Pyroprocess Experiments at ENEA Laboratories

Giorgio DE ANGELIS,Elio BAICCHI,Mauro CAPONE,Carlo FEDELI,Massimo SEPIELLI,Giuliano TIRANTI,Mirko DA ROS,Francesca GIACOBBO,Marco GIOLA,Elena MACERATA,Mario MARIANI 한국방사성폐기물학회 2015 방사성폐기물학회지 Vol.13 No.S

A new facility, known as Pyrel III, has been installed at ENEA laboratories for pyrochemical process studies under inactive conditions. It is a pilot plant which allows electrorefining and electroreduction experiments to be conducted on simulated fuel. The main component of the plant is a zirconia crucible. The crucible is heated by a furnace which is supported in an externally water-cooled well under the floor of a steel glove-box, where an argon atmosphere is maintained by a continual purge of about 10 L·min-1. The vessel is loaded with LiCl-KCl eutectic salt (59-41 mol%) and is currently operated at 460 °C. Several improvements on Pyrel II (the previous operating plant) have been introduced into Pyrel III. They are described in detail, together with the results from the first experimental campaign which used lanthanum metal. Moreover, studies about the treatment of chloride salt wastes from pyroprocesses have been conducted in parallel. They follow two main routes: on one hand, a matrix termed sodalite, a naturally occurring mineral containing chlorine, has been synthesized from a mix of nepheline, simulated exhausted salts and glass frit; on the other hand, a novel method proposed by Korea Atomic Energy Research Institute (KAERI) is under assessment. The final waste forms have been fully characterized with the support of the Politechnique of Milan, by means of density measurements, thermal analysis, and stereomicroscopy observations, FTIR, XRD, and RAMAN spectra, as well as leach tests under static conditions.

Preliminary design of a production automation framework for a pyroprocessing facility

신문수,류동석,한종희,김기호,손영준 한국원자력학회 2018 Nuclear Engineering and Technology Vol.50 No.3

Pyroprocessing technology has been regarded as a promising solution for recycling spent fuel in nuclearpower plants. The Korea Atomic Energy Research Institute has been studying the current status ofequipment and facilities for pyroprocessing and found that existing facilities are manually operated;therefore, their applications have been limited to laboratory scale because of low productivity and safetyconcerns. To extend the pyroprocessing technology to a commercial scale, the facility, including all theprocessing equipment and the material-handling devices, should be enhanced in view of automation. Inan automated pyroprocessing facility, a supervised control system is needed to handle and managematerial flow and associated operations. This article provides a preliminary design of the supervisingsystem for pyroprocessing. In particular, a manufacturing execution system intended for an automatedpyroprocessing facility, named Pyroprocessing Execution System, is proposed, by which the overallproduction process is automated via systematic collaboration with a planning system and a controlsystem. Moreover, a simulation-based prototype system is presented to illustrate the operability of theproposed Pyroprocessing Execution System, and a simulation study to demonstrate the interoperabilityof the material-handling equipment with processing equipment is also provided.

A CONCEPTUAL STUDY OF PYROPROCESSING FOR RECOVERING ACTINIDES FROM SPENT OXIDE FUELS

Yoo, Jae-Hyung,Seo, Chung-Seok,Kim, Eung-Ho,Lee, Han-Soo Korean Nuclear Society 2008 Nuclear Engineering and Technology Vol.40 No.7

In this study, a conceptual pyroprocess flowsheet has been devised by combining several dry-type unit processes; its applicability as an alternative fuel cycle technology was analyzed. A key point in the evaluation of its applicability to the fuel cycle was the recovery yield of fissile materials from spent fuels as well as the proliferation resistance of the process. The recovery yields of uranium and transuranic elements (TRU) were obtained from a material balance for every unit process composing the whole pyroprocess. The material balances for several elemental groups of interest such as uranium, TRU, rare earth, gaseous fission products, and heat generating elements were calculated on the basis of the knowledge base that is available from domestic and foreign experimental results or technical information presented in open literature. The calculated result of the material balance revealed that uranium and TRU could be recovered at 98.0% and 97.0%, respectively, from a typical PWR spent fuel. Furthermore, the anticipated TRU product was found to emit a non-negligible level of $\gamma$-ray and a significantly higher level of neutrons compared to that of a typical plutonium product obtained from the PUREX process. The results indicate that the product from this conceptual pyroprocessing should be handled in a shielded cell and that this will contribute favorably to retaining proliferation resistance.

Enhancing Operational Stability in Pyroprocessing: Study on UO2 Powder Removal in Electrorefining Process

Dalsung Yoon,Sang-Kwon Lee,Juho Lee,Seungwoo Paek,Taeho Kim,Chang Hwa Lee 한국방사성폐기물학회 2023 한국방사성폐기물학회 학술논문요약집 Vol.21 No.2

Pyroprocessing technology has emerged as a viable alternative for the treatment of metal/oxide used fuel within the nuclear fuel cycle. This innovative approach involves an oxide reduction process wherein spent fuel in oxide form is placed within a cathode basket immersed in a molten LiCl-Li2O salt operating at 923 K. The chemical reduction of these oxide materials into their metallic counterparts occurs through a reaction with Li metal, which is electrochemically deposited onto the cathode. However, during process, the generation of Li2O within the fuel basket is inevitable, and due to the limited reduction efficiency, a significant portion of rare earth oxides (REOx) remains in their oxide state. The presence of these impurities, specifically Li2O and REOx, necessitates their transfer into the electrorefining system, leading to several challenges. Both Li2O and REOx exhibit reactivity with UCl3, the primary electrolyte within the electrorefining system, causing a continuous reduction in UCl3 concentration throughout the process. Furthermore, the formation of fine UO2 powder within the salt system, resulting from chemical reactions, poses a potential long-term operational and safety concern within the electrorefining process.Various techniques have been developed to address the issue of UO2 fine particle removal from the salt, utilizing both chemical and mechanical methods. However, it is crucial that these methods do not interfere with the core pyroprocessing procedure. This study aims to investigate the impact of Li2O and REOx introduced from the electrolytic reduction process on the electrorefining system. Additionally, we propose a method to effectively eliminate the generated UO2 fine powder, thereby enhancing the long-term operational stability of the electrorefining process. The efficiency of this proposed solution in removing oxidized powder has been confirmed through laboratory-scale testing, and we will provide a comprehensive discussion of the detailed results.

Use of a single fuel containment material during pyroprocessing tests

Choi, Eun-Young,Won, Chan Yeon,Lee, Sung-Jai,Kang, Dae-Seung,Kim, Sung-Wook,Cha, Ju-Sun,Park, Wooshin,Im, Hun Suk,Hur, Jin-Mok Elsevier 2015 Annals of nuclear energy Vol.76 No.-

<P><B>Abstract</B></P> <P>The use of a single stainless steel (STS) wire mesh basket as the fuel containment material for a series of pyroprocessing steps has been studied. The use of a single basket minimizes fuel loss and was enabled by transporting and using the basket containing the fuel from one test to the next without unloading it. The series of tests consisted of electrolytic reduction, LiCl distillation, electrorefining, LiCl–KCl distillation, and finally a second electrolytic reduction and a subsequent LiCl distillation step. While the electrolytic reduction of UO<SUB>2</SUB> was conducted in a LiCl–Li<SUB>2</SUB>O molten salt electrolyte at 650°C using the STS wire mesh basket as the cathode, the electrorefining was carried out in a LiCl–KCl–UCl<SUB>3</SUB> molten salt electrolyte at 500°C, using the STS wire mesh basket as the anode. During the salt (LiCl and LiCl–KCl) distillation processes, the product of electrolytic reduction/electrorefining in the basket, which included metallic U and residual salts, was distilled at 850°C under vacuum. The electrolytic reduction, electrorefining, and salt distillation processes were successfully demonstrated with the use of a single STS wire mesh basket through the entire cycle. However, an unstable intermetallic U–Fe layer was observed between the reduction product and the STS basket, when a cross section of the basket was studied after the salt distillation steps. The influence of the U–Fe layer on the electrolysis steps needs to be studied further in order to understand and quantify the lifetime of a single STS wire mesh basket during pyroprocessing.</P> <P><B>Highlights</B></P> <P> <UL> <LI> Use of a single fuel containment basket demonstrated for pyroprocessing spent fuel. </LI> <LI> Achieved UO<SUB>2</SUB> to U conversion rates of over 99%. </LI> <LI> Unstable U–Fe layer observed on the fuel containing basket with repeated use. </LI> <LI> Studying the stability of the U–Fe layer key to determining lifetime of the basket. </LI> </UL> </P>

Investigations on the Pu-to-244Cm ratio method for Pu accountancy in pyroprocessing

Sunil S. Chirayath,부혁진,우승민 한국원자력학회 2023 Nuclear Engineering and Technology Vol.55 No.10

Non-uniformity of Pu and Cm composition in used nuclear fuel was analyzed to determine its effect on Pu accountancy in pyroprocessing, while employing the Pu-to-244Cm ratio method. Burnup simulation of a typical pressurized water reactor fuel assembly, required for the analysis, was carried out using MCNP code. Used fuel nuclide composition, as a function of nine axial and two radial meshes, were evaluated. The axial variation of neutron flux and self-shielding effects were found to affect the uniformity of Pu and Cm compositions and in turn the Pu-to-244Cm ratio. However, the results of the study showed that these non-uniformities do not affect the use of Pu-to-244Cm ratio method for Pu accountancy, if the measurement samples are drawn from the voloxidized powder at the feed step of pyroprocessing. ‘Material Unaccounted For’ and its uncertainty estimates are also presented for a pyrprocessing facility to verify safeguards monitoring requirements of the IAEA.

금속산화물의 열역학적 휘발 거동 및 휘발 산화 공정의 조건 분석

이영우,박소영,박병흥,Lee, Young Woo,Park, So Young,Park, Byung Heung 한국교통대학교 융복합기술연구소 2013 융ㆍ복합기술연구소 논문집 Vol.3 No.2

Metal oxides are known as stable materials during a thermal treatment. However, some oxides are readily evaporated at high temperatures. A voloxidation process is a head-end process for a pyroprocessing dealing with spent nuclear fuels (SF). In SFs, fission productions are in the form of oxides and some of them would be evaporated during the voloxidation process. Therefore, it is of importance to analyse the vapor pressures of metal oxides so that the material flows throughout the pyroprocessing could be estimated. In this work, vapor pressures of relevant metal oxides were calculated and presented to draw a baseline on the material flow of the pyroprocessing.

Rare earth removal from pyroprocessing fuel product for preparing MSR fuel

윤달성,백승우,이창화 한국원자력학회 2024 Nuclear Engineering and Technology Vol.56 No.3

A series of experiments were performed to produce a fuel source for a molten salt reactor (MSR) through pyroprocessing technology. A simulated LiCl–KCl–UCl3-NdCl3 salt system was prepared, and the U element was fully recovered using a liquid cadmium cathode (LCC) by applying a constant current. As a result, the salt was purified with an UCl3 concentration lower than 100 ppm. Subsequently, the U/RE ingot was prepared by melting U and RE metals in Y2O3 crucible at 1473 K as a surrogate for RE-rich ingot product from pyroprocessing. The produced ingot was sliced and used as a working electrode in LiCl–KCl–LaCl3 salt. Only RE elements were then anodically dissolved by applying potential at 1.7 V versus Ag/AgCl reference electrode. The RE-removed ingot product was used to produce UCl3 via the reaction with NH4Cl in a sealed reactor.

Numerical Heat Transfer Analysis of the Electrowinning Cell in the Pyroprocessing

윤달성,백승우,김시형,김광락,안도희 한국방사성폐기물학회 2009 방사성폐기물학회지 Vol.7 No.4

전해제련 공정은 악티늄족 원소를 동시에 회수하는 공정으로써, Pyroprocessing의 핵확산 저항성을 보장 하는 중요한 공정이다. 공학규모의 전해제련 장치를 설계하기 위한 기본 도구를 개발하기 위해서 실험실 규 모의 장치에 대한 열전달 해석을 수행하였다. 열전달 해석을 수치 해석적으로 계산하기 위해 ANSYS CXF 상 용 코드를 사용하였다. 열전달 해석 결과, 가열부의 길이가 수직으로 용융염의 높이보다 약 3배 이상이 되었 을 때, 용융염의 온도를 일정하게 유지할 수 있었으며, 냉각부의 길이는 그 영향이 미비하였다. 전해조 덮개아래의 아르곤 가스의 온도는 냉각 판의 개수에 따라 감소하였으며, 5개 이상 설치 할 경우 250 ℃ 이하로 유 지할 수 있음을 보였다. 이러한 계산 결과는 실제 실험 장치에서 측정된 장치 내부 온도 분포와 경향성이 일 치하는 것을 볼 수 있었다. 본 연구에서 해석된 전해제련 장치의 열 분포 특성은 공학규모 장치의 설계를 위 해 중요한 자료로 사용 될 수 있을 것이다.

A CONCEPTUAL STUDY OF PYROPROCESSING FOR RECOVERING ACTINIDES FROM SPENT OXIDE FUELS

유재형,CHUNG-SEOK SEO,EUNG-HO KIM,HAN-SOO LEE 한국원자력학회 2008 Nuclear Engineering and Technology Vol.40 No.7

In this study, a conceptual pyroprocess flowsheet has been devised by combining several dry-type unit processes; its applicability as an alternative fuel cycle technology was analyzed. A key point in the evaluation of its applicability to the fuel cycle was the recovery yield of fissile materials from spent fuels as well as the proliferation resistance of the process. The recovery yields of uranium and transuranic elements (TRU) were obtained from a material balance for every unit process composing the whole pyroprocess. The material balances for several elemental groups of interest such as uranium, TRU, rare earth, gaseous fission products, and heat generating elements were calculated on the basis of the knowledge base that is available from domestic and foreign experimental results or technical information presented in open literature. The calculated result of the material balance revealed that uranium and TRU could be recovered at 98.0% and 97.0%, respectively, from a typical PWR spent fuel. Furthermore, the anticipated TRU product was found to emit a non-negligible level of γ-ray and a significantly higher level of neutrons compared to that of a typical plutonium product obtained from the PUREX process. The results indicate that the product from this conceptual pyroprocessing should be handled in a shielded cell and that this will contribute favorably to retaining proliferation resistance.