Mamoru Kamoshida

A new concept for the nuclear fuel recycle system: Application of the fluoride volatility reprocessing

Abstract Hybrid Recycle System (HRS) is proposed as an advanced recycle system. The HRS consists of improved fluoride volatility reprocessing and vibration packing MOX fuel fabrication processing. For the former, a part of U is volatilized as hexafluoride with diluted F 2 gas, and then residual U and Pu are volatilized with concentrated F 2 gas. Plutonium content of the mixed fluoride gas can be adjusted as desired by controlling the U fluorination reaction in the first step. The U is highly decontaminated and the mixture gas of UF 6 and PuF 6 is not purified. The fluoride mixture is reacted with H 2 O and H 2 and directly converted to the mixed oxide grain for the vibration packing. The HRS can reduce the costs of reprocessing and fuel fabrication, the amounts of radioactive wastes and the probability of Pu proliferation.

Valence Control and Solvent Extraction of Americium in the Presence of Ammonium Phosphotungstate

Fundamental investigations on valence control and solvent extraction of americium were carried out to develop a method for americium separation from reprocessing solution. In order to adjust americium valency from III to IV and VI, (NH4)10P2W 17O61 synthesized was used as complexant stabilizing Am(IV). Oxidation behavior of americium was investigated as a function of (NH4)10P2W 17O61 americium ratio. Using 0.1M (NH4)2S2O8 and 0.01M AgNO3 as oxidation reagent, Am(IV) was obtained quantitatively at the ratio of 15. On decreasing the ratio to 0.6, 92% of americium was adjusted to Am(VI). The concentration of (NH4)2S2O8 could be reduced to 1/15 compared to the previously reported method in which no complexant was used. Americium(IV) was also prepared by reacting O3 and AgNO3 but no Am(VI) was obtained even at low (NH4)10P2W 17O61 to americium ratio. Americium(VI) could be extracted by tri-n-butyl phosphate stably without influence of (NH4)10P2W 17O61. The distribution coefficient of Am(VI) was 4 between 100% ...

Development of Inherently Safe Technologies for Large Scale BWRs: (1) Plant System

The Fukushima Daiichi nuclear accident and their consequences have led to some rethinking about the safety technologies used in boiling water reactors (BWRs). We have been developing the following various safe technologies: a passive water-cooling system, an infinite-time air-cooling system, a hydrogen explosion prevention system, and an operation support system to better deal with reactor accidents. The above mentioned technologies are referred to as “inherently safe technologies”.The passive water-cooling system and infinite-time air-cooling system achieve core cooling without electricity. These systems are intended to cope with a long-term station black out (SBO), such as that which occurred at the Fukushima facility. Both these cooling systems remove relatively high decay heat for the initial 10 days after an accident, and then the infinite-time air-cooling system is used alone to remove attenuated decay heat after 10 days.The hydrogen explosion prevention system consists of a high-temperature resistant fuel cladding made of silicon-carbide (SiC cladding) and a passive autocatalytic recombiner (PAR). Since the SiC cladding generates less hydrogen gas than the current zircaloy fuel cladding when core damage occurs, the risk of hydrogen leakage from a primary containment vessel (PCV) to a reactor building (R/B), such as an operating floor, can be reduced because the pressure in the PCV can be kept lower with less hydrogen gas generation. The leaked hydrogen gas is recombined by the PAR.When a large-scale natural disaster occurs, fast event diagnosis and recognition of damaged equipment are necessary. Therefore, the operation support system consists of event identification and progress prediction functions to reduce the occurrence of false recognitions and human errors.This paper describes the following items: the targeted plant system; the heat exchange tests conducted for both water-cooling and air-cooling systems; the air-cooling enhancing technology for air-cooling in a 4700 MW thermal power class reactor; hydrogen generation tests for SiC material; and the concept of the operation support system.Copyright

Development of Inherently Safe Technologies for Large Scale BWRs: (4) Hydrogen Explosion Prevention System Using SiC Fuel Claddings

In the aftermath of the lessons learned from the Fukushima Daiichi nuclear accident, we have been developing the following various safe technologies for boiling water reactors (BWRs), including a passive water-cooling system, an infinite-time air-cooling system, a hydrogen explosion prevention system, and an operation support system for reactor accidents.One of inherently safe technologies currently under development is a system to prevent hydrogen explosion during severe accidents (SAs). This hydrogen explosion prevention system consists of a high-temperature resistant fuel cladding of silicon carbide (SiC), and a passive autocatalytic recombiner (PAR). Replacing the zircaloy (Zry) claddings currently used in LWRs with the SiC claddings decreases the hydrogen generation and thus decreases the risk of hydrogen leakage from a primary containment vessel (PCV) to a reactor building (R/B) such as an operation floor. The PAR recombines the leaked hydrogen gas so as to maintain the hydrogen concentration at less than the explosion limit of 4 % in the R/B.The advantages of using SiC claddings in the system were examined through experiments and SA analysis. Results of steam oxidation tests confirmed that SiC was estimated to show 2 to 3 orders of magnitude lower hydrogen generation rates during oxidation in a high temperature steam environment than Zry. Results of SA analysis showed that the total amount of hydrogen generation from fuels was reduced to one fifth or less. Calculation also showed that the lower heat of the oxidation reaction of SiC moderated the steep generation with the temperature increase. We expected this moderated steep generation to reduce the pressure increase in the PCV as well as prevent excess amounts of leaked hydrogen from hydrogen disposal rate capacity using PARs.The SiC cladding under consideration consists of an inner metallic layer, a SiC/SiC composite substrate, and an outer environment barrier coating (EBC). A thin inner metallic layer in combination with a SiC/SiC composite substrate functions as a barrier for fission products. EBC is introduced to have both corrosion resistance in high temperature water environments during normal operation and oxidation resistance in high temperature steam environments during SA. Further reduction of the hydrogen generation rate in high temperature steam by improving the EBC is expected to decrease the total amount of hydrogen generation even more.Copyright

Americium Valence Control Experiments in the Dissolved Solution of Irradiated Nuclear Fuel

A method based on valence control and solvent extraction is presented for separating minor actinides (Am, Np) from nuclear fuel reprocessing solution. With this method, Am is adjusted from trivalen...

Application of “Hydration Model” to Evaluate Gas Phase Transfer of Ruthenium and Technetium from Reprocessing Solutions

In order to evaluate the amounts of gas phase transferred ruthenium (Ru), and technetium (Tc), simulations were made for the continuous evaporator used in a reprocessing plant to concentrate high level liquid waste. The concentrations and activities of nitric acid and water, which controlled the reaction rate and gas-liquid equilibrium in the evaporator solution, were evaluated using the previously developed “Hydration Model”. When the feed solution contained 2.7 M (=mol/dm3) of nitric acid, the nitric acid concentration in the evaporator solution reached its maximum at the concentration factor (CF) of 6 (CF: concentration ratio of FPs in evaporator and feed solutions). The activities of nitric acid and water were saturated at values of 0.01 and 0.43, respectively, after the CF reached 6. The simulation predicted decontamination factors DFs of 2×105 and 8×103 for Ru and Tc, respectively, for a typical evaporation conditions with an operational pressure of 6,700 Pa, and FPs of 0.02 to 1.4 M. The simulation...

Extraction of Am(VI) from Nitric Acid Solution Containing Phosphate Anion by TBP

The extraction behavior of Am(VI) was studied for a nitric acid solution containing ammonium dihydrogenphosphate using tri-n-butylphosphate (TBP) and other extractants. TBP and HDEHP solutions diluted with n-dodecane do not reduce Am(VI) in the extraction procedure, whereas CMPO and TOA do. In TBP extraction, phosphate anion raises the separation factor of Am from Nd as high as 120 with 1 MTBP solvent from 3 M HNO3-1 M NH4H2PO4, although it suppresses the extraction of both. A distribution ratio value of 6.6 was obtained using undiluted TBP. Dissociated phosphate anion H2PO4- forms a complex that lowers the distribution ratio of Am even in highly concentrated nitrate solution at low acidity. Americium(VI) is extracted as AmO2(NO3)2.2 TBP. The stability constant, log β, for Am(VI)- H2PO4- complex was estimated as 1.2.