A new circuit solution to minimize the consequences of the failure of the feed water cool down line of the BREST-OD-300

A new circuit solution to minimize the consequences of the failure of the feed water cool down line of the BREST-OD-300* Aleksey V. Popov1, Egor N. Kulikov1, Daria R. Danilova1, Alexander S. Lisyanskii1, Fyodor A. Svyatkin1, Sviatoslav O. Kiriyenko2, Egor V. Bubenshikov2, Irina V. Pavlova2, Victor V. Khodakovsky2, Valery N. Chigarev2 1 JSCPolzunovScientificandProductionAssociationfortheResearchandDesignofPowerEquipment,3/6AtamanskayaSt.,191167St. Peterburg,Russia 2 ProryvJSC,1RamenskiyBoulevard,119607Moscow,Russia Corresponding author: Aleksey V. Popov (PopovA[email protected]) Academic editor: Yury Kazansky ♦ Received 30 January 2025 ♦ Accepted 30 June 2025 ♦ Published 7 November 2025 Citation: Popov AV, Kulikov EN, Danilova DR, Lisyanskii AS, Svyatkin FA, Kiriyenko SO, Bubenshikov EV, Pavlova IV, Khodakovsky VV, Chigarev VN (2025) A new circuit solution to minimize the consequences of the failure of the feed water cool down line of the BREST-OD-300. Nuclear Energy and Technology 11(4): 243–249. https://doi.org/10.3897/nucet.11.176904 Abstract During the design of the secondary circuit for a nuclear power plant with the BREST-OD-300 reactor, engineers and designers faced specific challenges related to the properties of the lead coolant. The key issue was its high crystallization temperature, which necessitated fundamentally new approaches to the thermal circuit design of the power unit. Particular attention was paid to developing the feedwater heating system, as its parameters and configuration directly affect the thermal balance and operational stability of the reactor plant. Throughout the design process, the configuration of this system underwent multiple revisions, reflecting the complexity of achieving optimal technical solutions while meeting stringent reliability and safety requirements. Each modification to the feedwater heating system design requires comprehensive analysis of its impact on all key components of the power unit, including both the reactor and turbine circuits. Engineers employed an Integrated Сalculating Mathematical Model (ICMM) to evaluate the system performance in normal and emergency operating modes, incorporating interconnections between all technological processes. This approach not only helps identify potential issues at early design stages but also optimizes equipment operating parameters. This approach achieves significant improvements in the power unit’s economic efficiency while simultaneously meeting all reliability and safety requirements. The research placed special emphasis on analyzing accident scenarios, particularly the feedwater cooldown line failure. Engineers applied mathematical modeling to pinpoint critical parameters and design failure mitigation solutions. These included upgrades to backup heating systems and optimization of control algorithms. Thus, this work highlights the necessity of employing modern modeling methods when developing innovative nuclear reactors. Such approaches enable advance development of technical solutions to ensure safe and efficient operation of these advanced power units. Keywords NPP, BREST-OD-300, Integrated Сalculating Mathematical Model, Liquid Metal Coolant, Lead, Direct Contact Feed Water Heater (DCFH), HTFP (Hydroturbine Feed Pump) Copyright Popov AV 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(4): 243–249 DOI 10.3897/nucet.11.176904 Research Article * Russian text published: Izvestiya vuzov. Yadernaya Energetika (ISSN 0204-3327), 2025, n. 2, pp. 153–166. Popov AV et al.: Failure of the feed water cool down line of the BREST-OD-300244 Introduction Nuclear power plants are extra hazardous nuclear facilities which places emphasis on reliability and safety issues. Special attention is given therefore to designs using novel engineering solutions and, specifically, to molten lead cooled plants. To prevent crystallization of lead, the feedwater circuit includes a high-temperature circuit with nonstandard equipment, the parameters of which were updated in the process of engineering studies. Using mathematical models of complex technical systems is an effective tool of engineering design (Parshikov et al. 2016). They make it possible to find hidden links among systems in a variety of scenarios, including emergencies, at the design stage. The Integrated Сalculating Mathematical Model (ICMM) enables dynamic simulation of the power unit operation in nominal, transient and emergency modes. This allows one to identify inconsistencies in process parameters, evaluate operation of interconnected components, and check the input design data for being complete and consistent. Specific features of the lead cooled reactor feedwater unit Large-scale evolution of nuclear power, along with limited reserves of natural uranium about 1% of which is used in thermal reactors, is expected to require a closed nuclear fuel cycle and an expanded fuel base to be achieved via using liquid metal cooled fast breeder reactors (Adamov et al. 2012). Fast reactors not only make it possible to use in full all natural uranium, but allow as well afterburning of long-lived spent fuel components and disposal of weapon-grade plutonium. Economically, an advantage is a higher net efficiency amounting to 39% against 35% in the event of VVER-type NPP units (Glebov 2022). Currently in operation are two units with fast neutron reactors (BN-600 and BN-800) (Country Statistics: Russian Federation. IAEA: Power Reactor Information System). The rest were closed or decommissioned due to a decreased share of nuclear power in global generation following a number of major accidents. Sodium was used largely as coolant in such designs due to its good thermophysical properties, low cost, compatibility with structural materials and ease of maintaining the chemical composition in the course of operation. A competitive technology is liquid lead cooled reactors; both designs are a part of the Generation IV program (Locatelly et al. 2013). Using lead allows avoiding the following concerns associated with the use of sodium coolant: • danger of the coolant to react with air or water in the event of a circuit break; • need for an intermediate circuit. Use of lead coolant is a novel field involving much experimental and computational work. The key research activities, as defined in the IAEA roadmap (Gen IV International Forum 2020), include studies on fuel (Adamov et al. 2022), steam generator (SG) (Grabezhnaya et al. 2013) and pump (Beznosov et al. 2017) structures, structural materials (Ageev et al. 2009, thermal hydraulics (Belozerov et al. 2016) and reactor plant analysis codes (Bol’shov et al. 2016). As coolant, lead features a high melting point, which results in an energy-specific requirement placed on the feedwater temperature downstream of the SG to exclude the potential for the lead coolant solidification. This is a novel requirement for power units leading to the need for non-routine engineering solutions. The BN-350 design, developed to demonstrate the sodium coolant technology (tmelt = 97 °C), used a deaerator only for heat exchange in the feedwater line, had a feedwater temperature of 158 °C, and was started in the same way as thermal plant units. For later sodium cooled NPPs, the feedwater temperature was determined based on engineering and economic optimization (Fig. 1a). Similar solutions are applicable to lead-bismuth alloy (tmelt = 123 °C). The feedwater heating system of a lead cooled unit requires the feedwater temperature to be ~ 10 to 15 °C in excess of the melting point in all modes of operation (Fig. 1c), unlike the latest commissioned BN-800 unit (Fig. 1b) which has a margin of ~ 100 °C. Initially, standard solutions were proposed to be used in the lead cooled reactor unit design, the way it was done with the sodium coolant technology. However, use of lead, despite all its advantages as coolant, has major effect on the secondary circuit structure when water is used as the fluid in the steam-power cycle (Kondurov et al. 2024). Thermal circuit of the BREST-OD-300 In Russia, a pilot demonstration unit with a lead cooled fast neutron reactor continues to be built at the site of the Siberian Chemical Plant. It was decided that feedwater would be heated up to the required temperature in a vertical mixing feedwater heater (DCFH) upstream of the high-pressure heaters (HPH). A mixing heater used in the system requires a second-lift feed pump to deliver feedwater from the DCFH to the SG. For the initial circuit version (Fig. 2a) (Filin et al. 2001), developed by VTI in the early 2000s, a DCFH design with a pressure jet distribution was proposed, in which a steam-water mixture was fed laterally, and feedwater upstream of the HPHs was delivered to the steam space through a system of holes as jets (Somova et al. 2009). Feedwater in the DCFH steam space is heated as a result of the heating medium condensation on the jets. It was proposed that a steam-water mixture with parameters t = 352 °C and a steam quality of x = 0.8 be used as the heating medium. Further studies have shown that superheated steam may be used for heating, the temperature of which was selected based on the DCFH strength condition and amounts to 400 °C. For reliable cavitation-free operation of the second-lift feedwater pump (PEN-2), the system includes a continuously Nuclear Energy and Technology 11(4): 243–249 245 operating cooldown line with the fluid withdrawal downstream of the DCFH level control valve (3–4 of the flow rate) into the branch pipe of the water discharge pipeline, including in a mode with the accidental DCFH heating steam valve closure. The pressure drop for the supply of water for being cooled down was achieved by the resistance in the feedwater supply device (perforated pipe) in the DCFH. There was a control valve with a limiting washer installed in the cooldown line. The valve was intended only to adjust the flow rate in transient modes and to be activated only in the event the DCFH load and the feedwater temperature downstream of the HPHs change greatly. At the same time, even with the 30% loaded unit, erroneous opening of the valve with the HPHs being off did not lead to the feedwater temperature upstream of the SG decreasing to below 340 °С. Additionally, the cooldown valve could be used in the process of start-ups in the DCFH makeup modes with a low full-flow condensate rate. To this end, a line was installed to connect it to the pipeline for the DCFH nozzle group. This approach was continued until 2012 (Nesterov et al. 2011). However, it had a number of drawbacks: • lack of methods to calculate a design with a pressure jet distribution with sufficient accuracy; • the need to build a costly full-scale test bench with a pressure water distribution to confirm the equipment performance. Due to the above factors, an DCFH design with a non-pressurized jet two-stage distribution on perforated trays (Fig. 2b) was proposed by analogy with the deaerator columns (Trofimov 2002) to increase the jet length and, accordingly, the heat exchange area. Bench tests were undertaken at JSC “NPO CKTI” using DCFH models which confirm that the required heating is achieved in the process of contact heat exchange (Sukhorukov et al. 2021). Due to process complications in fabrication of PEN-2s with high water inlet Figure 1. Temperatures at representative points of the LMC unit feedwater system at rated power (a) and at partial power of the BN800 (b) and BREST-OD-300 (c) reactors: 1 – feedwater temperature downstream of HPHs; 2 – feedwater upstream of SG, 3 – LMC temperature downstream of SG, 4 – live steam temperature, 5 – LMC melting point; 6 – FW temperature margin to LMC freezing temperature; 7 – temperature of FW subcooling downstream of the HPHs to the LMC freezing temperature. Popov AV et al.: Failure of the feed water cool down line of the BREST-OD-300246 parameters (p = 15 MPa; t = 340 °C) and respective head and delivery rate values, a decision was made that water would be delivered from the DCFH using a hydroturbine feed pump (HTFP) (Fig. 3b), the design of which was optimized based on the NPP steam superheater condensate injection pumps (Kulakov et al. 2023; Popov et al. 2024). This solution makes it possible to design leak-tight pump sets using bearings operating with the hydraulic drive fluid, which ensures reliability of high-parameter water pumping (Shlemenzon et al. 2014). The design of the NPP’s reactor coolant pumps (RCP) and the fluid recirculation pumps of the thermal power plant supercritical pressure units (vertical single-stage apparatuses), which have similar parameters, cannot be used for the second-lift feed pump in the BREST-OD-300 reactor unit for the following reasons: • in the event of combined delivery and head values, a multi-stage pump shall be used with an impeller speed of 3000 rpm; • a vertical design of a multi-stage pump at a pumped fluid temperature of 340 °С is extremely difficult to develop; • the end seal with the specified parameters, taking into account the experience of designing reactor coolant pump seals, has an axial dimension of about 1 m, which requires a much longer distance between the pump supports. Figure 2. Design of the high-pressure feedwater mixing heater: a. Developed by VTI; b. Developed by CKTI; 1 – water supply device; 2 – protective shell; 3 – heating medium (steam) supply; 4 – feedwater inlet; 5 – vented steam; 6 – heated feedwater outlet; 7 – upper and lower trays; 8 – water feed for cooldown; 9 – splitter. Figure 3. DCFH actuation diagram: a. VTI-developed DCFH; b. СKTI-developed DCFH; 1 – DCFH; 2 – PEN; 3 – steam generator; 4 – heating medium (steam); 5 – feedwater from HPHs; 6 – steam to turbine; 7 – HTFP. Nuclear Energy and Technology 11(4): 243–249 247 Switching to free-flow water distribution in the DCFH, along with using a fundamentally new pump, led to the cooldown line tie-in installed upstream of the DCFH-level control valve, which has substantial effect on modes with anticipated operational occurrences, specifically with partial loads. Cool down line and ensuring its reliable operation The throughput of the cooldown control valve, Kv, is selected for the flow rate, G0, and the nominal pressure drop, Δp0. The tie-in installation upstream of the DCFH-level valve leads to the pressure drop in the valve growing noticeably in conditions of partial loads due to the uncontrolled PEN-1 drive, which is also the case with the feedwater density due to a decrease in its temperature downstream of the system. For example, at a reactor power of 30%, when the feedwater temperature downstream of the regeneration system is as small as possible, the flow rate through the fully opened valve will be approximately twice as large as in the event of the rated mode, which is determined as follows: ρ ρ (1) where G is the fluid mass flow rate through the valve, t/h; p is the fluid density, kg/m3; and Δp is the valve pressure drop, bar. With the minimum reactor power of 2%, switching on SG 1, where the cooldown valve opens fully due to the automatic equipment or operator error, leads to the flow rate through the cooldown line, which is the DCFH bypass, growing potentially tenfold, this causing a major decrease in the SG inlet feedwater temperature. According to current algorithms of control, there are no direct feedwater temperature safeguards since water in the saturation line was supplied earlier from the DCFH, and the cooldown line had no major effect in all unit operating modes. A feedwater temperature decrease upstream of the SG does not cause safeguards and interlocks to operate, since the feedwater control valves respond to the feedwater temperature decrease by reducing the feedwater flow rate, while maintaining the required lead coolant temperature downstream of the SG. To prevent a feedwater temperature decrease, a circuit design was proposed, which includes an installed cooldown limiter in the form of a safeguard CV (Fig. 4). Its capacity has been selected for the full flow rate and 30% of the total drop in the line. The safeguard valve regulator maintains the temperature downstream of the mixer at a level 4 °C smaller than the main regulator set point, and it is opened during normal operation. Erroneous regulator opening was simulated and the proposed design was tested at the BREST-OD-300 unit ICMM (Fedorovsky et al. 2024) deployed at Proryv JSC’s mathematical simulation site. The calculation was undertaken for two layout options and two initial states: the initial layout and the proposed layout at the minimum reactor power of 2% with water supplied to the SG and at the minimum power level of 30% of the reactor power. The ICMM initiating event is unauthorized opening of the cooldown valve. There is a noticeable decrease observed in the circuit temperature in the event of the option with no safeguard control valve (Fig. 5, lines 2 and 4). In low power modes (2 to 10%), the cooldown flow rate exceeds the SG flow rate which leads additionally to an increase in the DCFH level and the further protected shutdown of the feedwater pumps. The feedwater temperature upstream of the SG is 315 °С. In the 30% mode, the temperature upstream of the SG was 330 °С (Fig. 5, line 2). The lead coolant temperature did not practically change due to the feedwater flow rate reduction by the feedwater valve regulator. With a safeguard valve used in the layout, the feedwater temperature decreases locally after which the valve operates to cope with the disturbance caused by the main CV opening and maintains the temperature at 337 °C. Conclusion When used as reactor coolant, lead places a specific requirement on the secondary circuit which consists in that the feedwater temperature upstream of the SG should be maintained in all unit operation modes. Redesign of the feedwater heating system requires that reliability of operation be checked, including in modes with anticipated operational occurrences. The circuit designs with and without a safeguard valve installed were validated, as the project proposed, at the BREST-OD-300 unit ICMM. In the layout option with no redesigns, the cooldown layout opening at low reactor power levels (2% to 10%) leads to an increase in the DCFH level and subsequent shutdown of the feedwater pumps. In power modes, the cooldown valve opening causes the feedwater temperature upstream of the SG to decrease to 300 °C but with no primary circuit protections, as currently designed, starting to operate. In the event of the circuitry with the second safeguard valve installed, only a short-term feedwater temperature decrease is observed in all modes that does not exceed the permissible value. Figure 4. Proposed cooldown line layout: 1 – DCFH; 2 – cooldown valves; 3 – cooldown limiter; 4 – mixer; 5 – feedwater from HPHs; 6 – heating medium (steam); 7 – to HTFP Popov AV et al.: Failure of the feed water cool down line of the BREST-OD-300248 References Adamov EO, Dzhalavyan AV, Lopatkin AV, Molokanov NA, Muravyov EV, Orlov VV, Kal’akin SG, Rachkov VI, Troyanov VM, Avrorin EN, Ivanov VB, Aleksakhin RM (2012) Conceptual framework of a strategy for the development of nuclear power in Russia to 2100. 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