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The Integrated Pressurized Water Reactor (IPWR) designed with series of passive safety system has been regarded as a very promising reactor concept. In this paper, in order to assess the accident mitigation ability of passive safety systems in China 100 MW IPWR, the complete system models, including the reactor vessel, passive safety injection system and passive residual heat removal system, were established using RELAP5 software. The comprehensive and quantitative comparative analysis of auxiliary feed water system and passive safety system under typical accident scenarios, including loss of flow accident and Station Black-out (SBO), were investigated, respectively. Results show that the passive residual heat removal system has the similar accident mitigation capability compared with the auxiliary feed water system, which is adopted in most of traditional PWR design. And the IPWR could realize the reactor safe shutdown under typical accident scenarios with passive safety systems. Finally, the Small break loss of coolant accident (SBLOCA) induced by pressurizer surge line rupture was calculated to verify the function of passive safety systems for IPWR. Results demonstrate that the passive safety system could provide emergency coolant to the reactor and remove the reactor decay heat successfully. This work is meaningful for the quantitative evaluation of passive safety systems in IPWR.

During the small break loss of coolant accident (SBLOCA) in a pressurized water reactor (PWR), single and two-phase natural circulations (NCs) as well as reflux condensation can be the three major flow modes in the primary side to remove the residual core heat. The performance of U-tube steam generators (UTSGs) during the accident is of great significance for the reactor safety and integrity. However, flow behaviors in UTSG parallel U-tubes are considerably complicated and the reverse flow phenomenon may occur during both the single and two-phase NCs. Despite the rich researches on the phenomenon occurring under the single-phase NC, studies concerning that under the two-phase NC are still quite scarce. In this paper, the vapor-liquid two-phase flow condensed in the U-tube is numerically simulated using the computational fluid dynamics (CFD) method. The distribution of vapor volume fraction αg in the U-tube is investigated and analyzed. It is shown that a negative slope region obviously exists in the curve of pressure drop versus mass flow rate, and two-phase flow excursion can occur if the mass flow rate reduces to a critical value, namely the critical mass flow rate Mc. Afterwards, some influence factors on αg and Mc are numerically studied, which include the inlet vapor volume fraction, the bubble diameter, the inlet slip ratio, the U-tube wall temperature and the primary side pressure. The results can provide references for the long-term core cooling during a SBLOCA and the optimization design of SGs.

Thermal-hydraulic response of a two-loop pressurized water reactor is studied on loss of residual heat removal system (RHRS) during mid-loop operation. Decay heat during mid-loop operation is removed by RHRS. On loss of RHRS, no cooling system is available to cool the core. Transient thermal-hydraulic analysis of the plant is performed using RELAP/SCDAP on loss of RHRS. For this purpose, first steady-state is achieved at mid-loop conditions. System obtains steady-state at primary coolant inventory of 58000 kg, water level of 7.8m in the reactor pressure vessel (RPV) and upper plenum pressure of 104.41 kPa. On loss of RHRS, the core uncovery time, maximum reactor system pressure and cladding temperature are determined under four accidental scenarios in absence of any mitigation measure. The results show that the core uncovery time for the open case of reactor pressure vessel top vent valve (RPV top vent valve), pressurizer relief valve (PZR valve), steam generator man way and pressurizer man way are 423.1, 214.4, 50.1, 67.2 min respectively. The effect of reflux condensation for the above accidental configurations has also been studied. It is found that reflux condensation is a prominent phenomenon to increases core uncovery time for open case of RPV top vent valve and PZR valve.

In pressurized water reactor, one of the most important barriers to the release of radioactive material to the environment is the steam generator heat exchanger tubes. Degradation of steam generator tubes could result in increased primary-to-secondary leakage and delayed diagnosis of the event could eventually result in through-wall tube rupture. To prevent through-wall tube rupture, an effective trend monitor is essential. Moreover, to increase the cost-effectiveness and reliability of energy production from nuclear plants, reduction in maintenance and repair downtime caused by steam generator tube defects is crucial. This paper reviews the state-of-the-art diagnostic techniques and condition monitoring methods for heat exchanger tubes. In particular, it discusses steam generator tube degradation and integrity issues, the physical phenomena, and the analysis of theoretical and experimental research conducted in the past few decades on the inspection of steam generator tubes. In addition, plants’ monitoring systems are categorized and the predictive models utilized for the monitoring and evaluation of water chemistry is discussed. The major contribution is the presentation of critical parameter trends, deviations and empirical signatures observed when steam generator tube rupture occurs in a fully functional CNP1000 pressurized water reactor, and the analysis of machine learning approach for rupture event diagnosis. Furthermore, the operator response to thermal-hydraulic parameter trend indicating cracks and incipient leaks in steam generator tubes, as well as comparative advantages and demerits of water chemistry monitors are also presented. Finally, possible future research focus and likely challenges are discussed and a technique appropriate for effective online diagnosis of increased primary-to-secondary leakage event is recommended.

In this paper, the safety margin of IVR (in-vessel melt retention) for small IPWR IP200 (integrated pressurized water reactor) is evaluated by numerical simulation. The thermal hydraulics and core degradation are modeled by SCDAP/RELAP5 code to provide initial conditions for the IVR. Meanwhile, the lower head behaviors are modeled by the finite element code COUPLE to present the dynamic evolution of molten pool with detailed temperature distribution. The transient analyses of thermal load firstly focus on the progression of corium solidification and external cooling, and then analogize to the IVR studies of other IPWRs to discuss the impacts which caused by the inherent design features of IP200. Besides, the differences between FIBS steady model and SCDAP transient model are discussed in terms of the heat transfer governing equation to summarize the benefits of transient calculation. In order to explain the calculation rationality, the result of maximum surface heat flux is also validated with the dimensionless data of related experiment. Finally, the mass composition of molten pool and the gap thermal resistance of inner wall are selected for parametric sensitivity analyses to assess the conservatism of maximum heat flux. For the quantitative conclusions, the low coolant storage of IP200 makes the core degradation very fast. The duration is about 9500s and the maximum temperature of liquid corium is close to 2900 K. The initial inner heat of molten pool is very high that the transient heat flux can get approximately 195% above the steady value. At the moment of maximum thermal load, the peak heat flux reaches 0.46 MW/m2, still lower than the local CHF. The lowest safety margin locates at the bottom angle and corresponding q/CHF is 0.65. The present work can provide references for the safety design of small IPWR.

The flexible operating conditions of small pressurized water reactors (SPWRs) demand frequent movement of control rods, causing a serious problem of control rod wear and tear. This study focused on the optimization of the control rod travel for an SPWR. Three optimization strategies were proposed to improve the reactor power control scheme by introducing proportional-derivative, lead–lag, and incomplete differential control modules in the power deviation channel, and a weighting factor in the temperature deviation channel. The parameters of the three control modules and the weighting factor were tuned using the non-smooth optimization algorithm in MATLAB. The optimization strategies were implemented in a control simulation platform of the SPWR, based on which dynamic simulations of the SPWR were performed under typical load change transients at different power levels. The results demonstrate that the optimization strategies can significantly reduce the control rod travel of the SPWR with satisfactory control performance.

Resulting from a mechanical failure of a control rod system the reactor coolant pressure ejects the control rod assembly from the core. The control rod ejection accident belongs to the category of reactivity initiated accident (RIA) and is defined as a design basis accident. The U.S. Nuclear Regulatory Commission (U.S. NRC) has established the Phenomenon Identification and Ranking Tables (PIRT) for rod ejection accident in pressurized water reactor (PWR) which will be used in this study. The present work is focused on selected physical parameters which can be calculated by coupling two best estimate codes for VVER-1000/V-320. Namely, TRACE V5 p4 was used to develop a thermal-hydraulics model and PARCS 3.2 for establishing a detailed neutronics model of a reactor core. Both models were executed in a merged version TRACE V5 p4. The analysis was focused on evaluation of the following parameters according to PIRT: relative power in the reactor core and rejected fuel assembly, inserted reactivity, moderator properties (density and temperature), and inner/outer temperature of the heat structure. Nominal power of 100% and the control rod ejection time of 0.1 s were considered. The inserted reactivity has reached 0.45$ and the maximum power in the reactor core has risen to 179.7%.

Lead-cooled Fast Reactor (LFR) has been chosen as one of the six Generation-IV reactor concept candidates due to its high performance in neutronics, thermal hydraulics and inherent safety features. Passive residual heat removal system (PRHRS) is a passive system that guarantees the residual heat removal of LFR even in the hypothetical beyond-design basis conditions. Based on a 1000 MWth lead-bismuth cooled fast reactor, the design and performance analysis of PRHRS were conducted using RELAP5 code. First, a PRHRS removing heat by Independent Heat EXchanger (IHEX) was designed, which utilizes natural circulation, gravity field and has the characteristics of high reliability and inherent safety. Secondly, the lead-bismuth heat transfer model was developed and verified based on RELAP5/MOD 4.0 code, as the heat transfer model of Pressurized Water Reactor (PWR) is not suitable for the lead-bismuth heat transfer calculation. The RELAP5 model of LFR was established and verified by comparison with design parameters. The PRHRS performance was investigated under Loss of Power (LOP) transient to assess the feasibility of the PRHRS design. The results demonstrated that the heat removal rate of the PRHRS removing heat by IHEX exceeded the core decay rate at about 5 h. The cladding peak temperature of the PRHRS reached 562 °C, which was below the limit (750 °C). Therefore, the PRHRS removing heat by IHEX has the capability to remove the LFR residual heat and has the characteristics of high performance and inherent safety, which can be used as an independent residual heat removal system for commercial LFR. The work provides reference for the design and performance analysis of LFR PRHRS in the future.

Fusion-fission hybrid reactor is believed to be a feasible implementation of fusion energy in a short time as it requires less advanced technology in fusion physics and engineering than pure fusion reactor. In this paper, based on the mature technology of pressurized water reactor (PWR) and existing fusion reactor blanket concepts, a new conceptual design scheme of water-cooled pressure tube blanket was proposed for hybrid reactor and both the material selections and the structure design were performed. In the blanket, a trapezoidal U-shaped first wall (FW) was adopted to decrease the gaps between blanket modules. Due to high heat flux from plasma and intense irradiation by high-energy neutrons, it's essential to conduct thermal-mechanical coupling analyses and optimizations of the FW to ensure reasonable temperature and stress distributions on it. In this work, CFX-Workbench coupling method was adopted to perform the simulation. The influences of different FW geometrical configurations and boundary conditions on the temperature and stress distributions were researched in detail to optimize the FW structure. Besides, the thermal and mechanical characteristics of the helium tanks, the cooling channels and the breeding zones were researched in this paper. The calculation results verified the reasonability of the conceptual design scheme preliminarily and could provide a meaningful reference for the further conceptual design, analysis and optimization of the hybrid reactor blanket.

This article presents a review of the evolution of Pressurized Water Reactor Designs used for electric power generation in the United States from December 1957 through 2017. For most of this period the nuclear industry in the United States was the world leader in commercial nuclear PWR power plants and the evolution of PWRs in the world closely parallels the evolution in the United States. The Shippingport Nuclear Power Plant achieved its first core criticality on December 2, 1957. Since then there have been generation I, generation II, generation III and generation III + reactor designs developed and implemented. The article also addresses the development of pressurized water small modular reactors. The power output of the reactor has ranged from about 45 MWe for the SMR to over 1600 MWe for the current French EPR design. Plant safety systems have evolved from single failure tolerance systems to probabilistic approaches that have four separate safety divisions plus additional diverse non-safety systems and controls which dramatically decrease the probability of possible impacts on the public. The Three Mile Island nuclear accident is discussed and evaluated for its impact on the design of Pressurized Water Reactors. The more recent developments of passive safety systems enable reactors to dramatically decrease their reliance on AC power.

The core make-up tank (CMT) is a crucial device of the passive core cooling system in the Westinghouse AP-series pressurized water reactor nuclear power plants. In the prototype plant, the CMT is a vertical cylindrical vessel with hemispherical top and bottom end sockets. Although the structure of the CMT system is simple, it involves many thermal-hydraulic phenomena, most of which significantly impact the peak temperature of the fuel cladding and the liquid level of the core. Since the CMT not only provides full-pressure injection but also triggers the auto depressurization system by its liquid level signal during small-break loss of coolant accident (SBLOCA), its thermal hydraulic behaviors affect the performance of the passive core cooling system (PXS). Therefore, it is necessary to carry out separate effect tests to verify the feasibility of CMT applied to passive safety and injection systems and to study the operating characteristics of CMT. This article first presents the design of a CMT separate effect test facility and analysis of the operating characteristics of the CMT with the experimental data. The influence of pressure, non-condensable gas, wall condensation, and thermal resistance on the CMT behavior were investigated. The study found that the non-condensable gas would stratify in the CMT, which had negligible effect on the wall condensation, but could greatly reduce the gas-liquid interface condensation.

This work investigates the possibility of increasing the incineration rate of reactor-grade plutonium (rgPu) in the pressurized water reactor (PWR). In an attempt to reduce the production of plutonium isotopes from the conventional fuel and increase the rate of burning the accumulated plutonium, thorium-232 is suggested to be used as a fertile material in the standard fuel assembly and a suggested Blanket-seed assembly. The neutronic properties of Th-rgPu in both the standard fuel assembly and the BS assembly were analyzed and compared with the standard fuel to distinguish which of them is more effective in burning the rgPu. The performance of Th-rgPu in the standard fuel assembly and the BS assembly has been assessed by tracking the infinity multiplication factor (Kinf), the amount of the rgPu incinerated, reactor safety parameters, the most important fission products, the radioactivity of the spent fuel and the radial power distribution through the assembly. The neutronic analysis confirmed the viability of BS assembly in burning a large quantity of rgPu, maintaining the neutronic safety parameters at acceptable values.

Collapse of liquid level in reactor pressure vessel (RPV) and uncovery of core in typical pipeline rupture accidents, such as steam generator tube rupture (SGTR), main steam line break (MSLB), and loss of coolant accident (LOCA) for pressurized water reactor (PWR), may occur. This study used APROS to establish a full-scale simulation model of VVER440 reactor and evaluated different break sizes that can lead to RPV liquid level collapse and core uncovery during these accident processes. It is demonstrated that in SGTR accident, the liquid level in RPV collapsed under the condition of four or more heat transfer tube ruptures without the risk of core uncovery in all analyzed cases. In the MSLB accident, RPV liquid level collapse and core uncovery were absent due to the integrity of the primary circuit. In the LOCA accident, the RPV liquid level collapsed under a break size of 0.5% of pipe area (PA) of cold and hot leg. Notably, core uncovery occurred when the break area was 1.8% of PA in the cold leg and 65% of PA in the hot leg. The time required until core uncovery was earlier in the cold leg than that in the hot leg.

The Emergency Core Cooling System (ECCS) is actuated following Loss of Coolant Accidents (LOCA) to inject coolant into the primary loop through a T-junction configuration in nuclear power plants. The injection angle; i.e., the T-junction angle, in the cold leg injection system in a pressurized water reactor is an important parameter during a postulated LOCA. The amount of the break flow rate and the time at which the refill period initiates impact the thermal-hydraulic behavior of the reactor coolant system and eventually affect the integration of reactor nuclear core. In the current work, RELAP5 code is used to study the effect of the cold leg injection angle on LOCA scenario in a standard four-loop Westinghouse Pressurized Water Reactor (PWR). Injection angles investigated in the study are 45°, 90°, and 135°. The whole reactor system is simulated using RELAP5 code whilst the injection junction is numerically developed using ANSYS-FLUENT commercial code. The break flow, core recovery, and thermal-hydraulic parameters are calculated for the different injection angles. Research results are validated with published data available in the literature. The results manifest that the injection angle of 45° gives the highest and the most uniform core level during the core refilling phase, the best temperature behavior for both fuel centerline and maximum cladding temperatures, where both temperature distributions reach a stable value earlier than that of the other angles. ANSYS-FLUENT code is utilized to explain the complex mixing phenomena of fluids during the analysis.

In nuclear power plants (NPPs), one of the important yet uncertain phenomena during a station blackout (SBO) accident is the possibility for a counter-current natural circulation flow in the reactor coolant system (RCS) pressure boundary, which could divert the core melt progression and/or the in-vessel fission product (FP) distribution and, in turn, lead to a heatup and eventual creep rupture. From the risk and accident management points of view, one of key concerns is to predict more explicitly whether such a creep rupture can occur in the RCS pressure boundary, and if it indeed can, also predict the most probable time and the most likely location. This paper presents plant-specific analysis results on the potential characteristics of natural circulation-induced creep rupture of the RCS pressure boundary that could potentially take place during an SBO accident of a reference pressurized water reactor (APR1400). MELCOR-CFD (Computational Fluid Dynamics) coupled analysis was performed to determine more precisely the relevant physical parameter that could greatly influence on the behaviour of natural circulation and the ensuing creep rupture. Multiple sensitivity analyses were also carried out to investigate the influence that particular MELCOR modeling schemes and plant conditions have in estimating the RCS creep rupture characteristics. The analysis results and relevant insights were summarized in terms of particular points of interest.

The mainstream deterministic and probabilistic methods for safety analysis of the nuclear power plant (NPP) are the best estimate plus uncertainty (BEPU) analysis and probabilistic safety analysis (PSA), respectively. With the development of techniques, a more “realistic” simulation of the NPP has become a long-term goal to improve the economic efficiency. Nowadays, the combination of BEPU and PSA, also known as the risk-informed uncertainty method, is under development. In this paper, the cold leg double-ended break accident of a typical pressurized water reactor was selected to be the target scenario for a preliminary study of the combined method. A failure analysis of the emergency core cooling system based on PSA was carried out first, and the failure probability of each event sequence was determined. Then the conditional core damage probability of each event sequence was evaluated through BEPU analysis, and the core damage frequency of the reactor was finally determined.

In pressurized water reactor plants, the degradation of steam generator tubes has been one of the main economic and safety problems. Steam generator tubes are subjected to the degradation mechanism of stress corrosion cracking, in which an unstable crack growth may lead to their structural failure. In general, deterministic criteria provide the least conservative assessment in order to plug or sleeve the cracked tubes. As the crack growth rate under stress corrosion cracking is of high random nature, the structural integrity assessment of steam generator tubes must be performed taking into account the failure probability at current and next inspections. In this paper, a stochastic model is applied for performing the operational assessment of steam generator tubes for any time interval. The results obtained with this model are compared with others, such as Monte-Carlo simulation and non-evolutionary statistical model. The proposed methodology is able to provide efficiently a less conservative failure probability. The plausibility of stochastic models in the structural integrity assessment of steam generator tubes is reached in favor of their operational life extension.

In recent years, “Nanofluid” and “Energy-Efficient Cooling” have been achieved growing attention for use in new nuclear reactor technology. Also, the assurance of reactor safety has become of top priority in nuclear energy development because of the significant role of nuclear power in providing worldwide electricity. The use of nanofluid coolant is an effective way to enhance safety margins and heat transfer performance. Studies also prove that nanofluids are promising cooling fluids for applying as a future coolant in nuclear reactors. The current paper investigates the nanofluid effects as a coolant on the velocity of bulk, pressure drop, heat transfer coefficient, power peaking factors, and minimum departure from nucleate boiling ratio in a NuScale reactor, a natural circulation pressurized water reactor (PWR). It also analyzes the thermal characteristics of Al2O3 nanofluid with specific attention to their boiling heat transfer behavior. The study first presents the NuScale designed core with the use of water and different concentrations of Al2O3 nanofluid (0.001–10% volume fractions and 10–90 nm particle sizes). Then it shows the simulated equivalent cell with surrounding coolant utilizing computational fluid dynamic code (CFD). The study also describes the outcomes of replacing water with nanofluid coolant on natural circulation parameters, heat transfer coefficient, and minimum departure from nucleate boiling ratio with changes in nanoparticle concentrations and sizes. The results demonstrate the potential of this innovative coolant to enhance the minimum departure from nucleate boiling ratio and heat transfer coefficient. Overall, the presence of Alumina nanoparticles in water improves thermal performance and safety. The investigation also shows that the nanoparticles do not change the power peaking factor and neutronic performance significantly.

Two heterogeneous nodal methods based on the Variational Nodal Method (VNM) are investigated with diffusion approximation in three-dimensional Cartesian geometry. The first one is named as Function Expansion (FE) method while the second is Finite Sub-element (FS) method. Based on our previous work and the code Violet-Het1D in one-dimensional slab geometry, a code named Violet-Het3D was developed to handle the Pressurized Water Reactor (PWR) control rod cusping effect and pin-by-pin calculation by using either of these two methods. To eliminate the control rod cusping effect, Violet-Het3D provides a different idea from the existing methods. Neither homogenization procedure nor mesh adjustment is needed in Violet-Het3D by taking advantage of the treatment for heterogeneous nodes. Numerical results show that both the FE and FS methods can eliminate the cusping effect and obtain accurate power distribution while the FE method has relatively higher efficiency and accuracy. In contrast, for pin-by-pin calculation, the FS method obtains more accurate eigenvalue and pin power distribution than the FE method.

In a nuclear reactor severe accident, the molten core debris may cause steam explosion when contacting with water in the reactor cavity below the reactor pressure vessel (RPV), which can endanger the containment integrity of the nuclear power plant. In this study, the improved FCI computer code TEXAS-VI considering solidification effects and the original code TEXAS-V were used to analyze the TROI TS-3 test as a benchmark experiment. Then a series of ex-vessel steam explosion simulations were performed. In the experimental comparison, the melt penetration (leading front position in the coolant) in TS-3 was well predicted by TEXAS. The pressure at different positions simulated by TEXAS-VI was notably lower than that of TEXAS-V, but in good agreement with the experiment results. The simulation results showed that the modeling of corium solidification during the mixing process and the breakup criteria in TEXAS-VI largely mitigates the rapid breakup, thereby affecting the steam explosion. Our study continued by using TEXAS-VI to analyze ex-vessel steam explosion energetics for a range of different initial conditions, with the aim to examine the impact of the fuel melt initial diameter and velocity. For conservative estimation, the cross sectional area ratio of the fuel-coolant mixing region to the injected corium was set to maximize the peak explosion impulse based on a number of parametric calculations. In scope of this study conditions, the larger the diameter and the initial injected velocity of corium were, the larger the maximum pressure and the liquid kinetic energy from the steam explosion were. The void fraction can influence the fragmentation rate during the explosion stage and mitigate the propagation of the pressure pulse during the propagation stage. The liquid kinetic energy would be affected by the pressure from the steam explosion and the void fraction at the end of mixing stage.

The main purpose of this paper is neutronic-thermal hydraulics coupling analysis of the fuel channel of an advanced Small Modular Reactor (SMR), which is nominated as a near term option of the generation IV reactors. The CAREM25 is chosen as the reference SMR. By considering hexagonal and square lattices of the fuel assemblies in the SMR reactor core, neutronic-thermal hydraulic coupling of hot and average fuel channels were conducted via utilizing of the MCNPX and CFX codes. Central fuel assemblies are determined as the hottest fuel assembly in the proposed SMR cores with hexagonal and square fuel assemblies with power peaking factor of 1.778 and 1.674, respectively. The hottest fuel rod power peaking factor is calculated as 1.846 and 1.954 for hexagonal and square fuel channels, respectively. The calculated linear power densities of the hottest and the average fuel rod of hexagonal channels are 20.18 ( k W m ) and 10.77 ( k W m ) , respectively, as well as of square channels are 24.75 ( k W m ) and 10.36 ( k W m ) , respectively. Results show that the maximum axial power along the fuel rods occurred below the mid-plane of the rod. The ratio of the hot to average rod axial power peaking factor as a safety parameter used to calculate the maximum heat flux in the hottest channel, is calculated close to 2 and 2.4 for the core with hexagonal and square fuel assemblies, respectively. The calculated power peaking factors and linear power densities of fuel rods are in suitable range and in good agreement with PWR safety issues. The maximum temperature of fuel centerline in both types of the proposed SMRs is much lower than that of large PWRs (e.g., 2156 K for hot rod in VVER-1000). Between hexagonal and square fuel assemblies, the hexagonal FAs are preferred because of the two reasons; first, the axial fuel centerline temperature in the hottest hexagonal channel is lower than that of hottest square channel, so it is a good feature with respect of safety margins. Second, the outlet temperatures of the channels in the SMR core with hexagonal FAs are higher than the outlet temperature of the channels in the SMR core with square FAs. Thus the thermal efficiency of the reactor core with hexagonal FAs is more than that of the core with square FAs.

Fuel temperature (Doppler) feedback modeling in the coupled sub-channel thermal-hydraulic/time dependent neutron transport codes system CTF/TORT-TD was improved by accounting for the burnup dependence of the fuel thermal conductivity. TORT-TD is a three-dimensional (3D) time dependent neutron-kinetics code based on the discrete ordinates (SN) method. CTF is the Reactor Dynamics and Fuel Modeling Group (RDFMG) version of the sub-channel thermal-hydraulics code COBRA-TF (COlant Boiling in Rod Arrays – Two Fluid). A burnup-dependent fuel rod model, which takes into account the degradation of the fuel thermal conductivity at high burnups and the effects of burnable poisons, such as Gadolinium, was implemented in CTF. The model is applicable to UO2 (uranium dioxide) and MOX (mixed oxide) nuclear fuels – it includes the modified Nuclear Fuel Industries (NFI) model for UO2 fuels and the Duriez/Modified NFI model for MOX fuels. The in-pellet fuel temperature distributions predicted by CTF/TORT-TD were compared to reference CTF/TORT-TD/FRAPCON calculations, in which the fuel rods were modeled with the fuel performance code FRAPCON. These comparisons were carried out for a 4 × 4 pressurized water reactor (PWR) pin array at hot full power (HFP) steady state conditions. The CTF/TORT-TD fuel temperature predictions were consistent with the CTF/TORT-TD/FRAPCON results. This fact demonstrated that CTF with the new fuel thermal conductivity model can predict the temperature field within light water reactor (LWR) fuel rods as accurately as FRAPCON. Therefore, CTF/TORT-TD calculations can be carried out in fast scoping studies instead of the computationally expensive CTF/TORT-TD/FRAPCON calculations. The performed statistical analyses indicated an improved accuracy of fuel temperature calculations relative to the CTF/TORT-TD/FRAPCON reference numerical solution. Furthermore, better agreement between CTF/TORT-TD and CTF/TORT-TD/FRAPCON in calculated neutronic reactivity was found when fuel burnup effects were considered in CTF/TORT-TD. Therefore, the improved CTF/TORT-TD can be seen as a high fidelity multi-physics computational tool capable of providing accurate and efficient simulations for practical reactor core design and safety analysis.

The passive containment cooling system which utilizes natural phenomena to remove the energy released from reactor will prevent steel containment from overpressure during the postulated accident and provide long-term cooling to decrease the pressure and temperature of containment. The peak pressure in containment under the design basis accidents is important to the design and evaluation on the passive containment cooling system. In this paper, a dedicated code PCCSAP-3D for the evaluation on the performance of the passive containment cooling system is used to analyze the thermal-hydraulic transients in the containment of the advanced pressurized water reactor AP1000. Some factors are compared to study their e on the peak pressure of containment under the typical loss-of-coolant and main steam-line break accident. The results show that the amount of released coolant, the free volume and shell thickness of containment are dominant for containment pressure during the postulated accident, in comparison with the initial conditions. And the containment pressure response is more sensitive to internal initial conditions than to ambient conditions. The peak pressure under the postulated accidents is mainly contributed from rapid coolant release and it is not sensitive to the containment cooling performed by the passive containment cooling system.

Activated cooling water in nuclear facilities can present a significant radiation source around primary cooling system causing radiation damage to electrical components, increasing doses to personnel and in the case of fusion facilities additional heating to superconducting coils. This paper focuses on activation of oxygen isotopes in water and decay of this activated isotopes, i.e. 16N, 17N and 19O. An analysis of activation of water in pressurized water reactors and in fusion reactors was performed. Different evaluated nuclear data libraries were used in activation calculations (ENDF/B-VIII.0, FENDL-3.1b, JEFF-3.2 and TENDL-2015). The calculated activation rates with different nuclear data libraries agree well for the 16O(n,p)16N reaction and significantly differ for 17O(n,p)17N and 18O(n,γ)19O reactions. In fusion reactor the specific activity of activated water isotopes is in the order of 10 13 Bq/m3/MW, which is five orders of magnitude higher compared to specific activity in a typical fission pressurized water reactor, amounting to 10 9 Bq/m3/MW. The results of specific activity of cooling water were used to perform parametric analysis of dose rates around pipes of cooling system and dose field around a steam generator in a pressurized water reactor as a representative of heat exchangers. The analysis of dose rates around pipes include pipes featuring 1 mm to 8 cm thick walls and from 0.5 cm to 60 cm water radius. Results can be used to estimate dose rates for all studied isotopes, provided the specific activity is known. For heat exchangers the decay of 16N contributes majority to the dose rates in the air surrounding them while 17N and 19O decay contributes together less than 0.1%. For a typical 2 GW thermal power two loop pressurized water reactor the dose rates in air surrounding the stream generator are in the order of several mSv/h.

The objective of this work is to determine which of the main three fuel types (UO2, (Th,U)O2 and MOX fuel) can burn minor actinides (MAs) effectively and at the same time sustain the main operating parameters at an acceptable level. The effect of adding 1 wt % of MAs to the main three fuel types on the infinity multiplication factor (Kinf), thermal neutron flux and the actinides’ concentration is investigated using MCNPX code version 2.7. The MAs can be used to regulate the excess reactivity of the reactor. The variation of the MAs concentration with burnup is studied to distinguish the fuel type that has the ability to burn a large amount of MAs. The transmutations of MAs in the pressurized water reactor (PWR) affect the level of radioactivity of the spent fuel. The main operating parameters are calculated to assess the possibility of burning MAs in the PWR. A comprehensive mathematical study of the results is performed to assess the performance of three fuel types after adding the MAs. Hence, determine the best fuel type. The mathematical study utilizes one of the known multi-criteria decision-making methods, the technique of order preference by similarity to an ideal solution (TOPSIS).

The activated corrosion products are the leading source of radiation in Nuclear Power Plants (NPPs). Particularly in advanced integrated systems with a reduction in piping and advanced safety features such as system-integrated modular advanced reactor (SMART), the corrosion product activity is needed to be investigated. In this work, simulation of corrosion product activity has been evaluated for the SMART reactor under normal and shutdown conditions and compared with a typical Pressurized Water Reactor (PWR). For this purpose, a computer code ACP-SMART has been developed and implemented in MATLAB, which uses OpenMC as a subroutine. The effective group cross-section has been generated by OpenMC under the core design constraints. The results for 56Mn, 59Fe, 24Na, 99Mo, 58Co, and 60Co have been found as 0.072 μCi/cm3, 0.240 μCi/cm3, 0.290 μCi/cm3, 0.360 μCi/cm3, 0.015 μCi/cm3, and 0.020 μCi/cm3, respectively. The total specific activity due to corrosion products in the core, primary coolant, and inside the steam generator is 2.790 μCi/cm3, 0.183 μCi/cm3, and 0.145 μCi/cm3, respectively, which are 89.5%, 5.8%, and 4.7% of the overall contribution. Results indicate that the primary coolant and steam generator specific activity reach their saturation values fairly rapidly. Predominant corrosion product activity during normal operation is due to 56Mn while 58Co and 59Fe dominate after reactor shutdown. 99Mo remains dominant to 24Na and 59Fe during normal operation while 59Fe leads over 99Mo and 24Na after shutdown. A linearly increasing corrosion rate has been employed, and an effect on saturation activity has been investigated. For the linearly increasing corrosion rate, the specific activity behavior has changed considerably. The time taken to reach the saturation activity strongly depends upon the corrosion rate. The saturation activity in the primary coolant and steam generator depends on the CRUD removal rate from the core scale.

This paper presents an offsite modular construction approach, subject to transport constraints, for nuclear power plants. The modular construction principles underlying this work are generalizable across different reactor types; however, this paper focuses only on Small Modular Reactors (SMRs) that are of the Pressurized Water Reactor (PWR) type. Existing work on large nuclear reactor modularization is used to generate a module breakdown scheme for both small and large reactors and a module reconfiguration strategy is proposed that allows modules to fit within road transportable weight and dimension limits. Fully modularized SMRs can transport approximately 80% of their modules by road, compared to only 20% for large reactors. This has a significant effect on cost: a first-of-a-series, fully modular 300 MWe SMR can reduce overnight construction costs to $5470/kWe, a 45% reduction relative to its conventionally-built counterpart. This suggests that by using a standard design and radical modularization, SMRs could compete with current large reactor baseline construction costs of $6100/kWe. Standardization, another aspect of modular manufacturing best-practise, serves to remove repeat design costs and lowers nuclear construction costs by 15%. Additionally, SMRs can further reduce costs through shorter build schedules and production learning. These findings emphasize the importance of modularization for small reactor economics, highlighting the need for further research into the issues surrounding modularization, factory-centered offsite build, and supply chain strategy for small reactors.

Small Modular Reactors (SMRs) are currently being developed and licensed in many countries. The innovations in the designs provide many significant features, such as decreased construction cost and time, enhanced safety, modularity, possibility of factory production etc. Several new SMR designs are based on the concept of an integral Pressurized Water Reactor. One of such designs, namely the NuScale SMR was selected for this analysis. An input model was developed from scratch for the RELAP/SCDAPSIM code using publicly available data from the NuScale Final Safety Analysis Report provided by the U.S. NRC and some validation tests have already been conducted. An extremely severe event is analyzed and discussed in the current work. It is an inside containment Loss of Coolant Accident followed by failure of both the Emergency Core Cooling System and the Decay Heat Removal System. The simulation was conducted for 86400 s (24 h) with the core damage starting after ~4.8 h, which is comparable to the results for similar scenarios presented in the references.

One of the essential components in a pressurized water reactor is the steam generator (SG) that has an important role in efficiency of nuclear power plants. In the present work, a horizontal steam generator (HSG) was investigated based on entropy generation as a consequence of second law of thermodynamic. The thermal hydraulic operation parameters were calculated by considering a single-phase model for the primary loop related to the tube and two-phase model for the secondary loop related to steam generator shell. Then, the effect of each parameter against different constraints such as length-to-diameter ratio, mass flow rate, tube diameter, and inlet temperature on the generated entropy was investigated. To validate the physical equations and the framework used in this model, the results of the developed code are compared with the research articles related to enthalpy generation. The results showed that a significant increase in length-to-diameter ratio and tube diameter resulted in a decrease in entropy generation number (EGN). Therefore, it is concluded that decreasing the entropy generation enhances the performance of steam generators as well as their economic efficiency.

The control of a nuclear power plant is important due to its nonlinear dynamics. An optimized proportional–integral–derivative (PID) controller is a popular tool to this end. PID is tuned by various methods, including a genetic algorithm (GA). The quality and speed of GA optimization depend on the objective function as the performance index. In this paper, integral of absolute error (IAE), integral of the square error (ISE), integral of time-absolute error (ITAE), and integral of time-square error (ITSE) objective functions are evaluated. The results comparison is done based on overshoot/undershoot and the number of function evaluations (NFE) criteria. The PID gains are tuned to control of a typical pressurized water reactor (PWR), based on the point-kinetics model. The results show the high control performance compared to an empirically tuned PID. Also, ISE and ITSE are better performance indexes than IAE and ITAE in steady-states as critical situations. Also, these performance indexes have signified a good robustness against dynamics parametric uncertainties compared to an empirically tuned PID in every power mode. This comparative approach provides a good idea for experts to choose an appropriate performance index.

In pressurized water reactor (PWR) with multiple loops, abnormal working conditions occur when coolant pumps do not work in some loops. In these closed-loop pipelines, thermal stratification is formed under the effect of natural convection cooling and this poses a threat to the safe operation of the reactor. In this paper, the sequential function specification method (SFSM) is implemented to simultaneously predict the spatially and temporally varying internal fluid temperature and convective heat transfer coefficient of two-dimensional pipe under thermal stratification. In the direct problem, the finite volume method (FVM) is adopted to solve the governing equations subjected to initial and boundary conditions. In the inverse problem, the least-square method is employed to obtain the heat flux of the inner wall, and then Newton's law of cooling and the natural-convection correlation are adopted to obtain the fluid temperature and convective heat transfer coefficient. Numerical experiments of natural convection cooling are carried out under different outer wall boundary conditions to demonstrate the effectiveness of the proposed inversion method. The results of temperature estimation are consistent with the results of natural cooling numerical experiment and the relative error is below 4%, whereas the convective heat transfer coefficient is slightly worse due to the difference in the extraction positions. The effects of the number of involved future time steps and the measurement noise on the accuracy of the inverse analysis are also investigated in detail. Our results show that the fluid temperature and convective heat transfer coefficient of natural cooling may be effectively estimated using the proposed inverse method.