Simulation of an accident in conditions of cooling system malfunction and ventilation system failure in the spent fuel storage pool of the multipurpose fast research reactor

Simulation of an accident in conditions of cooling system malfunction and ventilation system failure in the spent fuel storage pool of the multipurpose fast research reactor* Anatoly A. Kazantsev1, Olga V. Supotnitskaya1, Evgeniya A. Ivanova1, Nataliya E. Astakhova1 1 IPPE JSC, 1 Bondarenko Sq., 249033 Obninsk, Kaluga reg., Russia Corresponding author: Anatoly A. Kazantsev ([email protected]) Academic editor: Georgy Tikhomirov ♦ Received 19 June 2025 ♦ Accepted 24 September 2025 ♦ Published 9 October 2025 Citation: Kazantsev AA, Supotnitskaya OV, Ivanova EA, Astakhova NE (2025) Simulation of an accident in conditions of cooling system malfunction and ventilation system failure in the spent fuel storage pool of the multipurpose fast research reactor. Nuclear Energy and Technology 11(3): 219–223. https://doi.org/10.3897/nucet.11.173061 Abstract The paper presents the results of the computational analysis of an accident scenario in the spent fuel storage pool of the multipurpose fast research reactor. The simultaneous failure of the spent fuel cooling system and the ventilation systems is being considered as beyond design basis accident (BDBA) conditions. The computational analysis of the BDBA scenario was performed with the computer program KUPOL-MT designed to simulate thermohydraulic processes in the room atmosphere of the reactor facility containment systems. During the computational analysis with KUPOLMT, a nodalization scheme for the spent fuel storage pool rooms was developed. The scheduled loading of spent fuel assemblies in the storage pool was considered in the accident simulation, together with the emergency unloading of the entire reactor core. It was shown that the water temperature in the spent fuel storage pool remained below the boiling point and there was no violation of the integrity of fuel cladding for three days after the start of the accident. As the fuel cladding did not collapse, there were no radiation consequences in the considered accident scenario. The calculation results demonstrated insignificant hydrogen stratification in the room above the spent fuel storage pool. It was shown that hydrogen content did not exceed the maximum concentration limit for three days after the start of the BDBA. Keywords Spent fuel storage pool, spent fuel assembly, Multipurpose Fast Research Reactor (MBIR), computer program KUPOLMT, beyond the design basis accident (BDBA), fuel element, hydrogen Introduction Nuclear research facilities (NRF) differ from nuclear power plants (NPP) in that they are intended not only to generate heat and electricity but also to be used for research activities. The Multipurpose Fast Research Reactor (MBIR) is designed to test materials and equipment for Generation IV reactors. The design thermal power of the MBIR is 150 MW (Technical characteristics of MBIR). The construction of the MBIR reactor at the site of NIIAR JSC in Dimitrovgrad, the Ulyanovsk Region, is expected to ensure the continuity of the experimental and research Copyright Kazantsev AA et al. This is an open access article distributed under the terms of the Creative Commons Attribution License (CC-BY 4.0), which permits unrestricted use, distribution, and reproduction in any medium, provided the original author and source are credited. Nuclear Energy and Technology 11(3): 219–223 DOI 10.3897/nucet.11.173061 Research Article * Russian text published: Izvestiya vuzov. Yadernaya Energetika (ISSN 0204-3327), 2025, n. 2, pp. 142–152. Kazantsev AA et al.: Simulation of an accident in conditions of cooling system malfunction220 programs implemented at the BOR-60 research reactor now in the process of decommissioning. The spent fuel storage pool (SFSP) is designed for safe long-term storage and cooling of spent fuel. The SFSP has an important role in ensuring the reactor safety at the normal operation (NO), refueling and further long-term fuel storage stages. The regulatory documents for the NRF SFSPs are NP-061-05 (NP-06105, 2005) and NP-049-17 (NP-049-17, 2017). The paper investigates the consequences of a beyond design basis accident (BDBA), in which a heat removal failure in the process of the SNF storage and transportation is considered an initiating event (as shown in the Appendix to NP-061-05 (NP-061-05, 2005). Nonoperation of the ventilation system is considered a failure. Fuel assembly and spent fuel storage pool design The MBIR SFSP comprises three storage compartments in the form of concrete boxes with a stainless steel lining. Two storage compartments are designed for scheduled FA storage. The two storage compartments provide redundancy in case one of them fails. The third compartment is designed for emergency unloading of the full reactor core. Fig. 1 presents the layout of the MBIR SFSP rooms. Each compartment is sectionalized for the FA accommodation. Reliable protection of personnel against ionizing radiation is provided by the water level above the FAs of about 6.5 m. And the water level from the compartment bottom is 9.1 m. The MBIR NRF is intended to replace the BOR-60 research reactor and has the same fuel element design with a diameter of 6 mm and the cladding thickness of 0.3 mm (Zhemkov I.Yu, 2014). Prior to the loading, spent fuel is cooled in the in-core storage (on the core periphery) for two cycles of 100 days (Technical characteristics of MBIR) and 30 days of refueling. Pre-unloading 230-day SFA cooling in the in-core storage leads to a substantial decrease in dose loads in the spent fuel refueling and storage process and to a decrease in the SFSP water temperature. To remove residual heat, the SFSP has a cooling system which comprises heat exchangers, pumps, pipelines, and valves. The volume of the room above the SFSP is 1507 m3. The volumes of the two SFA storage compartments are 384.5 m3, and the volume of the third compartment is 370 m3. In the slotted floor slab of the SFSP compartments there are hatches which are opened prior to the FA transportation from the rooms above the SFSP. The concrete wall lining and the slotted floor slabs are made of steel. In the process of normal operation (NO), the ventilation system removes air from the SFSP space above the water (under the slotted slab) and from the SFSP gas space (the rooms above the SFSP). For the NO mode, the following air parameters are maintained in the room above the SFSP: a pressure of 1 atm, a temperature of +8 to +30 °C, and a relative air humidity of up to 75%. The room above the SFSP is connected with the room for the SFSP inclined elevator (IH). The inclined elevator is used to transport FAs. The air parameters in the NO mode in the IH room are the same as in the room above SFSP. The SFSP rooms are provided with the main and the standby ventilation systems. Accident with the SFSP cooling failure A beyond-design-basis accident is considered involving the SNF cooling failure in the SFSP compartments which leads to the following MBIR systems becoming inoperative: • ventilation system (SFSP compartments, room above the SFSP, IH room); • standby ventilation systems (SFSP compartments, room above the SFSP, IH room). Figure 1. Layout of the MBIR SFSP rooms. Nuclear Energy and Technology 11(3): 219–223 221 No ventilation air ducts are shut off in the event the ventilation system fails. To estimate the maximum hazardous consequences of an accident, it is assumed that operating personnel to fail to take measures to manage the accident for 72 hours (three days) as defined in NP-001-15 (NP-001-15, 2016) and NP-033-11 (NP-033-11). In the event of such accident, namely, when the SFSP’s main cooling system fails, residual heat is removed from the SFSP surface. The water in the SFSP compartments starts to heat up, and the water level in the SFSP compartments decreases due to evaporation. Over time, the water temperatures in the SFSP and in the gas space above the SFSP surface equalize. In normal operating conditions of the SFSP, hydrogen is formed continuously in the SFSP compartments as a result of the water radiolysis due to photon absorption from SFAs, SFSPS assemblies and side screen assemblies stored in the cooling pool water. Radiolytic hydrogen is released into the gas space above the SFSP surface. In the event of an accident, due to the continuous release of hydrogen and no forced ventilation, the concentration of hydrogen in the SFSP gas space starts to increase. Computational accident simulation The combination of the processes taking place in the MBIR SFSP in the course of the accident under consideration involving the SFSP cooling failure is described by a set of models implemented in the KUPOL-MT containment code, version 1.0 (the KUPOL-MT hereinafter) (Kazantsev et al. 2019). The KUPOL-MT is designed to calculate the parameters within the steam/air containment atmosphere at the NPP with VVER type reactor. The code calculates the following values (Kazantsev et al. 2019): • pressure history in each room and pressure differences history between rooms; • temperature history in all rooms; • transient temperature distribution in structures and equipment installed in each room; • component concentration history of in each room; • water pool level history in rooms. The calculation takes into account the effects of transient heat and mass transport of the gas-droplet mixture in the containment rooms, taking into account the effects from natural convection, volume and surface condensation of steam in the presence of noncondensable gases, and heat and mass exchange of the room atmosphere with water in the sumps. The KUPOL-MT code uses a lumped parameter model. The design data of the SFSP compartments, the rooms above the SFSP and the inclined elevator room, which is connected with the space of the room above the SFSP through the ventilation system, are used as input data for the MBIR SFSP computational model construction using the KUPOL-MT code. The computational model used for the SFSP BDBA simulation in the KUPOL-MT code is presented in Fig. 2. The computational model comprises ten internal design volumes interconnected via twenty two junctions. Computational volume 1 simulates the inclined elevator room. Computational volumes 2, 3 and 4 simulate the SFSP storage compartments 1, 2 and 3. The room above the SFSP is broken down into two axial levels and comprises computational volumes 5 through 10. Fuel assemblies in the SFSP compartments are simulated by cylindrical multilayer heat structures (walls). The floor and walls of the SFSP compartments are also simulated by flat heat structures. The FA afterheat generation is assumed to be similar for compartments 1 and 2 and amount to 20.7 kW based on the scheduled SFA unloading into the SFSP after ten-year refueling mode. The afterheat generation in compartment 3 is taken as equal to 155 kW and corresponds to 30-day SFA cooling. The rates of radiolytic hydrogen generation in each SFSP compartment are calculated based on the radiochemical yield of hydrogen depending on the FA total thermal power in the compartment. The total rate of hydrogen generation is 1.87·10-5 kg/s. The temperature of the hydrogen released from the SFSP water is taken as equal to the water temperature in each SFSP compartment. The initial water temperature is 40 °C. From the space above the SFSP water and through breaks in the slotted floor slab above the SFA receptacles, the gas mixture enters the space in the room above the SFSP. The extent to which the slotted floor slab is leaktight, determined as the ratio of the slotted floor slab flow area to the total slotted floor slab area above the SFSP compartments, is estimated as 10%. Results of computational simulation using the KUPOL-MT code The potential risk of radioactivity release from the cooling pool leads to the need to maintain the SFSP atmosphere parameters. In the event of an accident with the SNF cooling failure in the SFSP water and the SFSP room ventilation failure, the parameters for monitoring in the process of computational simulation are Figure 2. Computational model of the SFSP rooms for the KUPOL-MT code. Kazantsev AA et al.: Simulation of an accident in conditions of cooling system malfunction222 • fuel cladding temperature in the SFSP compartments; • water temperature in the SFSP compartments; • water level in the SFSP compartments; • hydrogen concentration in the space above water in the SFSP compartments and in the SFSP room. Figs 3–6 show the results of the BDBA computational simulation using the KUPOL-MT code (Kazantsev et al. 2019). The calculation time is 300000 s (~ 3 days). The water temperature in the SFSP compartments is shown in Fig. 3. The water temperature in the SFSP compartments 1 and 2 reaches 49 and 51 °С respectively. Compartment 3 with emergency unloading has the water temperature reaching 80.6 °C. There is no water boiling in the SFSP compartments. The fuel cladding temperature in the SFSP compartments 1, 2 and 3 is shown in Fig. 4. The fuel cladding temperature in the SFSP compartment 3 reaches a maximum of 85 °С, and the fuel cladding temperatures reach 52 °С (compartment 1) and 49.4 °С (compartment 2) (Fig. 4), respectively. Fig. 5 shows the atmospheric temperature in the three SFSP compartments. The atmospheric temperature in compartment 3 reaches 77 °С (see Fig. 5). The atmospheric temperature in compartments 1 and 2 (boxes 2 and 3) approaches 82 and 72 °С respectively. The hydrogen concentration in the SFSP compartment gas space is shown in Fig. 6. Calculations show that the hydrogen concentration in the SFSP compartments does not exceed 2 vol.%. The maximum hydrogen concentrations in the SFSP compartments reach 1.22 vol.% in compartment 1, 1.18 vol.% in compartment 2, and 1.03 vol.% in compartment 3 (see Fig. 6). After it reaches the maximum values, the hydrogen concentration drops due to dilution with saturated steam, the partial pressure of which on the saturation line increases rapidly as the temperature grows. The maximum hydrogen concentration values are less than 4.1 vol.%, the smallest concentration required for the hydrogen ignition in a mixture with air (Gamburg et al. 1989; Gelfand et al. 2008). The hydrogen concentration change in the room space above the SFSP is shown in Fig. 6. The arrangement of computational volumes is shown in Fig. 2. It can be seen that there is no noticeable stratification of hydrogen in the room above the SFSP. The maximum concentration of hydrogen in all computational volumes of the room above the SFSP does not exceed 1.0 vol.% (see Fig. 6). The hydrogen concentration starts to fall then due to being diluted with steam as the partial steam pressure grows with the temperature grow in the room above the SFSP and decreases to 0.6 vol.%. The hydrogen concentration in the SFSP compartment is some 0.2 vol.% higher than in the room above the SFSP. Conclusions The paper presents the results of the computational simulation using the KUPOL-MT code for an accident involving an irradiated fuel cooling failure in the spent fuel storage pool accompanied by an additional failure in the Figure 3. Water temperature in three SFSP compartments. Figure 4. Fuel cladding temperature in the SFSP compartments. Figure 5. Temperature in the SFSP compartment gas space. Figure 6. Hydrogen concentration in the SFSP compartments. Nuclear Energy and Technology 11(3): 219–223 223 MBIR reactor. The additional failures assumed include failures of the main and standby ventilation systems in the SFSP compartments, and a ventilation failure in the room above the SFSP and in the SFSP inclined elevator room. 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