The radiological characterization includes: the collection and review of historical file; the performing calculation of radionuclide inventory in the reactor; in situ measurement; sampling analyses; the review and evaluation of the data obtained; the comparison of calculated result with measured date etc. The special attention should put to the information from the key part in reactor for end state radiological characterization. The sampling from the “hot spot” should not be lost; the number of the sampling should reasonable base on reliable statistics. The radioactivity density for site release should comply with the guide, standard and regulation of international atomic energy agency and our country.

The ROXf is a kind of inert matrix fuel U-238-free matrices, it has a high plutonium transmutation capability. The ROXf consists of chemically stable phases of fluorite ‘stabilized’ ZrO 2 or ThO 2 , and spinel MgAl 2 O 4 . In this fuel, PuO 2 is solidified in a fluorite phase. With U-238-Free matrices, a large part of the plutonium can be burned after irradiation in conventional LWRs. The spent ROXf consists of natural analogous geologically stable phases, and is disposed directly as high level wastes ‘HLWs’ after about 50 years cooling. From the high plutonium burn up rate and the high stability of the fuel, the ROXf-LWRs system has proliferation resistance and environmental safety. Characteristics of two types of ROXf, Zr-ROX and Th –ROX with weapons-Pu, in an LWR core arrangement are evaluated by cell burn up calculations and 2-D core calculations using the SRAC code system and JENDL-3 nuclear library. In an LWR of moderator to fuel volume ratio = 1.9, which corresponds to current PWRs, Pu transmutation rates with the two types of ROXf are large enough and more than 80% and 99% of Pu and Pu-239, respectively, can be burned. The calculated kinetic parameters indicate less moderate characteristics of ROX cores, especially with Zr-ROX. The fertile Th-232 in Th-ROX works like U-238 in the UO 2 fuel, making kinetic parameters more moderate and reactivity drop due to burn up smaller than that Zr-ROX. The neutron capture of Th-232 to generate U-233, causes the safeguards problem. Thus, the characteristics of Zr-ROX as a typical example, were investigated in this study.

A large number of nuclear reactors with graphite as moderator and reflector material are facing to be decommissioned now or later, and the radioactive graphite waste is a large part of the involved wastes. In addition, high temperature gas-cooled reactors being developed rapidly use a large quantity of graphite material (up to 95%) in the nuclear fuel elements, besides graphite material as their moderator and reflector material in the reactor cores. Therefore, it is very critical to manage these graphite wastes from the decommissioned and being decommissioned reactors. The part with low-level radioactive contamination that could not be reused now, may be disposed of as solid waste to reduce its volume, and the possibility of its being retrieved and reused in the future with advanced technology should be considered. The other graphite waste with high-level radioactive contamination requires much more consideration. Due to several factors, such as its large quantity, a lack of available disposal sites and public acceptance, it may not be disposed of directly in the repository any more. An option may be the transformation of the high-level radioactive graphite waste into low-level radioactive waste through physical and chemical processes. The current technologies involve, e.g., thermal treatment to release 36 Cl, capture of the 14 C from the gases of incineration of carbon material and decomposition of carbon dioxide into solid carbon. After these treatments the carbon material might be decontaminated and separated as low-level radioactive waste and a small amount of residual high-level waste could be disposed of ultimately. In order to achieve a sustainable development of graphite material, the maximum utility and the minimal disposal of radioactive graphite should be considered in the management of radioactive graphite waste. It is urgent to explore new technologies for decontaminating and recycling radioactive graphite.

A fine crystalline ammonium tungstophosphate (AWP) exchanger with high selectivity toward Cs + was encapsulated in biopolymer matrices (calcium alginate, CaALG). The characterization of the AWP-CaALG microcapsule was examined using SEM/WDS, IR and DTA/TG analyses, and the selective separation and recovery of 137 Cs were examined by the batch and column methods using simulated and real high-level liquid waste (HLLW). The free energy (Δ G 0 ) of the ion exchange ( NH 4 + ↔ Cs + ) for fine AWP crystals was determined at −13.2 kJ/mol, indicating the high selectivity of AWP towards Cs + . Spherical and elastic AWP-CaALG microcapsules (∼700 μm in diameter) were obtained and fine AWP crystals were uniformly immobilized in alginate matrices. Relatively large K d values of Cs + above 10 5 cm 3 /g were obtained in the presence of 10 −3 ∼1 M Ca(NO 3 ) 2 , resulting in a separation factor of Cs/Rb exceeding 10 2 . The irradiated samples ( 60 Co, 17.6 kGy) also exhibited large K d values exceeding 10 5 cm 3 /g in the presence of 2.5 M HNO 3 . The K d values in the presence of 0.1–9 M HNO 3 for 67 elements were determined and the order of K d value was Cs + ≫ Rb + > Ag + . The breakthrough curve of Cs + had an S-shaped profile, and the breakpoint increased with decreasing flow rate; the breakpoint and breakthrough capacity at a flow rate of 0.35 cm 3 /min for the column (0.7 g AWP-CaALG) were estimated at 25.2 cm 3 and 0.068 mmol/g, respectively. Good breakthrough and elution properties were retained even after thrice-repeated runs. The uptake (%) of Cs + in simulated HLLW (28 metal components-1.92 M HNO 3 , SW-11, JAEA) was estimated at 97%, and the distribution of Cs + and Zr/Ru into the AWP and alginate phases, respectively, were observed by WDS analysis. Further, the selective uptake of 137 Cs exceeding 99% was confirmed by using real HLLW (FBR “JOYO”, JAEA). The AWP-CaALG microcapsules are thus effective for the selective separation and recovery of Cs + from HLLWs.

Reduction of radioactive waste volume is one of the important issues for spent fuel reprocessing plants. JGC has intensively developed radioactive waste processing technologies aiming at waste volume reduction and at immobilization with environmentally low impact for Purex type reprocessing plants. These technologies are Electrolytic Decomposition Process of Nitric Acid (E-DeP), Catalytic Decomposition Process of Nitrate Ion (C-DeP) and Electrodialytic Na Separation Process (E-SeP). An effective combination of each technology results in reduction of numbers of waste packages by one-twentieth.

Core loading pattern design has great influence on nuclear power plant operation. An excellent core loading pattern can not only enhance operation factor, reduce operation cost, but also increase operation safety. Under the premise of nuclear safety, AP1000 first core loading pattern achieves the goal of low leakage loading by simulating the reactivity distribution of the 18-month Equilibrium Cycle design. The fuel management presented in this paper illustrates the economic performance and technical feasibility of the advanced 18-month cycle first core fuel. The advanced feature of this first core design include: the development of loading pattern that simulates the reactivity distribution of a typical low leakage reload core in order to reduced leakage, the use of radial enrichment zoning in higher enriched assemblies to lower peaking factors, and the use of axial burnable absorber zoning to improve axial power shape control and reduce axial peaking. The discussion provided in this paper demonstrates the ability of the advanced first core design operate safety and efficiently, and the core is designed with adequate peaking factor margin in both base-load and load-follow operation. Finally, this paper analyses the impact brought about by multi-enrichment on the nuclear power plant operation capacity and operation cost, and arrive at a conclusion that precise fuel cycle evaluation such that the enrich uranium and operation costs can be accurately quantified and control in a detailed economic evaluation.

Reactivity Initiated Accident (RIA) leads to an unwanted increase in fission rate and power in a region of the reactor core confined around the position of occurrence. The power excursion due to such events may cause fuel rods failures and a subsequent release of radioactive material into the primary coolant of reactor, in severe cases, this release could damage nearby fuel assemblies. In nuclear power plants, RIAs are due to control system faults, e. g. control elements ejection/insertion, or rapid changes in temperature or pressure of moderator. In Boiling Water Reactors (BWRs), the control rod drop accidents (RDAs) at cold zero power have been deeply investigated, in fact, notwithstanding they are less frequent in comparison with the control rod ejection event in PWRs, in this kind of plant these conditions are the most severe in case of a RIA occurrence. RDA transient, comprised in the design basis events considered in safety analysis, may cause rod failures especially at high burnup. To simulate a RIA, a peaked power pulse is applied to a pre-irradiated and re-instrumented rodlet aiming at investigating the most important phenomena that could lead to the rupture of cladding tubes. This paper is focused on the investigation of the TRANSURANUS fuel performance code capability to predict the thermomechanical state of rodlets subjected to RIA tests. To this purpose the FK-1 test, carried out at the Nuclear Safety Research Reactor (NSRR) of Japan Atomic Energy Research Institute (JAERI), was simulated. This experiment is part of a set of 12 tests performed at the NSRR facility to study the performance under a reactivity initiated accident of BWR rodlets with burnup between 41 and 61 MWd/kg HM . In the FK-1 test, a STEP I BWR rodlet, previously irradiated in the Fukushima Daiichi Nuclear Power Station (Unit 3) operated by the Tokyo Electric Power COmpany (TEPCO) up to 45 MWd/kg HM , was subjected to a peak enthalpy insertion of 544 J/g. In this paper the code findings for the FK-1 test are discussed on the basis of the experimental data and the predictions of other stand-alone codes for transient analysis. The FK-1 predictions of FRAPTRAN (2001), FALCON (2003) and SCANAIR (ver. 3–2) are reported. The choice of fuel relocation model and important cladding properties (swelling, thermal expansion, thermal conductivity) was made relying on preliminary calculations whose results are also presented. Notwithstanding a satisfactory agreement between predictions and experimental data and a good agreement in the presented code-to-code comparison were envisaged, these results also emphasized the need to improve the models for FGR, heat transfer to plenum. Investigations are also required to ascertain possible contribution from fission gas to pelles thermal expansion. Ongoing modeling activity, performed at the ITU Joint Research Centre, is focused on a new model for FGR, ENEA (Agenzia nazionale per le nuove tecnologie, l’energia e lo sviluppo economico sostenibile) is expected, in the near term, to give a contribution to refine the model for plenum gas temperature. These activities should improve the description of RIA transient and further investigations on NSRR tests will be performed with newly developed models. The work presented in this paper will be part of ENEA contribution in FUMEX III project leaded by the International Atomic Energy Agency (IAEA) and aimed at the improvement of fuel codes predictions at high burnup.

The removal efficiencies of strontium, cesium and cobalt by inorganic membrane with cut-off 8kD were studied. In order to improve the removal efficiency of non-active nuclides, soluble macropolymers sodium poly(acrylic) acid (NaPAA) of different molecule sizes were selected as chemical assistant reagents to make comparative experimental studies. The molecule of 8000,50000 and 100000 D NaPAA were used. The flux and removal rates of different systems were mainly studied. The results show that for inorganic membrane with pore size of 8kD molecule of 50000 Da NaPAA is the best selection for assistant reagents.

The HTR-10 is a pebble bed High-temperature gas-cooled reactor (HTGR) with a nominal thermal power of 10 MW. The on-line nondestructive burn-up (BU) measurement for the spherical fuel elements with TRISO coated fuel particles is an important issue in the fuel management of the HTR-10. The HTR-10 employs an HPGe gamma spectroscopy system to determine the BU values of the discharged fuel elements, and the accuracy of the BU measurement system is a crucial issue. The calibration requirement and the absolute calibration method are discussed and analyzed in this paper. In this work, the long-lived fission product 137 Cs is demonstrated as the best BU indicator in the HTR-10, whereas the inner-calibrating source method using the ratio of the amounts of 134 Cs to 137 Cs is not suitable. The BU value of a fuel element is proportional to the intensity of 137 Cs gamma-ray and independent on the irradiation history of the fuel element, except for a small correction term related to the radioactive decay of 137 Cs. The efficiency of the measurement system is calibrated by using a 137 Cs standard source embedded into the center of a graphite sphere with the same size and the same graphite material as the fuel element, which was located at the measurement position of the fuel sphere, to eliminate the possible systematic errors from the measurement system. The difference in geometry and self-absorption between the BU measurement and the calibration measurement is corrected by applying the MCNP simulation. The error sources of the calibration results are analyzed, in which the uncertainty of the source activity (∼5%) is found to dominate. The other major error sources include the fluctuation of the uranium loading in the fuel elements, the errors of the nuclear data, the counting errors of the detector, and the correction procedure for the detection efficiency. More than 1,900 fuel elements discharged from the HTR-10 have been measured by far, and the BU values of these elements spread in the range from 8.4±0.5 GWD/t to 20±1.0 GWD/t, where the relative errors of the burn-ups consist of the calibration errors about 5.2% and the counting errors around the order of 1%. The behavior of the measured burn-ups has been discussed qualitatively.

This paper presents the design of the low-level solid waste encapsulation device for the 200L barrels. The device is used to compress and package the low-level contaminated waste such as cotton goods, rubber, plastic products, leather products and paper products.

Water-soluble oxa-diamide ligand, N,N,N ′ ,N ′ -tetra-methyl-3-oxy-pentane-1,5-diamid (TMPDA) has been synthesized and purified. Its crystal structure, melting point, decomposition temperature, solubilities in aqueous phase and organic phase, distribution ratio between aqueous and organic phase, etc. are reported. The effect of TMPDA concentration in aqueous phase and HNO 3 concentration in the equilibrium aqueous phase on the extraction efficiency of La(III), Ce(III), Pr(III), Nd(III), Zr(IV), Fe(III), Y(III), Mo(VI), Ru(III) and Pd(II) by 30% TRPO/kerosene have been studied. The results indicate that TMPDA dissolve well in aqueous phase but almost insoluble in kerosene or 30%TRPO/kerosene in the bi-phase system. It can effectively reduce the extraction of Ln(III), Y(III) and Zr(IV) into 30%TRPO/kerosene at a moderate acid system (0.24mol/L∼0.27mol/L HNO 3 ). TMPDA is a promising stripping agent for Ln(III), Y(III) and Zr(IV) from loaded TRPO.

AREVA as a worldwide PWR fuel provider has to have a fleet of fresh UO 2 shipping casks being agreed within a lot of countries and to accommodate foreseen EPR™ Nuclear Power Plants fuel buildings. To reach this target the AREVA NP Fuel Sector together with TN International (a subsidiary of AREVA NC) decided to develop an up to date shipping cask gathering experience feedback of the today fleet and an improved safety allowing the design to comply with international regulations (NRC and IAEA) and other local foreign Safety Authorities requirements. Based on pre design features a safety case was set up to highlight safety margins. Criticality hypothetical accidental assumptions were defined: • Preferential flooding, • Fuel rod array expansion (so called “bird caging” effect), • Fuel sliding, • Neutron absorber penalty, • …. The French criticality code package CRISTAL is used to check several configurations reactivity and derived safety margins. The CRISTAL code package relies on a wide validation process helping defining uncertainties as required by regulations in force. The shipping cask, containing two fuel assemblies, is designed to maximize fuel isolation inside the cask and with neighboring ones even for large array configuration cases. Few and proven industrial products are used: • Stainless steel for the structural frame, • Balsa wood for impact limiters, • BORA ® resin as neutron absorber. The cask is designed to handle mainly the EPR™ fuel assembly type and may be extended to other contents such as APWR fuel assembly type. After a brief presentation of the computer codes and the description of the shipping cask, the CRISTAL calculation results as well as the applied uncertainties will be discussed.

With world stockpiles of used nuclear fuel increasing, the need to address the long-term utilization of this resource is being studied. Many of the transuranic (TRU) actinides in nuclear spent fuel produce decay heat for long durations, resulting in significant nuclear waste management challenges. These actinides can be transmuted to shorter-lived isotopes to reduce the decay heat period or consumed as fuel in a CANDU ® reactor. Many of the design features of the CANDU reactor make it uniquely adaptable to actinide transmutation. The small, simple fuel bundle simplifies the fabrication and handling of active fuels. Online refuelling allows precise management of core reactivity and separate insertion of the actinides and fuel bundles into the core. The high neutron economy of the CANDU reactor results in high TRU destruction to fissile-loading ratio. This paper provides a summary of actinide transmutation schemes that have been studied in CANDU reactors at AECL, including the works performed in the past [1–4]. The schemes studied include homogeneous scenarios in which actinides are uniformly distributed in all fuel bundles in the reactor, as well as heterogeneous scenarios in which dedicated channels in the reactor are loaded with actinide targets and the rest of the reactor is loaded with fuel. The transmutation schemes that are presented reflect several different partitioning schemes. Separation of americium, often with curium, from the other actinides enables targeted destruction of americium, which is a main contributor to the decay heat 100 to 1000 years after discharge from the reactor. Another scheme is group-extracted transuranic elements, in which all of the transuranic elements, plutonium (Pu), neptunium (Np), americium (Am), and curium (Cm) are extracted together and then transmuted. This paper also addresses ways of utilizing the recycled uranium, another stream from the separation of spent nuclear fuel, in order to drive the transmutation of other actinides.

By proposing a credible response to the growing demand for energy while emitting no greenhouse gas, nuclear power will more than likely expand in the future, leading to increased quantities of used nuclear fuel to manage. Recycling the energy from this used fuel and an efficient waste management are key components for a sustainable development of nuclear energy. Current closed fuel cycle enables 25% in uranium savings, reduces the volume of waste by a factor 5 while its radiotoxicity is divided by a factor 10. Excellent track records of existing recycling plants participate to the competitiveness of MOX (Mixed Oxide) and ERU (Enriched Reprocessed Uranium) fuels compared to ENU (Enriched Natural Uranium) fuel. By offering a sound solution for nuclear waste management, recycling also contribute to favour nuclear acceptance. The advantages of closed fuel cycle have been demonstrated by the successful policy of recycling implemented in Europe for more than 30 years, with 35 reactors using MOX fuel with an excellent return of experience. The AREVA industrial recycling platform (La Hague and Melox plants) has treated more than 24,500 tons of used fuel and fabricated more than 1,600 tons of MOX fuel. With the new AREVA EPR™ reactor, recycling will take another step forward, enabling MOX fuel loading of up to 100%, thus offering optimized management options of recycled fuel, and giving more flexibility to its customers.

The sorption of 131 I − , 131 IO 3 − , I 2 and CH 3 131 I from water solutions at 25°C on new composite materials obtained by modifying of cation-exchange resin KU-2 was investigated. It was established, that the given materials are capable to absorb I 2 both from distilled water, and from a water coolant of the WWER-type NPPs, with distribution factors K d more than 10 3 cm 3 /g at V/m = 100. Thus, it was found, that practically full I 2 absorption (more than 95.0%) was achieved for 15 min. It was shown, that anion-exchange resin AV-18 is capable to adsorb CH 3 131 I from a water solutions with distribution factors K d more than 300 cm 3 /g at V/m = 100.

The R&D of coating process of fuel particles is one of the most important key technologies in the research work of high-temperature gas-cooled reactor (HTGR). A safe and reliable coating process is expected and related to the prospect of large-scale utilization of nuclear energy. The related research of the carbon black powder which is the main byproduct generated in the coating process is important, because it relates to the impact of coating process on the surrounding environment and is also helpful to understand the deposit mechanism of PyC (pyrolytic carbon) layer coated on the fuel particle. In the present study, the microstructure of the pyrolytic carbon powder were systematically investigated by scanning electron microscope (SEM), transmission electron microscopy (TEM), laser Raman spectroscopy and particle size analysis (PSA). It can be found that the carbon powder in the cyclone separator is composed of the nano-spherical particles with the diameter of about 50nm. The ring-layered nano-structured carbon particles could be found from the electron micrographs. The comparison between Raman spectra of carbon powder and pyrolytic carbon coated on the fuel particle showed that the droplet deposit mechanism was suitable to explain the PyC deposition process. The nano-particles agglomerate into the irregular floc and the diameters of the stable particle clusters are mainly 1 μm and 10 μm. The disposal methods of carbon black powder are also discussed.

With an increased population and an increasing demand for power, nuclear power has attracted an increasing attention and mass nuclear power plant have been built in different countries in the past several decades. At present, about ten thousands ton spent fuels are discharged from nuclear power plant every year and the estimated capacity will approximately add up to 5×10 5 ton. Therefore, spent fuel reprocessing, by which the co-extraction and separation as well as purification of Uranium and Plutonium could be realized and ensure the recycle of uranium resources and the management of nuclear waste, is a vital step in nuclear fuel cycle including two major strategies, i.e. once-through cycle and closed fuel cycle. It is worth noting that the utilization of MOX fuel made by plutonium mixed with uranium has been successfully achieved in thermal reactor. Fortunately, the middle experiment plant of china spent fuel reprocessing has been being debugged and will be operated completely in future two years. Various reprocessing schemes have been proposed for the extraction of actinides from fission products and other elements presented in spent nuclear fuel. However, after numerous studies of alternate reprocessing methods and intensive searches for better solvents, the PUREX process remains the prime reprocessing method for spent nuclear fuels throughout the world. High burning and strong radioactive spent fuel resulting from the evolution of various reactors drive the development of the advanced PUREX technology, which emphasizes the separation of neptunium and technetium besides the separation of the Uranium and Plutonium from the majority of highly active fission products. In addition, through Partitioning and Transmutation method, some benefits such as segregating the actinides and long life fission products from the high level waste can be obtained. The GANEX process exploited by CEA, which roots in COEX process belonged to advanced PUREX process, considers the separation of the actinides and long life fission products. The study on the pyro-chemical processing such as the method of electro-deposition from molten salts has still not replaced the traditional PUREX process due to various reasons. In conclusion, the future PUREX process will focus on the modified process including predigesting the technical flowsheets and reducing reprocessing costs and using salt-less reagent in order to minimize the waste production.

Acetohydroxamic acid (AHA) is an organic ligand planned for use in the Uranium Extraction (UREX) process. It reduces neptunium and plutonium, and the resultant hydrophilic complexes are separated from uranium by extraction with tributyl phosphate (TBP) in a hydrocarbon diluent. AHA undergoes hydrolysis to acetic acid which will impede the recycling of nitric acid. During recent discussions of the UREX process, it has been proposed to replace AHA by formohydroxamic acid (FHA). FHA will undergo hydrolysis to formic acid which is volatile, thus allowing the recycling of nitric acid. The reported reduction potentials of AHA and pertechnetate ( TcO 4 − ) indicated that it may be possible for AHA to reduce technetium, altering its fate in the fuel cycle. At UNLV, it has been demonstrated that TcO 4 − undergoes reductive nitrosylation by AHA under a variety of conditions. The resulting divalent technetium is complexed by AHA to form the pseudo-octahedral trans -aquonitrosyl (diacetohydroxamic)-technetium(II) complex ([Tc II (NO)(AHA) 2 H 2 O] + ). In this paper, we are reporting the synthesis of FHA and its complex formation with technetium along with the characterization of FHA crystals achieved by NMR and IR spectroscopy. Two experiments were conducted to investigate the complexation of FHA with Tc and the results were compared with previous data on AHA. The first experiment involved the elution of Tc from a Reillex HP anion exchange resin, and the second one monitored the complexation of technetium with FHA by UV-visible spectrophotometry.

Several experiments have been carried out with a Nd-YAG laser as a future dismantling cutting tool. At the beginning, industrial lasers could cut only small thicknesses due their small power (1 kW) but now, with the rise in industrial power (8 kW), it is possible to cut plates up to 100 mm thickness. This technique is practical with the use of optical fibers which allow to maintain the laser generator in a non-radioactive zone. This paper provides a synthesis of the measurements of aerosols and gases produced in different configurations (cutting in air and underwater) and some comparisons with other cutting tools.

As an alternative grouting material for the geological repository of long-lived radioactive waste, the “ S uper f ine S pherical silica G rout” (SFSG) material is developed using a fine spherical silica and a fine calcium hydroxide. The developed SFSG material takes an advantage of its smaller particle size distribution (max. ∼1 micron or less) than those of the cementitious materials, and also provides a low alkaline environment so as to reduce unfavorable effects on the long-term performance of geological disposal system. The SFSG is a mixture of the “super fine silica powder”, the superfine calcium hydroxide and additives such as superplasticizer. Presently, the mixture being investigated for grouting materials is focused on water/binder ratio (W/B) of 1.2. Some preliminary laboratory experiments were carried out to characterize its fundamental properties from the viewpoint of practical use for geological disposal, which is required to be equivalent with the conventional cementitious materials in terms of penetrability, strength, pH performance and workability. From a series of experiments, it was concluded that SFSG is expected to become an alternative grouting material for a geological repository.