Slab Cores Containing 10% Enriched Uranyl Nitrate Solution without Neutron Isolator

The reactivity effect of neutron interaction between two identical units containing low enriched (10% 235U enrichment) uranyl nitrate solution without neutron isolator was measured in the STACY. The unit has 350 mm of thickness and 690 mm of width and distance between those two units was adjustable from 0 to 1,450 mm. Condition of the solution was about 290gU/l in uranium concentration, about 0.8 N in free nitric acidity, 24–27°C in temperature and about 1.4g/cm3 in solution density. The reactivity effect was estimated from variation of critical solution level from 495 to 763 mm depending on the core distance. The reactivity effect was also evaluated by the solid angle method and a computational method using the continuous energy Monte Carlo code MCNP-4C and the nuclear data library JENDL 3.2. Those estimations were compared and the distance to isolate neutron interaction was overestimated by the solid angle method. The computational method considering neutron reflection from surrounding structures reproduced well the isolation distance given by the experiment.

Benchmark Evaluation on Single Core System Composed of 10% Enriched Uranyl Nitrate Solution at STACY

Critical experiments on 10% enriched uranyl nitrate solution have been conducted for obtaining benchmark data using a Static Experiment Critical Facility, STACY, in the Nuclear Fuel Cycle Safety Engineering Research Facility, NUCEF. Two cylindrical and one slab tank were used to construct a single core system. Critical solution levels were measured for three core tanks with changing the uranium concentration of uranyl nitrate solution. Fundamental data were obtained for unreflected or water reflected cores. The sensitivities of the fuel composition and tank geometry on the criticality were calculated, and total uncertainty of the experiments are evaluated less than 0.15% Δ k. Calculations of the neutron multiplication factor keff for the critical conditions were performed using a continuous-energy Monte Carlo code MCNP-4B employing JENDL-3.2. This paper presents the core configurations of STACY, their experimental uncertainties and calculation results with above calculation system.

Criticality Safety Benchmark Experiment on 10% Enriched Uranyl Nitrate Solution Using a 28-cm-Thickness Slab Core

The second series of critical experiments with 10% enriched uranyl nitrate solution using a 28-cm-thick slab core have been performed with the Static Experiment Critical Facility of the Japan Atomic Energy Research Institute. Systematic critical data were obtained by changing the uranium concentration of the fuel solution from 464 to 300 gU/l under various reflector conditions. In this paper, the thirteen critical configurations for water-reflected cores and un reflected cores are identified and evaluated. The effects of uncertainties in the experimental data on k eff are quantified by sensitivity studies. Benchmark model specifications that are necessary to construct a calculational model are given. The uncertainties of k eff's included in the benchmark model specifications are approximately 0.1%δk eff. The thirteen critical configurations are judged to be acceptable benchmark data. Using the benchmark model specifications, sample calculation results are provided with several sets of standard codes and cross section data.

Criticality Safety Benchmark Experiment on 10% Enriched Uranyl Nitrate Solution Using a 28-cm-Thickness Slab Core.

The second series of critical experiments with 10% enriched uranyl nitrate solution using a 28-cm-thick slab core have been performed with the Static Experiment Critical Facility of the Japan Atomic Energy Research Institute. Systematic critical data were obtained by changing the uranium concentration of the fuel solution from 464 to 300 gU/l under various reflector conditions. In this paper, the thirteen critical configurations for water-reflected cores and un reflected cores are identified and evaluated. The effects of uncertainties in the experimental data on k eff are quantified by sensitivity studies. Benchmark model specifications that are necessary to construct a calculational model are given. The uncertainties of k eff's included in the benchmark model specifications are approximately 0.1%δk eff. The thirteen critical configurations are judged to be acceptable benchmark data. Using the benchmark model specifications, sample calculation results are provided with several sets of standard codes and cross section data.

Measurement of Neutron and Gamma-Ray Absorbed Doses Under Criticality Accident Conditions at TRACY Using Tissue-Equivalent Dosimeters

To evaluate neutron and gamma-ray absorbed doses in human bodies at criticality accidents, two kinds of tissue-equivalent dosimeters, a polymer-alanine dosimeter and a thermoluminescent dosimeter (TLD) made of 7Li211B4O7, were applied to dosimetry experiments with ˜10% enriched uranyl nitrate solution at the Transient Experiment Critical Facility (TRACY) in the Japan Atomic Energy Research Institute. For the experiments, five transient operations were conducted to simulate criticality accidents by varying the conditions of reactivity addition. Very high doses from both neutrons and gamma rays were successfully measured in the range of 1.5 to 1600 Gy by using polymer-alanine dosimeters. The gamma-ray doses were able to be determined in the range of 1 to 900 Gy by using 7Li211B4O7 TLDs. In addition, it is confirmed that the doses are proportional to integrated power during transient operations although the conditions of reactivity addition are different. Since the sensitivity of 7Li211B4O7 to gamma rays is almost the same as that of alanine, the neutron doses are easily evaluated without any complicated correction by subtracting the gamma-ray doses obtained by the 7Li211B4O7 TLDs from the sum of neutron and gamma-ray doses by the polymer-alanine dosimeters. As a result of computational analyses by the MCNP4B code, it is also found that calculated doses agree with measured ones within 95% confidence intervals and the MCNP4B is applicable to the evaluation of neutron and gamma-ray absorbed doses during the transient.

Summary of the JCO criticality accident in Tokai-mura and a dose assessment.

A criticality accident occurred on September 30, 1999, in a conversion test facility at the JCO Tokai site. The accident was triggered by pouring an 18.8% enriched uranyl nitrate solution into a precipitation vessel beyond the critical mass. The accident continued for about 19 hours before the criticality could be stopped. during which time neutrons and gamma-rays were emitted continuously due to fission reactions. The total number of fission reactions was 2.5 x 10(18), which was estimated by an activity analysis of the fission products in the solution of the precipitation vessel. The accident gave serious radiation dose to 3 employees and fatal dose to 2 of them. Neutrons and gamma-rays emitted by the accident caused meaningful doses to the residents of the surrounding area of JCO. The dominant dose to the residents and JCO employees was brought by neutrons and gamma-rays from the precipitation vessel, while the contribution of radioactive plume was negligible. The individual dose was estimated for 234 resident, 169 JCO employees and 260 emergency personnel. The maximum doses were 21 mSv for the residents, 48mSv for the JCO employees, and 9.4mSv for the emergency personnel, respectively. No deterministic effect, however, has been observed, except for the 3 workers.

Descriptions of High and Low Enriched U 235 Critical Experiments in 174-1 Spheres

A wide variety of criticality experiments, using different geometries and solutions, was performed at Oak Ridge National Laboratory in the 1950s. Two sets performed in 174-1 spheres are analyzed to provide benchmark descriptions for validation of computational tools used by nuclear criticality specialists. One set was performed in an unreflected sphere and contained a highly enriched uranyl nitrate solution (93.18% U 235). Three of the four measurements from this set were poisoned with boron. The second set consists of one unreflected and two water-reflected measurements and used low-enriched uranyl fluoride solutions (4.9% U 235). One of these experiments was critical with the sphere 83% full. An uncertainty in the experimental k eff was found by sensitivity studies on reported measurement uncertainties, inconsistencies, and omissions in experimental parameters. Benchmark configurations were found for all experiments. To be functional for all computer codes, one-dimensional configurations were used when possible. The sensitivity study of the experiments was performed using ONEDANT with 27-Group ENDF/B-IV cross sections and MCNP with continuous energy ENDF/B-V cross section data. K eff values for the benchmark configurations were found using MCNP with ENDF/B-V and ENDF/B-VI cross section data, ONEDANT and KENO-Va with Hansen-Roach and 27-group ENDF/B-IV cross sections. © 1997 Elsevier Science Ltd.

Critical Experiments on 10% Enriched Uranyl Nitrate Solution Using A 60-cm-Diameter Cylindrical Core

A series of critical experiments with 10% enriched uranyl nitrate solution using a cylindrical core tank 60 cm in diameter have been performed with the Static Experiment Critical Facility at the Nuclear Fuel Cycle Safety Engineering Research Facility in the Tokai research establishment of the Japan Atomic Energy Research Institute. In the first series of experiments using the cylindrical core tank, systematic data of the critical height for water-reflected cores and unreflected cores were obtained by changing the uranium concentration of the fuel solution from 313 to 225 g U/ℓ. As the reactivity of each core is controlled only by solution height, these criticality configurations, which have simple cylindrical shapes, are available for the validation of calculation codes used in criticality safety designs of nuclear fuel cycle facilities.The neutron multiplication factors of experimental cores were calculated with the two-dimensional transport code TWOTRAN in the SRAC code system and with the continuous-energy Monte Carlo code MCNP4A, employing the Japanese evaluated nuclear data library JENDL-3.2. The calculations from the combination of these calculation codes and the nuclear data library reproduce the neutron multiplication factors within an error of 0.9% for the experimental configuration of critical cores.