• Title/Summary/Keyword: Simulated fission products

Search Result 22, Processing Time 0.165 seconds

Fabrication of Nitride Fuel Pellets by Using Simulated Spent Nuclear Fuel (모의 사용후 핵연료를 이용한 질화물 핵연료 소결체 제조)

  • Ryu, Ho-Jin;Lee, Jae-Won;Lee, Young-Woo;Lee, Jung-Won;Park, Geun-Il
    • Journal of Powder Materials
    • /
    • v.15 no.2
    • /
    • pp.87-94
    • /
    • 2008
  • In order to investigate a nitriding process of spent oxide fuel and the subsequent change in thermal properties after nitriding, simulated spent fuel powder was converted into a nitride pellet with simulated fission product elements through a carbothermic reduction process. Nitriding rate of simulated spent fuel was decreased with increasing of the amount of fission products. Contents of Ba and Sr in simulated spent fuel were decreased after the carbothermic reduction process. The thermal conductivity of the nitride pellet was decreased by an addition of fission product element but was higher than that of the oxide fuel containing fission product elements.

Phase analysis of simulated nuclear fuel debris synthesized using UO2, Zr, and stainless steel and leaching behavior of the fission products and matrix elements

  • Ryutaro Tonna;Takayuki Sasaki;Yuji Kodama;Taishi Kobayashi;Daisuke Akiyama;Akira Kirishima;Nobuaki Sato;Yuta Kumagai;Ryoji Kusaka;Masayuki Watanabe
    • Nuclear Engineering and Technology
    • /
    • v.55 no.4
    • /
    • pp.1300-1309
    • /
    • 2023
  • Simulated debris was synthesized using UO2, Zr, and stainless steel and a heat treatment method under inert or oxidizing conditions. The primary U solid phase of the debris synthesized at 1473 K under inert conditions was UO2, whereas a (U, Zr)O2 solid solution formed at 1873 K. Under oxidizing conditions, a mixture of U3O8 and (Fe, Cr)UO4 phases formed at 1473 K, whereas a (U, Zr)O2+x solid solution formed at 1873 K. The leaching behavior of the fission products from the simulated debris was evaluated using two methods: the irradiation method, for which fission products were produced via neutron irradiation, and the doping method, for which trace amounts of non-radioactive elements were doped into the debris. The dissolution behavior of U depended on the properties of the debris and aqueous solution for immersion. Cs, Sr, and Ba leached out regardless of the primary solid phases. The leaching of high-valence Eu and Ru ions was suppressed, possibly owing to their solid-solution reaction with or incorporation into the uranium compounds of the simulated debris.

Determination of Plutonium Present in Highly Radioactive Irradiated Fuel Solution by Spectrophotometric Method

  • Dhamodharan, Krishnan;Pius, Anitha
    • Nuclear Engineering and Technology
    • /
    • v.48 no.3
    • /
    • pp.727-732
    • /
    • 2016
  • A simple and rapid spectrophotometric method has been developed to enable the determination of plutonium concentration in an irradiated fuel solution in the presence of all fission products. An excess of ceric ammonium nitrate solution was employed to oxidize all the valence states of plutonium to +6 oxidation state. Interference due to the presence of fission products such as ruthenium and zirconium, and corrosion products such as iron in the envisaged concentration range, as in the irradiated fuel solution, was studied in the determination of plutonium concentration by the direct spectrophotometric method. The stability of plutonium in +6 oxidation state was monitored under experimental conditions as a function of time. Results obtained are reproducible, and this method is applicable to radioactive samples resulting before the solvent extraction process during the reprocessing of fast reactor spent fuel. An analysis of the concentration of plutonium shows a relative standard deviation of <1.2% in standard as well as in simulated conditions. This reflects the fast reactor fuel composition with respect to uranium, plutonium, fission products such as ruthenium and zirconium, and corrosion products such as iron.

ESTIMATION OF THE FISSION PRODUCTS, ACTINIDES AND TRITIUM OF HTR-10

  • Jeong, Hye-Dong;Chang, Soon-Heung
    • Nuclear Engineering and Technology
    • /
    • v.41 no.5
    • /
    • pp.729-738
    • /
    • 2009
  • Given the evolution of High-Temperature Gas-cooled Reactor(HTGR) designs, the source terms for licensing must be developed. There are three potential source terms: fission products, actinides in the fuel and tritium in the coolant. It is necessary to provide first an inventory of the source terms under normal operations. An analysis of source terms has yet to be performed for HTGRs. The previous code, which can estimate the inventory of the source terms for LWRs, cannot be used for HTGRs because the general data of a typical neutron cross-section and flux has not been developed. Thus, this paper uses a combination of the MCNP, ORIGEN, and MONTETEBURNS codes for an estimation of the source terms. A method in which the HTR-10 core is constructed using the unit lattice of a body-centered cubic is developed for core modeling. Based on this modeling method by MCNP, the generation of fission products, actinides and tritium with an increase in the burnup ratio is simulated. The model developed by MCNP appears feasible through a comparison with models developed in previous studies. Continuous fuel management is divided into five periods for the feeding and discharging of fuel pebbles. This discrete fuel management scheme is employed using the MONTEBURNS code. Finally, the work is investigated for 22 isotope fission products of nuclides, 22 actinides in the core, and tritium in the coolant. The activities are mainly distributed within the range of $10^{15}{\sim}10^{17}$ Bq in the equilibrium core of HTR-10. The results appear to be highly probable, and they would be informative when the spent fuel of HTGRs is taken into account. The tritium inventory in the primary coolant is also taken into account without a helium purification system. This article can lay a foundation for future work on analyses of source terms as a platform for safety assessment in HTGRs.

Reprocessing of simulated voloxidized uranium-oxide SNF in the CARBEX process

  • Boyarintsev, Alexander V.;Stepanov, Sergei I.;Kostikova, Galina V.;Zhilov, Valeriy I.;Chekmarev, Alexander M.;Tsivadze, Aslan Yu.
    • Nuclear Engineering and Technology
    • /
    • v.51 no.7
    • /
    • pp.1799-1804
    • /
    • 2019
  • The concept of a new method, the CARBEX (CARBonate EXtraction) process, was proposed for reprocessing of spent uranium oxide fuel. The proposed process is based on use of water solutions of $Na_2CO_3$ or $(NH_4)_2CO_3$ and solvent extraction (SE) by the quaternary ammonium compounds for selective recovery and purification of U from the fission products (FPs). Applying of SE allows to reach high degree of purification of U from FPs. Carrying out the processes in poorly aggressive alkaline carbonate media leads to increasing safety of SNF's reprocessing and better selectivity of separation of lanthanides and actinides. Moreover carbonate reprocessing media allows to carry out a recycling and regeneration of reagents. We have been done laboratory scale experiments on the extraction components of simulated voloxidated spent fuel in the solutions of NaOH or $Na_2CO_3-H_2O_2$ and recovery of U from carbonate solutions by SE method using carbonate of methyltrioctylammonium in toluene. It was shown that the purification factors of U from impurities of simulated FPs reached values $10^3-10^5$. The received results support our opinion that CARBEX after the further development can become more safe, simple and profitable method of spent fuel reprocessing.

Effect of Spray System on Fission Product Distribution in Containment During a Severe Accident in a Two-Loop Pressurized Water Reactor

  • Dehjourian, Mehdi;Rahgoshay, Mohammad;Sayareh, Reza;Jahanfarnia, Gholamreza;Shirani, Amir Saied
    • Nuclear Engineering and Technology
    • /
    • v.48 no.4
    • /
    • pp.975-981
    • /
    • 2016
  • The containment response during the first 24 hours of a low-pressure severe accident scenario in a nuclear power plant with a two-loop Westinghouse-type pressurized water reactor was simulated with the CONTAIN 2.0 computer code. The accident considered in this study is a large-break loss-of-coolant accident, which is not successfully mitigated by the action of safety systems. The analysis includes pressure and temperature responses, as well as investigation into the influence of spray on the retention of fission products and the prevention of hydrogen combustion in the containment.

FEA Study on Hoop Stress of Multilayered SiC Composite Tube for Nuclear Fuel Cladding (핵연료 피복관용 다중층 SiC 복합체 튜브의 Hoop Stress 전산모사 연구)

  • Lee, Hyeon-Geun;Kim, Daejong;Park, Ji Yeon;Kim, Weon-Ju
    • Journal of the Korean Ceramic Society
    • /
    • v.51 no.5
    • /
    • pp.435-441
    • /
    • 2014
  • Silicon carbide-based ceramics and their composites have been studied for application to fusion and advanced fission energy systems. For fission reactors, $SiC_f$/SiC composites can be applied to core structural materials. Multilayered SiC composite fuel cladding, owing to its superior high temperature strength and low hydrogen generation under severe accident conditions, is a candidate for the replacement of zirconium alloy cladding. The SiC composite cladding has to retain its mechanical properties and original structure under the inner pressure caused by fission products; as such it can be applied as a cladding in fission reactor. A hoop strength test using an expandable polyurethane plug was designed in order to evaluate the mechanical properties of the fuel cladding. In this paper, a hoop strength test of the multilayered SiC composite tube for nuclear fuel cladding was simulated using FEA. The stress caused by the plug was distributed nonuniformly because of the friction coefficient difference between the inner surface of the tube and the plug. Hoop stress and shear stress at the tube was evaluated and the relationship between the concentrated stress at the inner layer of the tube and the fracture behavior of the tube was investigated.

The measurement of oxygen and metal ratio of simulated spent fuels by wet and dry chemical analysis (습식 및 건식법에 의한 모의 사용후핵연료의 O/M비 측정)

  • Choi, Ke Chon;Lee, Chang Heon;Kim, Won Ho
    • Analytical Science and Technology
    • /
    • v.16 no.2
    • /
    • pp.117-124
    • /
    • 2003
  • Oxygen to metal ratio has been measured by wet and dry chemical analysis to study the properties of sintered $UO_2$ pellets and $U_3O_8$ in the lithium reduction process of spent pressurized water reactor fuels. Uranium dioxide pellets simulated for the spent PWR fuels with burnup values of 20,000~60,000 MWd/MtU were prepared by mixing $UO_2$ powder and oxides of fission product elements, pelleting the powder mixture and sintering it at $1,700^{\circ}C$ under a hydrogen atmosphere. For wet chemical analysis, the simulated spent fuels were dissolved with mixed acid (10 M HCl : 8 M $HNO_3$, 2.5 : 1, v/v) using acid digestion bomb technique. The total amount of uranium and fission products added in the simulated spent fuels were measured using inductively coupled plasma atomic emission spectrometry. Weight change of the simulated fuel during its oxydation was measured by thermogravimetry and then the O/M ratio result was compared to that obtained by wet chemical analysis. Influence of $Mo_{0.4}-Ru_{0.4}-Rh_{0.1}-Pd_{0.1}$, quaternary alloy, on the determination of O/M ratio was investigated.

The Characteristics of an Oxidative Dissolution of Simulated Fission Product Oxides in $(NH_4)_2CO_3$ Solution Containing $H_2O_2$ ($H_2O_2$ 함유 $(NH_4)_2CO_3$ 용액에서 모의 FP-산화물의 산화용해 특성)

  • Lee, Eil-Hee;Lim, Jae-Gwan;Chung, Dong-Yong;Yang, Han-Beum;Kim, Kwang-Wook
    • Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
    • /
    • v.7 no.2
    • /
    • pp.93-100
    • /
    • 2009
  • This study has been carried out to look into the characteristics of an oxidative-dissolution of fission products (FP) co-dissolved with uranium (U) in a $(NH_4)_2CO_3$ carbonate solution. Simulated FP-oxides which contained 12 components have been added to the solution to examine their dissolution characteristics. It is found that $H_2O_2$ is an effective oxidant to minimize the oxidative-dissolution of FP. In the 0.5 M $(NH_4)_2CO_3$-0.5 M $H_2O_2$ solution, some elements such as Re, Te, Cs and Mo seem to be dissolved together with U, while 98${\pm}$2% for Re and Te, 94${\pm}$2% for Cs, and 29${\pm}$2 % for Mo are dissolved for 2 hours. It is revealed that dissolution rates of Re, Te and Cs are high (completely dissolved within 10${\sim}$20 minutes) due to their high solubility in the $(NH_4)_2CO_3$ solution regardless of the addition of $H_2O_2$, and independent of the concentrations of $Na_2CO_3$ and $H_2O_2$. However, the dissolution ratio of Mo seems to be slightly increased with time and about 33 % for 4 hours, indicating a very slow dissolution rate and also independent of the $(NH_4)_2CO_3$ concentration. It is found that the most important factor for the oxidative-dissolution of FP is the pH of the solution and an effective dissolution is achieved at a pH between 9${\sim}$10 in order to minimize the dissolution of FP.

  • PDF