Design Basis Conditions Research Articles

Advanced Analysis of Structural Performance in Novel Steel-Plate Concrete Containment Structures

This paper investigates the structural performance of novel steel-plate concrete containment structures, focusing on third-generation nuclear power plants. To address the challenges of increased complexities and costs associated with double-layer containment designs, this study explores the potential of steel-plate concrete structures to enhance safety, economic efficiency, and construction simplicity. The steel-plate concrete structure, characterized by its core concrete and dual steel plates, shows superior compressive strength, bending resistance, and elastoplasticity. Extensive numerical analyses, including finite element modeling and thermal-stress coupling, were conducted under various load conditions. Under structural integrity test conditions, the maximum radial displacement observed was 24.59 mm. Under design basis conditions, the maximum radial displacement was 47.61 mm; under severe accident conditions, it was 53.83 mm. The ultimate bearing capacity was 0.91 MPa, 2.17 times the design pressure. This study concludes that the steel-plate concrete containment structure maintains a high safety margin under all tested conditions, with stress and strain well within acceptable limits. It can effectively serve as a robust barrier against radioactive leakage and malicious impacts, providing a viable alternative to conventional containment designs.

Design standardisation and seismic protection of SMRs through modular metafoundations

This paper investigates the seismic protection of the Nuward™ small modular reactor (SMR) building, focusing on design loading and beyond design basis earthquake (bDBE) conditions. The study aims to achieve two primary objectives: i) to enhance seismic mitigation of a SMR building under bDBE conditions, through the use of the innovative modular single-layer (SLM) and multi-layer (MLM) metafoundations (MFs); ii) to effectively standardise and harmonise SMR building designs in locations prone to beyond design basis conditions. To accomplish these goals and demonstrate the protective capabilities of the MFs, the study employs non-linear time-history analyses (NLTHAs) for both DBE and bDBE conditions. Along these lines, a reduced-order model was developed from a refined finite element (FE) model of the SMR building using the Craig-Bampton mode synthesis technique. Then, finite locally resonant modular MFs were designed and analysed using NLTHAs. Specifically, physics-based ground motion models (GMMs) were used to generate and select seismic triplets that mimicked DBE and bDBE scenarios for NLTHAs. Successively to achieve improved seismic performance, the optimization of the MFs was pursued by targeting the optimal number of columns, resonator parameters, and unit cell dimensions. Additionally, the deployment of inerters was considered, to significantly reduce the size of the MFs and enable their application in multiple layers for ultra-low frequency attenuation. The overall findings suggest that modular MFs meet seismic protection requirements, and positively contribute to the standardization process of SMR buildings, even in areas characterized by beyond-design seismic conditions.

Research on two novel hydrogen-electricity-heat polygeneration systems using very-high-temperature gas-cooled reactor and hybrid-sulfur cycle

The combination of the hybrid-sulfur (HyS) cycle with a very-high-temperature gas-cooled reactor (VHTR) presents a promising nuclear hydrogen production technology. This study proposes two novel hydrogen-electricity-heat polygeneration systems using VHTR and HyS cycle. A thermodynamic model is developed via energy and exergy analyses. Thermodynamic performance analysis and comparison are conducted on the two systems. For the two systems under design basis condition, the recommended variation range of the split ratio is 0–0.6395 and the hydrogen production rate is 0–55.7 mol/s. For systems 1 and 2, the energy utilization efficiencies are 65.91 % and 64.53 %, and the exergy efficiencies are 68.55 % and 67.36 %, at a hydrogen production rate of 55.7 mol/s. The process heat supply unit, steam generator (SG) and HyS cycle exhibit the top three largest exergy loss coefficients of 6.32 %, 6.26 % and 4.21 % at the split ratio of 0.5; therefore, they are key units for improving system performance. The intermediate heat exchanger's primary-side helium outlet temperature and SG's secondary-side main steam temperature both affect the variation range of split ratio. A slightly lower main steam temperature can increase the maximum hydrogen production rate and maximum energy utilization efficiency.

Total Loss of Feedwater Analysis of Pressurized Water Reactor Using RELAP5

Abstract In Europe, the design extension conditions (DEC) were introduced after the Fukushima Dai-ichi accident as preferred method for giving due consideration to the complex sequences and severe accidents without including them in the design basis conditions. The objective of the study is to determine available elapsed time before core uncovery and needed DEC safety features for total loss of all feedwater (TLOFW) in a two-loop pressurized water reactor. RELAP5/MOD3.3 computer code has been used for calculations. The initiating event for TLOFW is multiple failures in which, in addition to the loss of main feedwater, the auxiliary feedwater is also lost. The scenarios without DEC safety features and the scenarios with DEC safety feature assumed have been simulated. The results showed that after TLOFW event initiation, it is very important to trip the reactor as soon as possible. In case of loss of offsite power, the reactor coolant pumps stop, and the reactor very quickly trips on low reactor coolant pump flow. When normal operation systems are assumed, the reactor trip occurs on low-low steam generator narrow level few tens of seconds after accident initiation, resulting in less time available before core uncovery occurence. The results for TLOFW scenarios with normal operation systems and DEC safety features assumed demonstrated that secondary side bleed and feed can prevent core uncovery in case when no operator actions are credited in the first 30 min of event. When primary side bleed and feed is used, less time is available for operator actions.

Safety considerations related to boron redistribution during Loss-of-Coolant-Accident initiated beyond design basis events for the NuScale power module

The authors, as staff at the U.S. Nuclear Regulatory Commission (NRC), evaluated the potential consequences of boron dilution under beyond design basis conditions in the NuScale Power Module to inform NRC staff’s safety evaluation report for the NuScale design certification review. The authors identified the small break loss of coolant accident with coincident failure of control rods to insert as the scenario with the most adverse boron dilution and reactivity insertion consequences. The authors evaluated the sequence of events for the scenario to determine which conditions or phenomena could produce the potential for a core flow surge to yield core damage because of a super-prompt reactivity excursion. The authors screened the conditions and phenomena that could potentially cause a core flow surge based on the dynamic speed of the incursion compared to key reactivity feedback mechanisms: thermal–hydraulic coupling via rod heat transfer and convective mixing. Based on the screening process, the authors were able to identify two processes that could potentially lead to a power excursion. However, the authors’ analysis relies heavily on assumptions regarding convective mixing, making this an important phenomenon worth careful consideration in any evaluation of boron redistribution.

Oxidation behavior of CrAl-Mo coated Zircaloy-4 in DB and BDB scenarios

Cr coatings have been widely applied on Zircaloy for accident tolerant fuel materials. However, it has long been a great challenge in the nuclear community for Cr coatings surviving in the beyond design basis conditions (BDB) because of the unstable chromia layer and eutectic reactions between Cr and Zr. In this study, bi-layered CrAl-Mo coatings have been designed to improve the performance of Cr coatings in the BDB conditions. Results showed that the CrAl-Mo coated samples can tolerant in steam at temperatures as high as1400 °C because of the formation of a more stable alumina layer on the surface and the inhibition of eutectic reactions near the coating/substrate interface. Moreover, diffusion of Mo facilitated the formation of a CrAlMo layer, which stabilized the outer alumina layer and reserved the Al inventory within the coating.

Safety analysis of representative accidental transients for a 100 MWe small modular LBE-cooled reactor

Nuclear energy plays an important role in the green low-carbon energy. The lead-based reactor is one of the most popular small modular reactor types. The pre-conceptual design of a 100 MW(e) small modular lead-bismuth eutectic cooled reactor was developed by Shanghai Nuclear Engineering Research & Design Institute CO., LTD. to meet the market demand of advanced small-scale nuclear plant. Representative accidents were analyzed for the 100 MW(e) small modular lead-bismuth reactor using ATHLET/MOD 3. The result shows that under the design basis conditions, the performance of the reactor can meet the safety criteria with the protection system working normally. The coolant solidification in the primary circuit under these transients should be prevented. Under the design extension condition, the highest cladding temperature is much lower than the melting point. In the unprotected reactivity insertion condition, there may be a short time of local melting of the fuel, which also meet the acceptance criteria. The safety analysis demonstrated the passive safety of the design with the negative temperature coefficient, strong natural circulation capability, and the passive residual heat removal system.

Real-time modeling of transient behavior of the primary coolant system in pool-type sodium cooled fast reactor

In the paper, a real-time simulation code THCSA-RT is developed to analyze the primary coolant system of pool-type sodium cooled fast reactor which includes reactor core, hot and cold sodium pools, pumps, and intermediate heat exchangers. Different numerical methods including Gear multi-value method and Quasi-Newton iteration method are used to solve the mass and energy equations set and the momentum equations set, respectively. Several typical design basis conditions of CEFR such as planned shutdown condition and RIA are analyzed, and the calculation results agree well with those of the final safety analysis report calculated by DINROS. The simulation time is presented to prove the effectiveness of the new numerical methods developed in the present work, and the new numerical methods could be applied in the similar system analysis codes as well. The newly developed code THCSA-RT could be used in the system real-time simulation of pool-type sodium cooled fast reactor and has been adopted in the development of CEFR full scope simulator.

Study on Typical Design Basis Conditions of HPR1000 With Nuclear Safety Analysis Code ATHLET

The third-generation nuclear power plant Hua-long Pressurized Reactor (HPR1000) is developed based on the experience of Chinese commercial Nuclear Power Plant (NPP) designs, construction, operation and maintenance. It improves the concept of defense in depth and strengthens severe accident prevention and mitigation strategies. The HPR1000 has implemented a number of active and passive innovative safety systems and accident management procedures for design basis conditions, e.g., the employment of Medium Pressure Rapid Cooldown (MCD) and Atmospheric Steam Dump System (ASDS) for the activation of Middle Head Safety Injection (MHSI), the application of Secondary Passive Residual Heat Removal System (SPRHR) for the residual heat removal. In the article, calculations are carried out for HPR1000 nuclear power plant with nuclear system safety analysis code ATHLET (Analysis of Thermal-Hydraulics of Leaks and Transient) 3.1 [1]. By means of conservative deterministic safety analysis approach, transient analyses concerning selected typical design basis conditions, i.e., Large Break Loss-Of-Coolant Accident (LB-LOCA), Small Break Loss-Of-Coolant Accident (SB-LOCA), Steam Generator Tube Rupture accident (SGTR) and Feed water Line Break (FLB) are performed. The ATHLET results are also compared with the results performed by CGN (China General Nuclear) developed code LOCUST with similar assumptions. The comparisons indicate that, although some discrepancies are detected, the trends of system responses predicted by the two codes are generally in agreement with each other for different accident scenarios. The results also demonstrate that the acceptance criteria for each accident can be met with significant safety margin. Thus, the effectiveness of safety system configuration and accident management procedures is guaranteed.

Development and validation of a numeric tool for partially degraded core quenching under low injection rates

Quenching of reactor core has been studied numerically with the help of several numeric tools reported in literature. These tools have been developed to simulate core quenching under design basis conditions. Hence, most of these tools make use of classical homogeneous equilibrium model of two phase flow which are reported to perform better under high pressure conditions with relatively small temperature gradients. System codes such as RELAP5 make use of two fluid model. Such system codes can simulate core quenching under design basis injection rates (1 g/s per unit length of single fuel pin) prescribed under Emergency Operating Procedures (EOPs). However, because of possibilities of flow path breakage under Severe Accident conditions, the injection rate reaching the core is expected to be lesser than the typical injection rate used for Severe Accident Management Guidelines (SAMG). A numeric tool ‘Program for Degraded Reactor Core Reflood’ (PDRCR) has been developed to simulate quenching of partially degraded reactor core under low injection flow rates. PDRCR has specific modules to tackle issues such as water packing, high temperature gradients in the fuel and simulation of parallel flow paths to account for flow re-distribution in partially ballooned fuel pin condition. This paper presents validation results for PDRCR. The predictions of PDRCR for SEFLEX and DRCRE test facilities and comparison with the experimental results are presented. The predictions of PDRCR are found to be better than RELAP5 for low injection flow conditions.

Experimental investigation of the coolability of blocked hexagonal bundles

The CODEX-COOL experimental series was carried out in order to evaluate the effect of ballooning and pellet relocation in hexagonal bundles on the coolability of fuel rods after a LOCA event. The effects of blockage geometry, coolant flowrate, initial temperature and axial profile were investigated. The experimental results confirmed that a VVER bundle up to 80% blockage rate remains coolable after a LOCA event under design basis conditions. The ballooned section creates some obstacles for the cooling water during reflood of the bundle, but this effect causes only a short delay in the cooling down of the hot fuel rods. The accumulation of fuel pellet debris in the ballooned volume results in a local power peak, which leads to further slowing down of quench front.

Containment leakage characterization with BARCOM test results for design and over pressure conditions

BARC containment (BARCOM) is a 1:4 scale test model of 540MWe IPHWR which was constructed at Tarapur for carrying out leakage studies, at various overpressure conditions, apart from performing ultimate load capacity assessment of the containment model. Various leakage tests were conducted up to the design pressure (Phase II experiments) as well as beyond the design pressure (Phase III experiments).A mathematical model has been developed, by simulating the containment model, to assess the leakage contribution from containment model by various modes such as through the concrete porosity, penetrations and cracks. The modeling involves solving mass and energy equation for the containment together with the evaluation of stress and strain in containment structures and tendons which were further used for calculation of pre-stressing pressure and for estimating crack width. The paper presents the methodology adopted for leakage characterization studies, useful for nuclear containment performance assessment under beyond the design basis conditions. The limitations associated with the existing predictive leakage models based on the present study are highlighted with an objective to further improve these models in future based on the present experimental data and analysis methodology evolved in this paper.

Comparison of CATHARE results with the experimental results of cold leg intermediate break LOCA obtained during ROSA-2/LSTF test 7

Thermal-hydraulic analysis is a key part in support of regulatory work and nuclear power plant design and operation. In the field of Loss Of Coolant Accident, evolutions of the regulations are discussed in various countries taking into account the very unlikely character of a double-ended guillotine break and questioning the necessity to study such an event with Design Basis Conditions assumptions. As a consequence, the consideration of intermediate size piping rupture becomes more and more important. The paper presents the modeling of the Test Facility ROSA-2/LSTF in the calculation code CATHARE 2.V2.5. OECD/NEA ROSA-2 Project Test 7 was conducted with the Large Scale Test Facility on June 14, 2012. The experiment simulated the thermal-hydraulic responses during a PWR 13% cold leg Intermediate Break Loss Of Coolant Accident (IBLOCA). The break was simulated by a cold leg upwardly mounted long break nozzle. The facility and the experiment conditions are modeled in CATHARE. The vessel is modeled by using a 3D module. A thermal-hydraulic analysis is conducted and the obtained results are subsequently compared with the experimental results from ROSA-2/LSTF Test 7. Evaluation of the differences between experimental and calculated results is discussed.

Accident analysis on supercritical water-cooled reactor during design basis and beyond design-basis condition

Supercritical water-cooled reactor (SCWR) is the only water-cooled reactor among six Generation IV reactor concepts. Safety analysis is one of the most important tasks for SCWR design. A typical thermal spectrum SCWR with passive safety system during design-basis accident (DBA) and beyond design-basis accident (BDBA) is performed. For DBA, reactor system is modeled based on a revised code ATHLET-SC. Loss of coolant accident is chosen to perform safety analysis and sensitive analysis. The results achieved demonstrate the feasibility of proposed passive cooling system to provide sufficient cooling. However, it should be noted that if one of safety systems fails to actuate during loss of coolant accident, although the likelihood is fairly low, there is potential risk of cladding failure. Consequently, the DBA will develop into the BDBA. For BDBA, a postulated severe accident is analyzed after melt pool is formed in the lower plenum. Heat transfer behavior in the melt pool as well as two-dimensional heat transfer effect in the lower head wall is discussed. Then, key parameters are chosen to perform parametric analysis. Results show that the safety margin to critical heat flux is significant. After considering two-dimensional heat conduction effect in the lower head, the safety margin could be further increased.

Preliminary Safety Analysis of Mixed Oxide Sphere-pac Driver Fuels in the CP-ESFR WH Core

Abstract Future sodium cooled fast reactors (SFRs) with Minor Actinide (MA) burning potential have to demonstrate an enhanced safety, minimal waste production and increased proliferation resistance in addition to their high economical potential. In this context, the PELGRIMM project, part of the European Framework Program 7 (FP7), investigates new MA-bearing oxide fuel developments for SFRs. Two fuel designs are studied: the standard pellet and the innovative sphere-pac fuel. Due to its favorable swelling behavior expected under irradiation and its dustless fabrication, the latter one appears to be a promising option. A special work package in PELGRIMM addresses the safety performance of this fuel under operational, design basis (DBC) and design extension conditions (DEC). For the preliminary safety assessment, the Working Horse (WH) core, designed and analyzed in the former CP-ESFR project was chosen. The SIMMER-III code was applied after implementing key features as the specific thermal conductivity of sphere-pac fuels. The feasibility of loading the core with MOX sphere-pac instead of pellet fuel, regarding operation and safety limitations, was first proven. Both fuel designs for the WH core at beginning of life and after irradiation were then compared at nominal and under unprotected loss of flow conditions. Results show a similar behavior of these fuels under irradiation, opening a perspective to the practical application of sphere-pac fuels in SFR systems.

Upper limits to americium concentration in large sized sodium-cooled fast reactors loaded with metallic fuel

Transient analysis of a large sized sodium-cooled reactor loaded with metallic fuel modified by different fractions of americium have been performed. Unprotected loss-of-offsite power, unprotected loss-of-flow and unprotected transient-over-power accidents were simulated with the SAS4A/SASSYS code based on the geometrical model of an IFR1For acronyms and abbreviations, see the Nomenclature at the end of the paper.1 with power rating of 2500MWth, using safety parameters obtained with the SERPENT Monte Carlo code. The Ti-modified austenitic D9 steel, having higher creep rupture strength, was considered as the cladding and structural material apart from the ferritic/martensitic HT9 steel.For the reference case of U–12Pu–1Am–10Zr fuel at EOEC, the margin to fuel melt during a design basis condition UTOP is about 50K for a maximum linear rating of 30kW/m. In order to maintain a margin of 50K to fuel failure, the linear power rating has to be reduced by ∼3% and 6% for 2wt.% and 3wt.% Am introduction into the fuel respectively.Hence, an Am concentration of 2–3wt.% in the fuel would lead to a power penalty of 3–6%, permitting a consumption rate of 3.0–5.1kgAm/TWhth. This consumption rate is significantly higher than the one previously obtained for oxide fuelled SFRs.

Critical Section Selection Methodology for the U.S. EPRTM Standard Nuclear Power Plant

A three-tier critical section selection methodology that is used to identify critical sections for the U.S. EPRTM standard nuclear power plant (NPP) is presented. The critical section selection methodology includes three complementary approaches: qualitative, quantitative, and supplementary. These three approaches are applied to Seismic Category I structures in a complementary fashion to identify the most critical portions of the building whose structural integrity needs to be maintained for postulated design basis events and conditions. Once the design of critical sections for a particular Seismic Category I structure is complete, the design for that structure is essentially complete for safety evaluation purposes. Critical sections, taken as a whole, are analytically representative of an “essentially complete” U.S. EPRTM design; their structural design adequacy provides reasonable assurance of overall U.S. EPRTM structural design adequacy.

Transient analyses of the 400MWth-class EFIT accelerator driven transmuter with the multi-physics code: SIMMER-III

In the EUROTRANS Programme of the 6th FP of the EC, EFIT, the European Facility for Industrial Transmutation, a generic conceptual design of a 400MWth ADS transmuter loaded with a CERCER U-free fuel with an MgO matrix, is developed. The safety objectives for the EFIT are achieved on the basis of the defense-in-depth concept. For the safety assessment various protected and unprotected design basis conditions (DBC) and design extension conditions (DEC) transients have been analyzed as e.g. the protected and unprotected loss of flow (PLOF/ULOF), beam trip transients, over power transients and especially the unprotected blockage accidents (UBAs). Key safety analyses of the EFIT core have been performed with the SIMMER-III code, currently the only code which can handle core disruptive conditions for heavy liquid metal cooled reactors. Unprotected transients set upper safety limits and play an important role in the overall safety assessment and transient behaviors of an ADS with MgO based fuel and T91 cladding. The high temperature range is still affected by significant uncertainties, therefore, unprotected accidents with a potential of fuel failure and fission gas release deserve special attention. In this paper, analyses of important and challenging transient cases will be presented. The SIMMER-III simulated results indicate that under the current conditions, the EFIT core can survive in most of the transient cases without pin failure except in the UBA cases. The analyses thus demonstrate the general high safety potential of an ADS like EFIT. Results of the UBA cases show that pin failure happens in the core center and propagation of the pin failure could be possible, while the core will finally arrive at a lower power condition because of the sweeping out of the fuel from the active core.

공기구동 게이트밸브의 운전 성능평가 방법에 관한 연구

The valve performance has been evaluated from the theoretical equation based on design information such as packing thrust, spring preload and friction coefficient(μ). The accuracy of those data can be lower than that of vendor's initial design data. Especially, the friction coefficient can be degraded with time than the original condition and the valve performance calculated using the previous friction coefficient can not be available. Accordingly, this paper is describing a new performance evaluation method of valve based on diagnostic test data which are acquired from a site valve tested in static and dynamic conditions. Especially, this paper provides a new method using friction coefficient(μ) which is derived from the diagnostic test data acquired in the valve's design basis condition.

Feasibility Study of a Passive DHR System with Heat Transfer Enhancement Mechanism in a Lead-Cooled Fast Reactor

The KLFR is a pool-type lead-cooled fast reactor, which has a core thermal output of 900 MW(thermal), and a reactor vessel auxiliary cooling system (RVACS) is employed to secure reliable decay heat removal (DHR) during the worst anticipated design-basis condition. Since the RVACS design is based on reliable and economic considerations, a sufficiently large DHR capacity and compact reactor vessel size are desirable. However, these two requirements compete with each other because a sufficient DHR capacity can be achieved by a larger vessel size with a consequential heavy lead coolant weight. An advanced RVACS concept that has a larger capacity with a more compact vessel size was developed. To increase the DHR capacity of the KLFR, which uses natural-air circulation cooling, the feasibility of heat transfer enhancement by introducing new design concepts to essentially reduce the heat transfer resistance of the radial heat transfer elements was investigated. As a result of this work, the parametric analysis results showed that the passive DHR capacity of the KLFR can be substantially increased by up to 24% when compared with the classical RVACS concept, and this feature makes a compact reactor vessel very feasible. With the proposed advanced RVACS concept, one could expect that the heat removal capacity of an RVACS-type passive DHR system will be increased.