Conference Agenda

The TRIGA Mark-II research reactor is a globally recognized pool-type model supplied by General Atomics of the United States since the late 1950s. Beginning with its first international supply to Italy in 1959, this reactor type was commissioned in the Republic of Korea (KRR-1) in 1962 and in Slovenia (JSI TRIGA) in 1966, establishing a cornerstone for nuclear research, isotope production, and workforce training. While the design life of these reactors was typically 30 to 40 years, the unique operational characteristics and low thermal power of research facilities have allowed many units to operate far beyond their initial structural limits. Today, these reactors have either entered or are approaching the decommissioning phase. In Slovenia, the recent decision to extend the operation of the Jožef Stefan Institute’s TRIGA reactor until 2043 means that at least 15 years remain before the commencement of decommissioning. However, given the myriad of technical variables inherent in the decommissioning process, proactive preparation is indispensable for ensuring a seamless transition. KRR-1 in the Republic of Korea, which is identical in design to JSI TRIGA, began its decommissioning in 2000 and is currently scheduled for unrestricted site release by 2027. With an initial target completion date as early as 2008, the actual process has spanned nearly 30 years. This study analyzes the KRR-1 experience to identify the key considerations that affected the decommissioning process and shares lessons learned to facilitate the successful decommissioning of the JSI TRIGA.

First, the project prioritized the minimization of waste through active decontamination and clearance. Approximately 82% of the total generated decommissioning waste (about 2,600 drums or 520 tons) was disposed of as clearance waste, while the remaining 18% was disposed of in a Low-and-Intermediate Level Waste (LILW) repository. While the operator’s initial projection for clearance was at 70–80%, the project surpassed this target through the extensive decontamination of bulk concrete and soil waste. By downgrading the contamination level from LILW to clearance through decontamination, the project prevented the unnecessary and excessive facility disposal of radioactive waste, thereby significantly securing the disposal margin of the LILW repository.

Second, the study highlights the development of an isotopic inventory evaluation model tailored to the unique characteristics of research reactors. Unlike commercial power plants with standardized operational patterns, KRR-1 featured irregular activation patterns and varied experimental configurations. Therefore, the repository operator required an isotopic inventory assessment that specifically reflected these research-scale characteristics. However, due to the failure to present a sufficiently reliable evaluation model at the outset, the waste was initially rejected for acceptance. This necessitated the construction of a new, validated isotopic inventory evaluation model by backtracking through decades of operational records, which led to a substantial and unforeseen extension of the overall decommissioning period. This case underscores that technical readiness for waste characterization must precede physical dismantling to prevent the decommissioning process from stalling.

Third, the analysis concerns the management and acceptance of "mixed waste." During the decommissioning process of KRR-1, various materials such as lead, graphite, and aluminum were generated as radioactive waste. These materials possess not only radiological hazards but also secondary risks including chemical toxicity, flammability, and gas generation, which could compromise the integrity of the LILW repository. In this instance, various mixed wastes were generated without a comprehensive initial identification of their types and origins. Attempting to dispose of these mixed wastes alongside general radioactive waste without a dedicated evaluation or consideration of their specific characteristics resulted rejection of waste acceptance due to non-compliance with the Waste Acceptance Criteria (WAC). Even today, significant quantities of these mixed wastes remain in temporary on-site storage. Although consultations between the reactor operator and the repository operator are ongoing, technical hurdles such as the need for re-evaluation of repository safety assessments and the necessity of revising the WAC have left the issue unresolved. This case suggests that demonstrating the long-term safety of such complex waste is a time-consuming process; the failure to consider disposal pathways for mixed wastes in advance was a primary driver of the project’s long-term delay.In conclusion, the KRR-1 experience demonstrates that the successful decommissioning of a TRIGA Mark-II research reactor is not merely a task of mechanical dismantling but a complex endeavor requiring early-stage regulatory and technical harmonization. To ensure a planned and efficient decommissioning and waste management process, it is critical to proactively predict and catalogue the physical, chemical, and radiological characteristics of the various waste types that will be generated. Furthermore, a life-cycle approach is essential, requiring active coordination from the initial facility design stage to ensure that decommissioning waste characteristics are fully integrated into the WAC. By incorporating both the successes and the limitations identified in this case study, the decommissioning of the JSI TRIGA can be conducted by facilitating systematic, expeditious, and safe decommissioning activities alongside efficient waste management, based on these preemptive insights.

As many countries embark on an ambitious expansion or renewal of their nuclear fleets, the question of spent nuclear fuel (SNF) management is becoming increasingly central. Beyond the focus on reactor technologies deployment, the credibility of new nuclear programs depends on the ability to define and implement robust, acceptable, and financially sustainable fuel cycle and waste management strategies.

While SNF interim storage solutions have enabled safe nuclear electricity production for decades, they were never designed to serve as indefinite solutions. Deep Geological Repositories (DGR) disposal is widely recognized and established as the reference long-term pathway for Intermediate & High-Level Waste (ILW & HLW) elimination. However, as illustrated by leading projects in Finland, Sweden or France, the implementation of DGR remains complex, capital-intensive, subject to delays and some financial uncertainties.

The core challenge for evaluating SNF and I&HLW disposal routes for new nuclear projects is not the absence of technical solutions, but the growing misalignment between the initial Business Plan hypotheses in an uncertain future:

- SNF pools are progressively reaching saturation, while dry storage facilities are constrained by both design lifetimes and SNF long-term characterization challenges.
Uncertainties surrounding the commissioning schedules of DGRs create a structural gap. Any delay directly translates into additional operational, maintenance, and monitoring costs for interim storage systems.

- These uncertainties are not purely technical, but also financial and societal. In Slovenia, more than 38% of citizens perceive spent fuel storage as a challenging aspect against nuclear energy, highlighting the importance of public acceptance in shaping long-term strategies.

In addition, concepts such as shared or multinational repositories, although attractive in principle, have so far proven difficult to implement due to political, regulatory, and financial considerations.

In this context, SNF reprocessing offers a strategic and industrial lever to restore flexibility and improve the predictability of SNF management pathways.

By enabling the separation of valuable materials (Uranium and Plutonium) and the conditioning of waste, reprocessing significantly reduces both the volume by a factor of 5 and the radiotoxicity of ultimate waste by a factor of 10. This directly alleviates and delays constraints on national strategies planning for “short-term” storage systems and DGRs.

Reprocessing also contributes to the desaturation of spent fuel pools in the short to medium term, providing operators with additional time to align infrastructure development with actual needs.

From a financial perspective, it allows for an optimization of the CAPEX timing. By smoothing capital expenditures over time and leveraging the dynamics of dedicated provision funds, operators can strengthen their capacity to finance future DGR projects under more favorable conditions while reducing the risks of SNF management, replaced by optimized waste forms. Such “reprocessed waste forms” (i.e. vitrified and compacted waste canisters) benefit from well-characterized thermal and radiotoxic behavior over more than a century, offering a high level of predictability for long-term management.

Finally, from the perspective of potential shared storage or disposal solutions, the standardization and optimization of waste packages resulting from reprocessing represent a key advantage, both technically and politically.

In a context where uncertainty is becoming the main driver of risk in SNF management, the ability to introduce flexibility and improve predictability is essential.

Reprocessing should not be seen as an alternative to geological disposal, but as a complementary and enabling solution.

By integrating reprocessing into a comprehensive SNF management strategy, nuclear stakeholders can move from a constrained and reactive approach to a more controlled, resilient, and sustainable trajectory.

In the state of the art, the characterisation of Low and Intermediate Level Waste (LILW) is focused on the determination of gamma emitters. However, limits in the amount of alpha activity in LILW repositories also require the determination of actinides. is an expensive and time-consuming process. This is largely driven by the need to properly determine the concentration of actinides, most of them alpha-emitters, in the waste packages. The short range of alpha-particles in matter forces the use of destructive measurement techniques, that require sampling and radiochemical analysis, to determine the concentration of actinides relative to that of measurable gamma-emitting isotopes (Scale Factors), such as Cs-137 or Co-60. This process presents a challenge for the Spanish authorities, as the inventory of LILW generated in the country is expected to double in the following years.

Passive neutron counting presents an opportunity to streamline and save cost in the characterisation of LILW by detecting neutrons from (alpha,n) reactions of directly form neutron-emitting actinides (Cm-244). This characterisation technique can, potentially, be used to characterise great volumes of LILW through non-destructive measurements, due to the larger travel distances of neutrons in matter, specially metals. In addition, since there is no need for sampling and chemical analysis, the measurements can be performed on-site, further streamlining the characterisation process. However, some challenges are expected. Mainly, a low neutron yield is expected from actinide sources, which may be comparable to the neutron background at declassification activity levels (~1 Bq/g).

This work studies the viability of passive neutron detection for the characterisation of large volumes of LILW. For this purpose, a neutron detection system based on He-3 detector arrays was designed and evaluated using Monte Carlo (MC) simulation tools. Several preliminary results have been achieved so far. First, the system’s response was calculated for the dominant neutron production mechanisms in LILW, namely (α,n) reactions and spontaneous fission, across different waste container materials commonly found in LILW. Additionally, using a mixture of MC simulations and experimental measurements, the system’s response to background neutrons was estimated. For fission sources, the system’s response in coincidence counting mode was calculated as well, as this technique is expected to offer improved background discrimination levels. Minimum Detectable Activities (MDA) of actinides in LILW waste containers, in function of the measurement time have been determined in this way. The final aim is designing a detection system sensible enough to characterise LILW down to de declassification level.