scholarly journals A new direction in PWR simplification

2021 ◽  
Vol 7 ◽  
pp. 2
Author(s):  
Nicolaas M. Bonhomme
Keyword(s):  
High Pressure ◽  
Power Level ◽  
Saturation Temperature ◽  
Heat Removal ◽  
Pressure Control ◽  
Severe Accident ◽  
Injection System ◽  
Primary System ◽  
Ambient Conditions ◽  
Gravity Driven

A new approach to PWR simplification is presented, in which a compact Reactor Coolant System (RCS) configuration is introduced, particularly suited for a power level in the range of 600 MWe. Customary PWR primary system components are eliminated to achieve this RCS simplification. For example, RCS pressure control through a “self-pressurization” mode, with core exit at saturation temperature with less than 1% steam, allows elimination of a pressurizer. Also, mechanical control rods are replaced by reactivity control using negative moderator void and temperature coefficient together with variable speed primary pumps, and with an upgrade in the safety boration function. Decay heat removal in shutdown conditions is realized through the secondary side rather than through primary side equipment. The compact RCS can be installed in a small volume, high-pressure containment. The containment is divided into two leak-tight zones separated by a partition plate. Safety equipment installed in one of the two zones will be protected against adverse ambient conditions from leaks or breaks in the other zone. The partition facilitates management of coolant inventory within the RCS and the containment following RCS leaks or breaks. In particular, the safety injection system as commonly known, consisting of accumulators and multiple stages of injection pumps can be discarded and replaced by gravity-driven flooding tanks. Space available around major RCS components is adequate to avoid compromising accessibility during maintenance or in-service inspection operations. In addition, the two-zone, high-pressure containment provides extra margins in severe accident mitigation. Finally, the proposed containment has a much smaller size than customary large dry containments in PWR practice and it can be anticipated that Nuclear Island building size will similarly be reduced.

Level 2 PSA Overview of HPR1000 Nuclear Power Plant

2020 ◽  
Author(s):  
Wang Ziguan ◽  
Lu Fang ◽  
Yang Benlin ◽  
Chen Shi ◽  
Hu Lingsheng
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Heat Removal ◽  
Accident Prevention ◽  
Severe Accident ◽  
Mitigation Measures ◽  
Injection System ◽  
Risk Level ◽  
Level 2

Abstract Risk-informed design approaches are comprehensively implemented in the design and verification process of HPR1000 nuclear power plant. Particularly, Level 2 PSA is applied in the optimization of severe accident prevention and mitigation measures to avoid the extravagant redundancy of system configurations. HPR1000 preliminary level 2 PSA practices consider internal events of the reactor in the context of at-power condition. Severe accidents mitigation and prevention system and its impact on the overall large release frequency (LRF) level are evaluated. The results showed that severe accident prevention and mitigation systems, such as fast depressurization system, the cavity injection system and the passive containment heat removal system perform well in reducing LRF and overall risk level of HPR1000 NPP. Bypass events, reactor rapture events, and the containment bottom melt-through induced by MCCI are among the dominant factors of the LRF. The level 2 PSA analysis results indicate that HPR1000 design is reliable with no major weaknesses.

Experimental Research on Characteristics of Passive Residual Heat Removal System for Small Modular Reactors Under the Station Blackout Accident

2016 ◽  
Author(s):  
Xiaodong Lu ◽  
Chuanxin Peng ◽  
Yan Zhang ◽  
Xuesong Bai ◽  
Yuanfeng Zan ◽  
...  
Keyword(s):  
Experimental Research ◽  
Heat Removal ◽  
Injection System ◽  
Primary System ◽  
Water Tank ◽  
The Core ◽  
Station Blackout ◽  
Heat Removal System ◽  
Removal System ◽  
Residual Heat

An experimental research on performance characteristics of passive residual heat removal system (PRHRS) for the small modular reactor designed by Nuclear Power Institute of China (NPIC) under the station blackout accident was performed in the CREAS facility, which consists of the primary system, the secondary system, the passive safety injection system, the passive residual heat removal system, the overpressure protection system and the auxiliary system. The experimental results show that, after the station blackout accident, a stable two-phase natural circulation between the steam generators and the heat exchanger in the PRHRS was established with a mass flow of 0.4T/h, thus the heat from the primary system was removed to the water in the containment water tank (CWT). During this period, the core decay residual heat and the sensible heat were removed from the primary system by the PRHRS effectively. The cold water from the core makeup tanks was injected into the reactor pressure vessel for core cooling. The peaked primary pressure was 16.3MPa and less than relief valve opening pressure 16.9MPa. In addition, the average coolant temperature of the reactor core reduced below 483 K, and the reactor operated safely.

Instantaneous torque, energy saving and flow rate ripple analysis of a common rail pump equipped with different delivery-pressure control systems

2017 ◽  
Vol 19 (10) ◽  
pp. 1036-1047 ◽  
Author(s):  
Alessandro Ferrari ◽  
Ruggero Vitali
Keyword(s):  
High Pressure ◽  
Energy Saving ◽  
Flow Rate ◽  
Control Strategy ◽  
Pressure Control ◽  
Common Rail ◽  
Injection System ◽  
Rail Pressure ◽  
High Pressure Pump ◽  
Rail Pressure Control

A mechanical model of a high-pressure pump of a common rail fuel injection system is presented and validated by comparison with experimental instantaneous pump shaft torque and pump piston lift data. The instantaneous torque has been measured with a high-performance torque meter installed on a hydraulic rig for testing pieces of injection apparatus. In the model, the mechanics of the piston plunger and the forces exchanged between pistons and cam are simulated, and friction losses between mobile parts are taken into account. The numerical tool is used to investigate the dynamical performance of the high-pressure pump and to analyse the impact of the rail pressure control strategy on instantaneous torque, energy saving and flow rate ripple. The rail pressure control strategy, based on the application of a fuel metering valve at the pump inlet, gives rise to an improved hydraulic efficiency of the injection system at part loads and to a moderate rate of pressure increase in the pumping chamber at part loads. However, the rail pressure control strategy based on the installation of a pressure control valve at one rail extremity leads to a reduction in the pump flow rate ripple and to a diminution in the fatigue stress. Furthermore, cavitation problems can occur during intake and early compression phases of the pump cycle when the fuel metering unit is working.

An Evaluation of a Direct Corium Cooling Method for the Ex-Vessel Melt Retention

2010 ◽  
Author(s):  
Kwang Soon Ha ◽  
Hwan Yeol Kim ◽  
Jongtae Kim ◽  
Jong Hwa Park
Keyword(s):  
Steam Explosion ◽  
Power Level ◽  
Water Injection ◽  
Conservation Equation ◽  
Severe Accident ◽  
Injection System ◽  
Steam Generation ◽  
The Core ◽  
Cooling Method ◽  
Core Catcher

An evaluation of the ex-vessel core catcher system of a sample advanced light water reactor was presented. The core catcher was designed to cool down the molten corium through a combined injection of water and gas from the bottom of the molten corium, which could be effective in the reduction of rapid steam generation and in the suppression of a steam explosion. By using the MELCOR code, a scenario analysis was performed for a representative severe accident scenario of the ALWR, that is, the 6-inches large break loss of coolant accident without safe injection. The corium spreading regime was estimated by an asymptotic calculation. The composition of the molten corium, the decay power level, and the sacrificial concrete ablation depth with time were obtained by a sacrificial concrete ablation analysis. The corium cooling history in the core catcher during the coolant injection was evaluated to calculate the temporal steam generation rate by considering an energy conservation equation. These were used as the major inputs for the temporal calculations of containment pressure which was performed by using the GASFLOW code. Several cases with change of water and gas injection rates were calculated. It was confirmed that the bottom water injection system was an effective corium cooling method in the ex-vessel core catcher to preclude a possible steam explosion and to suppress the quick release of steam.

scholarly journals Source Term Estimation under the SBLOCA-Induced Severe Accident Condition in the SMART

2021 ◽  
Vol 2021 ◽  
pp. 1-14
Author(s):  
Jaehyun Ham ◽  
Sang Ho Kim ◽  
Sung Il Kim ◽  
Byeonghee Lee ◽  
Jong-Hwa Park ◽  
...  
Keyword(s):  
Source Term ◽  
Thermal Power ◽  
Reactor Core ◽  
Heat Removal ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Injection System ◽  
Spray System ◽  
Loss Of Coolant ◽  
Accident Condition

The SMART is a system-integrated modular reactor in which a nuclear steam supply system with a thermal power of 365 MW is contained inside of the reactor vessel. Although the probability is very low, the reactor core can be damaged during a small break loss-of-coolant accident when both the passive safety injection system and the passive residual heat removal system are completely unavailable. In this work, a total of five cases were analyzed considering the reactor vessel condition and the availability of the radioactivity removal tanks and the ancillary containment spray system as containment condition variables using MELCOR code. It was estimated that there is no containment failure based on pressure, hydrogen mole fraction, and ablation depth, so that the release fractions of the 12 classes of fission products in MELCOR were evaluated considering design leak only for all cases. The overall source term of the case in which the integrity of the reactor vessel is maintained by the early initiation of the cavity flooding system was similar to that of the reactor vessel failure case. While the release fraction of cesium to the environment was analyzed to increase when there is no water in the radioactivity removal tanks, the fraction is small enough at which the radioactivity of the released cesium-137 remains well below 100 TBq, a regulatory limit. Moreover, it was found that the source term can be cut in half if the ancillary containment spray system is available. The results of this study verify the safety performance of the SMART under the small break loss-of-coolant severe accident condition with respect to the source term of interest.

Effect of Regulation Valves Installation in High Pressure Injection System Pipelines on the Formation of Reactor Vessel Thermal Stress Conditions

2014 ◽  
Author(s):  
Aleksander Mazurok ◽  
Maksym Vyshemirskyi
Keyword(s):  
High Pressure ◽  
Thermal Stress ◽  
Flow Rate ◽  
Water Flow Rate ◽  
Stress Conditions ◽  
Severe Accident ◽  
Injection System ◽  
Pressure Injection ◽  
Cooling Water Flow Rate ◽  
High Pressure Injection

Effect of regulation valves (RV) installation in high pressure injection system (HPIS) pipelines on the formation of reactor pressure vessel (RPV) thermal stress conditions was analyzed. Modernization is implemented at South-Ukrainian nuclear power plant (SUNPP) Unit 1 within the framework of life extension, which finished by the end of 2013. The main goal of the modernization is to expand the HPIS functionality for small leak accident and protection against the cold overpressurization due to flow rate and primary pressure effectively regulation. The thermal hydraulic model for RELAP5/mod3.2 code with detailed downcomer (DC) model and changes in accordance with modernization was used for calculations. Detailed (realistic) modeling of piping and equipment was performed. Also, an algorithm for the RVs was developed. Applying of cooling water flow rate regulation avoids excessive primary cooling and, consequently, helps to preserve the RPV integrity and to prevent reaching through crack formation, which can lead to a severe accident.

Hydrogen bonding in liquid methanol at ambient conditions and at high pressure

2000 ◽  
Vol 98 (3) ◽  
pp. 125-134 ◽  
Author(s):  
T. Weitkamp, J. Neuefeind, H. E. Fisch
Keyword(s):  
High Pressure ◽  
Hydrogen Bonding ◽  
Ambient Conditions
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