Preliminary Study for Evaluation of Instant Release Fractions for PWR Spent Fuels Based on Literature Review

Kwang Pyo Choi,Ser Gi Hong 한국방사성폐기물학회 2023 한국방사성폐기물학회 학술논문요약집 Vol.21 No.2

While many countries consider direct disposal of the spent nuclear fuels, they need to consider long-term disposal scenarios with severe accidents such as the contact between underwater and the spent nuclear fuel due to large defect of the canister. Radionuclides releases rapidly with contacting water or slowly with dissolution of UO2 matrix. The former is known as the ‘Instant Release’, and the latter is ‘Congruential Release’. Even though the instant release fractions (IRF) are much smaller than the congruential ones, IRF has to be treated carefully due to the fact that the instant releases lead to much larger value of the exposure dose rates than the congruential ones which proceed very slowly. It is known that the exposure dose rates by the instant releases are ~25 times larger than the one by the congruent release. The radionuclides from UO2 matrix migrate to the grain boundary, make bubbles, and make tunnels, which leads to instant releases of some radionuclides. The radionuclides in the gap between UO2 pellet and cladding can be also instantly released. In addition, the radionuclides in the crud are instantly released. But in this paper, nuclides from the crud are not regarded, due to the lack of the leaching data. Meanwhile, there’re some nuclides that released from the construction materials like the cladding, the Rod Cluster Control Assembly (RCCA), or the other metal parts. In this work, IRF values for major IRF nuclides such as Cs, I, Cl, Se for the reference PWR spent fuels of South Korea were evaluated based on the rationale from literatures’ review. In particular, these evaluations were done as the function of fission gas release (FGR), average discharge burnup, and fuel dimensions. In addition, the values of IRF for the other nuclides were also suggested based on the other institutes.

Towards grain-scale modelling of the release of radioactive fission gas from oxide fuel. Part II: Coupling SCIANTIX with TRANSURANUS

Zullo G.,Pizzocri D.,Magni A.,Van Uffelen P.,Schubert A.,Luzzi L. 한국원자력학회 2022 Nuclear Engineering and Technology Vol.54 No.12

The behaviour of the fission gas plays an important role in the fuel rod performance. In a previous work, we presented a physics-based model describing intra- and inter-granular behaviour of radioactive fission gas. The model was implemented in SCIANTIX, a mesoscale module for fission gas behaviour, and assessed against the CONTACT 1 irradiation experiment. In this work, we present the multi-scale coupling between the TRANSURANUS fuel performance code and SCIANTIX, used as mechanistic module for stable and radioactive fission gas behaviour. We exploit the coupled code version to reproduce two integral irradiation experiments involving standard fuel rod segments in steady-state operation (CONTACT 1) and during successive power transients (HATAC C2). The simulation results demonstrate the predictive capabilities of the code coupling and contribute to the integral validation of the models implemented in SCIANTIX.

Application of the SCIANTIX fission gas behaviour module to the integral pin performance in sodium fast reactor irradiation conditions

Magni A.,Pizzocri D.,Luzzi L.,Lainet M.,Michel B. 한국원자력학회 2022 Nuclear Engineering and Technology Vol.54 No.7

The sodium-cooled fast reactor is among the innovative nuclear technologies selected in the framework of the development of Generation IV concepts, allowing the irradiation of uranium-plutonium mixed oxide fuels (MOX). A fundamental step for the safety assessment of MOX-fuelled pins for fast reactor applications is the evaluation, by means of fuel performance codes, of the integral thermal-mechanical behaviour under irradiation, involving the fission gas behaviour and release in the fuel-cladding gap. This work is dedicated to the performance analysis of an inner-core fuel pin representative of the ASTRID sodium-cooled concept design, selected as case study for the benchmark between the GERMINAL and TRANSURANUS fuel performance codes. The focus is on fission gas-related mechanisms and integral outcomes as predicted by means of the SCIANTIX module (allowing the physics-based treatment of inert gas behaviour and release) coupled to both fuel performance codes. The benchmark activity involves the application of both GERMINAL and TRANSURANUS in their “pre-INSPYRE” versions, i.e., adopting the state-of-the-art recommended correlations available in the codes, compared with the “post-INSPYRE” code results, obtained by implementing novel models for MOX fuel properties and phenomena (SCIANTIX included) developed in the framework of the INSPYRE H2020 Project. The SCIANTIX modelling includes the consideration of burst releases of the fission gas stored at the grain boundaries occurring during power transients of shutdown and start-up, whose effect on a fast reactor fuel concept is analysed. A clear need to further extend and validate the SCIANTIX module for application to fast reactor MOX emerges from this work; nevertheless, the GERMINAL-TRANSURANUS benchmark on the ASTRID case study highlights the achieved code capabilities for fast reactor conditions and paves the way towards the proper application of fuel performance codes to safety evaluations on Generation IV reactor concepts.

Review of Instant Release Fractions of Long-lived Radionuclides in CANDU and PWR Spent Nuclear Fuels Under the Geological Disposal Conditions

최희주,구양현,조동건 한국방사성폐기물학회 2022 방사성폐기물학회지 Vol.20 No.2

Several countries, including Korea, are considering the direct disposal of spent nuclear fuels. The radiological safety assessment results published after a geological repository closure indicate that the instant release is the main radiation source rather than the congruent release. Three Safety Case reports recently published were reviewed and the IRF values of seven long-lived radionuclides, including relevant experimental results, were compared. According to the literature review, the IRF values of both the CANDU and low burnup PWR spent fuel have been experimentally measured and used reasonably. In particular, the IRF values of volatile long-lived nuclides, such as 129I and 135Cs, were estimated from the FGR value. Because experimental leaching data regarding high burnup spent nuclear fuels are extremely scarce, a mathematical modelling approach proposed by Johnson and McGinnes was successfully applied to the domestic high burnup PWR spent nuclear fuel to derive the IRF values of iodine and cesium. The best estimate of the IRF was 5.5% at a discharge burnup of 55 GWd tHM−1.

HIGH BURNUP CHANGES IN UO₂ FUELS IRRADIATED UP TO 83 GWD/T IN M5® CLADDINGS

J. NOIROT,I. AUBRUN,L. DESGRANGES,K. HANIFI,J. LAMONTAGNE,B. PASQUET,C. VALOT,P. BLANPAIN,H. COGNON 한국원자력학회 2009 Nuclear Engineering and Technology Vol.41 No.2

Since the 90’s, EDF and AREVA-NP have irradiated, up to very high burnups, lead assemblies housing M5® cladded fuels. Post-irradiation examination of high burnup UO₂ pellets show an increase in the fission-gas release rate, an increase in fuel swelling, and formation of fission-gas bubbles throughout the pellets. Xenon abundances were quantified, and phenomena leading to this bubble formation were identified. All examinations provided valuable data on the complex state of the fuel during irradiation. They show the good behavior of these fuels, exhibiting various microstructures at very high burnups, none of which is likely to lead to problems during irradiation

HIGH BURNUP CHANGES IN UO₂ FUELS IRRADIATED UP TO 83 GWD/T IN M5^(R) CLADDINGS

Noirot, J.,Aubrun, I.,Desgranges, L.,Hanifi, K.,Lamontagne, J.,Pasquet, B.,Valot, C.,Blanpain, P.,Cognon, H. Korean Nuclear Society 2009 Nuclear Engineering and Technology Vol.41 No.2

Since the 90's, EDF and AREVA-NP have irradiated, up to very high burnups, lead assemblies housing $M5^{(R)}$ cladded fuels. Post-irradiation examination of high burnup $UO_2$ pellets show an increase in the fission-gas release rate, an increase in fuel swelling, and formation of fission-gas bubbles throughout the pellets. Xenon abundances were quantified, and phenomena leading to this bubble formation were identified. All examinations provided valuable data on the complex state of the fuel during irradiation. They show the good behavior of these fuels, exhibiting various microstructures at very high burnups, none of which is likely to lead to problems during irradiation.

A New Mechanistic and Engineering Fission Gas Release Model for a Uranium Dioxide Fuel

LEE, Chan Bock,YANG, Yong Sik,KIM, Dae Ho,KIM, Sun Ki,BANG, Je Geun Atomic Energy Society of Japan 2008 Journal of nuclear science and technology Vol.45 No.1

<P>A mechanistic and engineering fission gas release model (MEGA) for uranium dioxide (UO<SUB>2</SUB>) fuel was developed. It was based upon the diffusional release of fission gases from inside the grain to the grain boundary and the release of fission gases from the grain boundary to the external surface by the interconnection of the fission gas bubbles in the grain boundary. The capability of the MEGA model was validated by a comparison with the fission gas release data base and the sensitivity analyses of the parameters. It was found that the MEGA model correctly predicts the fission gas release in the broad range of fuel burnups up to 98 MWd/kgU. Especially, the enhancement of fission gas release in a high-burnup fuel, and the reduction of fission gas release at a high burnup by increasing the UO<SUB>2</SUB> grain size were found to be correctly predicted by the MEGA model without using any artificial factor.</P>

Towards grain-scale modelling of the release of radioactive fission gas from oxide fuel. Part I: SCIANTIX

Zullo G.,Pizzocri D.,Magni A.,Van Uffelen P.,Schubert A.,Luzzi L. 한국원자력학회 2022 Nuclear Engineering and Technology Vol.54 No.8

When assessing the radiological consequences of postulated accident scenarios, it is of primary interest to determine the amount of radioactive fission gas accumulated in the fuel rod free volume. The state-ofthe-art semi-empirical approach (ANS 5.4e2010) is reviewed and compared with a mechanistic approach to evaluate the release of radioactive fission gases. At the intra-granular level, the diffusiondecay equation is handled by a spectral diffusion algorithm. At the inter-granular level, a mechanistic description of the grain boundary is considered: bubble growth and coalescence are treated as interrelated phenomena, resulting in the grain-boundary venting as the onset for the release from the fuel pellets. The outcome is a kinetic description of the release of radioactive fission gases, of interest when assessing normal and off-normal conditions. We implement the model in SCIANTIX and reproduce the release of short-lived fission gases, during the CONTACT 1 experiments. The results show a satisfactory agreement with the measurement and with the state-of-the-art methodology, demonstrating the model soundness. A second work will follow, providing integral fuel rod analysis by coupling the code SCIANTIX with the thermo-mechanical code TRANSURANUS

Simulation of Pore Interlinkage in the Rim Region of High Burnup $UO_2$Fuel

Koo, Yang-Hyun,Oh, Je-Yong,Lee, Byung-Ho,Cheon, Jin-Sik,Joo, Hyung-Koo,Sohn, Dong-Seong Korean Nuclear Society 2003 Nuclear Engineering and Technology Vol.35 No.1

Threshold porosity above which fission gas release channels would be formed in the rim egion of high burnup UO$_2$ fuel was estimated by the Monte Carlo method and Hoshen-Kopelman algorithm. With the assumption that both rim pore and rim grain can be represented by cube, pore distribution in the rim was simulated 3-dimensionally by the Monte Carlo method according to porosity and pore size distribution. Then, using the Hoshen-Kopelman algorithm, the fraction of open rim pores interlinked to the outer surface of a fuel pellet was derived as a function of rim porosity. The simulation showed that porosity of 24-25% is the threshold above which the number of rim pores forming release channels increases very rapidly. On the other hand, channels would not be formed if the porosity is less than about 23.5%. This is consistent with the observation that, for porosity less than 23.5%, almost no fission gas is released in the rim. However, once the rim porosity reaches beyond 25%, extensive open paths would be developed and considerable fission gas release would start in the rim.

FUEL PERFORMANCE CODE COSMOS FOR ANALYSIS OF LWR UO2 AND MOX FUEL

BYUNG-HO LEE,구양현,JAE-YONG OH,Jin-SikCheon,YOUNG-WOOK TAHK,손동성 한국원자력학회 2011 Nuclear Engineering and Technology Vol.43 No.6

The paper briefs a fuel performance code, COSMOS, which can be utilized for an analysis of the thermal behavior and fission gas release of fuel, up to a high burnup. Of particular concern are the models for the fuel thermal conductivity, the fission gas release, and the cladding corrosion and creep in UO2 fuel. In addition, the code was developed so as to consider the inhomogeneity of MOX fuel, which requires restructuring the thermal conductivity and fission gas release models. These improvements enhanced COSMOS’s precision for predicting the in-pile behavior of MOX fuel. The COSMOS code also extends its applicability to the instrumented fuel test in a research reactor. The various in-pile test results were analyzed and compared with the code’s prediction. The database consists of the UO2 irradiation test up to an ultra-high burnup, power ramp test of MOX fuel, and instrumented MOX fuel test in a research reactor after base irradiation in a commercial reactor. The comparisons demonstrated that the COSMOS code predicted the in-pile behaviors well, such as the fuel temperature, rod internal pressure, fission gas release, and cladding properties of MOX and UO2 fuel. This sufficient accuracy reveals that the COSMOS can be utilized by both fuel vendors for fuel design, and license organizations for an understanding of fuel in-pile behaviors.