Tinjauan peraturan keselamatan utilisasi dan modifikasi instalasi nuklir nonreaktor

The service life of Non-Reactor Nuclear Installations (INNR) in Indonesia is quite long, so it is essential to make modifications. In addition, developments of technology and market demand with the products could modify INNR, such as replacing and adding a control system to interim storage of nuclear-spent fuel facility (KHIPSB3) [1]. Due to the absence of technical regulations related to the modification and utilization of the new INNR and to provide uniformity of format and content in carrying out the further utilization and modification of the INNR, it is necessary to study the literature, compare regulations and consult with resource persons. From the results of these studies and consultations, an overview and solution of arrangements before, during, and after the new utilization or modification of INNR are obtained, making it easier for permit holders or evaluators to carry out activities related to the further utilization or modification of INNR. From the review results, we can conclude that special regulations related to the safety of new utilization or modification of INNR need to be issued immediately by the regulatory body. This review is expected to be a reference in making regulations for the further utilization and modification of INNR, which regulates the format and content of the modification or new utilization of INNR, which becomes a guideline for licensee and evaluators in implementing further utilization and modification of INNR. Keywords: modification, utilization, INNR

Radioactive Fission Waste from Molybdenum-99 Production and Proliferation Risks

Molybdenum-99 (99Mo) is a parent radioisotope of Technetium-99m (99mTc) widely used in nuclear diagnostics. The production of this radioisotope by PT. INUKI generated radioactive fission waste (RFW) that theoretically contains239Pu and235U, posing a nuclear proliferation risk. This paper discusses the determination of radionuclides inventory in the RFW and the proposed strategy for its management. The radionuclides inventory in the RFW was calculated using ORIGEN 2.1 code. The input parameters were obtained from one batch of 99Mo production using high enriched uranium in PT. INUKI. The result showed that the RFW contained activation products, actinides, and fission products, including239Pu and235U. This result was then used for consideration of the management of the RFW. The concentration of 235U was reduced by a down-blending method. The proposed strategy to further manage the down-blended RFW was converting it to U3O8 solid form, placed in a canister, and eventually stored in the interim storage for high-level waste located in The Radioactive Waste Technology Center.

Consolidated nuclear waste storage in Andrews, Texas: An integrated technical and policy risk analysis

Over the last 50 years, nuclear energy has reduced US energy-related CO2 emissions by over 30 gigatons compared to if the same electricity were produced by fossil fuels such as coal and natural gas. However, many kilotons of spent nuclear fuel have accumulated at different sites across the country, and sociopolitical factors have frustrated efforts to address the challenge of nuclear waste disposal. Presently, a consolidated interim storage facility in Andrews, Texas, provides a promising temporary solution. In this paper, we compare the technical and policy risks of the project to continued storage at independent spent fuel storage installations. Our results indicate that the cost of the radiological risk is low (<$30,000) for both scenarios. However, policy and societal considerations will impact the viability of the proposed consolidated interim storage facility. The safety and suitability of this interim storage facility will be affected by when a permanent repository becomes available, whether insurance for offsite waste storage is available, and the impact of climate risks. Although a consolidated interim storage facility at Andrews can potentially serve as a safe and economically advantageous solution, we highlight why these concerns must be addressed for the successful implementation of this facility, and more broadly for the future of the US nuclear industry.

Development of a mobile, automated, optical inspection system for radioactive barrels

Due to the delayed construction and commissioning of a German repository for intermediate- and low-level radioactive waste, waste inventories from several decades are now located at the interim storage sites, the safe custody of which must also be ensured for an indefinite period of interim storage. The usual practice in the interim storage facilities is recurrent inspections, which are carried out almost exclusively manually and without electronic comparative recordings as well as without mechanical documentation and archiving. Remote or automated inspection does not take place. The inspections are carried out visually and are therefore very subjective and thus subject to errors. Manual performance is labor intensive and requires the use of personnel exposed to radiation. Neither are uniform inspection criteria of the visual inspections applied, nor are the inspections performed uniformly between sites. Based on these facts, the Institute for Technology and Management in Construction, Department of Deconstruction and Decommissioning of Conventional and Nuclear Buildings, together with the Institute for Photogrammetry and Remote Sensing, is developing an automated drum inspection system as part of the funding measure FORKA – Research for the Deconstruction of Nuclear Facilities. EMOS is a mobile inspection unit that remotely and automatically records the entire surface of the drum, including lid and bottom, optically; evaluates it analytically; and both stores it electronically and outputs the results in the form of an inspection report. In this way, recurring inspections of the drum stock can be completed under the same inspection conditions each time. A decisive advantage is the possibility of carrying out the inspection remotely in order to reduce the radiation dose to the employees on site. The optical evaluation, display and output of the results will ensure a more precise inspection and analysis of the drum surfaces through software to be specially developed than is possible through manual and visual inspections as currently performed in the interim storage facilities. The continuous monitoring of the stored drums will be facilitated and also the tracing of possible damage development through the comparison of archived measurement results is a novel and powerful tool that helps to increase and ensure the safety aspects of interim storage in the long term. Changes in drum geometry as well as in the surface condition (e.g. corrosion formation, etc.) can be identified at an early stage with the help of the inspection unit, and measures can be taken at an early stage to counteract the loss of integrity of the storage containers.

Spectroscopic and chemical investigations on volatile fission and activation products within the fuel-cladding interface of irradiated pressurised water reactor fuel rod segments

In Germany, the present waste management concept foresees the direct disposal of spent nuclear fuel (SNF) in deep geological repositories for high-level waste available by 2050, at best. Until then, SNF is encapsulated in dual-purpose casks and stored in dry interim storage facilities. Licenses for both casks and facilities will expire after 40 years following loading of the cask and emplacement of the first cask in the storage location. Yet, due to considerable delays in the site selection process and the estimated duration for construction and commissioning of a final repository of at least 2 decades, a prolonged dry interim storage of SNF is inevitable (ESK, 2015). Concerning these considerable timespans, integrity of the cladding is of utmost importance regarding the ultimately conditioning of the fuel assemblies for final disposal. Various processes strain the structural integrity of Zircaloy cladding during reactor operation and beyond such as delayed hydride cracking, fuel-cladding chemical interactions or irradiation damage induced by α-emitters present in the fuel pellet's rim zone (Ewing, 2015). Especially with higher burn-up, the gap between fuel and cladding closes and results in the formation of an interaction layer, in which precipitates of fission and activation products are present, displaying an interface for degradation processes. For chemical analysis and speciation of these agglomerates, Zircaloy-4 and SNF specimens were sampled from fuel rod segments irradiated in commercial pressurised water reactors during the 1980s. Zircaloy-4 specimens were taken from an UOX (50.4 GWdtHM-1) and mixed oxide fuel (MOX) (38.0 GWdtHM-1). In addition, SNF fragments were sampled from the closed gap of both fuel types to examine volatile activation and fission products, which had been segregated from the centre to the pellet periphery during irradiation and thus contribute to the possible chemically assisted cladding degradation effect of the precipitates within the fuel-cladding interface. Spectroscopic analysis of precipitates within the interface layer between fuel and cladding were performed by optical microscopy, X-ray absorption and X-ray photoelectron spectroscopy, as well as by energy-dispersive scanning electron microscopy. Moreover, the radionuclide inventory of the respective Zircaloy-4, fuel and interaction layers was determined using liquid scintillation counting, γ-spectroscopy, gas mass spectrometry, ion chromatography and inductive-coupled plasma mass spectrometry and compared to results received by MCNP/CINDER and webKORIGEN calculations. In this study, we provide results regarding the speciation and chemical composition of previously identified Cs-U-O-Zr-Cl-I bearing compounds found in the interaction layer of irradiated nuclear fuel and inventory analyses of radionuclides present therein, with particular emphasis on Cl-36 and I-129. Furthermore, the agglomerates within the fuel-cladding interface were characterised for the first time utilising synchrotron radiation-based Cl K-edge and I K-edge measurements, resulting in compounds with structural similarities to CsCl and CsI. The outcomes obtained from this study provide further insights into the complex chemistry within the fuel-cladding interface with respect to the aging management and integrity of SNF under the conditions of interim storage. In future studies we will examine whether the different compounds at the fuel-cladding interface have the potential to affect the mechanical properties of Zircaloy cladding.

Impact of Radiation-Induced Microstructures on the Integrity of Spent Nuclear Fuel (SNF) Elements in Long-Term Storage

The current safety concept provides for a period in the range of 40 years for interim storage of spent fuel elements. Since the requirement for proof of safety for to up to 100 years arises, the integrity of the spent fuel elements in prolonged interim storage and long-term repositories is becoming a critical issue. In response to this safety matter, this study aims to assess the impact of radiation-induced microstructures on the mechanical properties of spent fuel elements, in order to provide reliable structural performance limits and safety margins. The physical processes involved in radiation damage and the effect of radiation damage on mechanical properties are inherently multiscalar and hierarchical. Damage evolution under irradiation begins at the atomic scale, with primary knock-on atoms (PKAs) resulting in displacement cascades (primary damage), followed by the defect clusters leading to microstructural deformations. In this context, we have developed and applied a multiscale simulation methodology consistent with the multistage damage mechanisms and the corresponding effects on the mechanical properties of spent fuel cladding and its integrity. Within the improved hierarchical modelling sequence, the effect of the radiation field on the fuel element cladding material (Zircalloy-4) is assessed using Monte Carlo methods. A molecular dynamics method is employed to model damage formation by PKAs and primary damage defect configurations. The formation of clusters and evolution of microstructures are simulated by extending the simulation sequence to a longer time scale with the kinetic Monte Carlo (KMC) method. Transferring the calculated radiation-induced microstructures into macroscopic quantities is ultimately decisive for the structural/mechanical behaviour and stability of the cladding material, and thus for long-term integrity of the spent fuel elements. Results of the multiscale modelling and simulations as well as a comparison with experimental results will be presented at the conference session.

Management of high-level radioactive waste in Germany: roads from storage towards disposal – need and options for action

The road towards final disposal of high-level radioactive waste (HAW) produced in Germany requires extensive and foresighted management. To date, HAW has been stored in dual-purpose casks inside 15 interim storage facilities. Finally, it is disposed of in a deep geological repository. A site-selection process for this repository, taking into account the whole national territory, started in 2017. The road from interim storage to final disposal is not yet planned in detail: neither temporally nor spatially nor technically. Important parameters are still unknown. The last operating licenses of the existing interim storage facilities, originally built to last for up to 40 years, will end in 2047, and a concept for prolonged interim storage does not exist. The dates for the decision on the repository site and the start of its operation are plagued by uncertainties, as well as the development of safety concepts for different potential host rocks or knowledge on the long-time behavior of disused fuel assemblies during dry interim storage. According to the German site-selection law (Deutscher Bundestag, 2017) the siting decision for the final repository is planned to be made in 2031; Thomauske and Kudla (2016) drew up timelines for the site-selection process to end between 2059 and 2096. The research project WERA – Management of high-level radioactive waste in Germany: Roads from storage towards disposal – addressed these uncertainties through the development of different design options for the four main steps of the German road to disposal and of a variety of scenarios combining these steps, covering a broad range of potential future designs of the road to disposal. These scenarios have been analyzed in detail. Need for technical and political action along the road to final disposal has been identified. Options for action were named, and their preconditions and consequences were listed. The design options and the scenarios derived form the basis of societal discourse on the disposal of high-level radioactive waste. Thus, the research project WERA contributes toward the politically and societally active integration of the different disposal steps (interim storage, receiving storage facility, waste conditioning, and final disposal).