Relief Valve Impact Analysis

2020 ◽  
Author(s):  
Alton Reich
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Impact Analysis ◽  
High Pressures ◽  
Primary System ◽  
Relief Valve ◽  
Piston Velocity ◽  
Threaded Connection ◽  
System Pressure ◽  
The Impact

Abstract In nuclear power plants power actuated pressure relief valves serve several purposes. They act as safety valves and open automatically in response to unusually high pressures in the primary system. They also act as power operated valves and are used to relieve steam in response to automatic or manually initiated control signals. These valves are required to lift completely over a short duration from the time that they receive an actuation signal, or the system pressure exceeds the set point. This short lift time results in the valve disk moving at high velocities, and can result in high impact forces on the piston and stem when the valve fully opens. In order to evaluate and improve the performance of a two-stage power actuated relief valve, an analysis was performed to calculate the impact force on the main disk piston when it opened and the resulting stresses. The analysis was based on the main disk piston velocity measured during valve testing. Of particular interest were the stresses in the threaded connection between the stem and the main disk piston.

Dynamic Analysis of a High-Pressure Relief Valve During Opening

2020 ◽  
Vol 143 (1) ◽  
Author(s):  
John Bossard ◽  
Alton Reich ◽  
Alex DiMeo
Keyword(s):  
Power Plants ◽  
High Pressures ◽  
Experimental Testing ◽  
Pressure Relief Valve ◽  
Primary System ◽  
Relief Valve ◽  
Predictive Tool ◽  
Pressure Relief ◽  
System Pressure ◽  
Valve Operation

Abstract In nuclear power plants, power actuated pressure relief valves serve several purposes. They act as safety valves and open automatically in response to unusually high pressures in the primary system. They also act as power-operated valves and are used to relieve steam in response to automatic or manually initiated control signals. These valves are required to lift completely over a short duration from the time that they receive an actuation signal, or the system pressure exceeds the set point. This short lift time results in the valve disk moving at high velocities, and can result in high impact forces on the piston and stem when the valve fully opens. To quantitatively evaluate the dynamic performance of the Target Rock Pressure Relief Valve, an analysis effort was undertaken which would accommodate both the fluid dynamic features of the valve operation, as well as the kinematic characteristics of the valve, during pressure relief valve operation. To execute the analysis, the Generalized Fluid System Simulation Program (GFSSP) was used. GFSSP is a network flow solver computational fluid dynamics (CFD) code developed by NASA that has the ability to analyze transient, multiphase flows, and conjugate heat transfer, along with the inclusion of custom user subroutines developed by the user which can accommodate other simulation requirements. In this paper, we present the GFSSP model developed, and the computed results that could be compared with corresponding parameters as measured from experimental testing for the pressure relief valve. Adjustments to GFSSP input parameters allow the anchoring of the GFSSP valve model to test data. This makes it possible to use the GFSSP model as a predictive tool for understanding valve dynamics, as well as evaluating proposed pressure relief valve modifications for performance improvements.

scholarly journals Blowout Accident Impact Analysis Method for the Siting of Offshore Floating Nuclear Power Plant in Offshore Oil Fields

2019 ◽  
Vol 2019 ◽  
pp. 1-7
Author(s):  
Zhigang Lan
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Impact Analysis ◽  
Oil Spills ◽  
Oil Field ◽  
Oil Fields ◽  
Offshore Oil ◽  
The Impact

Focused on the utilization of nuclear energy in offshore oil fields, the correspondence between various hazards caused by blowout accidents (including associated, secondary, and derivative hazards) and the initiating events that may lead to accidents of offshore floating nuclear power plant (OFNPP) is established. The risk source, risk characteristics, risk evolution, and risk action mode of blowout accidents in offshore oil fields are summarized and analyzed. The impacts of blowout accident in offshore oil field on OFNPP are comprehensively analyzed, including injection combustion and spilled oil combustion induced by well blowout, drifting and explosion of deflagration vapor clouds formed by well blowouts, seawater pollution caused by blowout oil spills, the toxic gas cloud caused by well blowout, and the impact of mobile fire source formed by a burning oil spill on OFNPP at sea. The preliminary analysis methods and corresponding procedures are established for the impact of blowout accidents on offshore floating nuclear power plants in offshore oil fields, and a calculation example is given in order to further illustrate the methods.

Proposal for Evaluation With Strain-Based Criterion of Impact Phenomena Between Steel Structure and Tornado Missile

2016 ◽  
Author(s):  
Yuko Sakamoto ◽  
Koji Shirai ◽  
Toshiko Udagawa ◽  
Shunsuke Kondo
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Impact Analysis ◽  
Structural Integrity ◽  
Steel Structures ◽  
Severe Accident ◽  
Numerical Code ◽  
Pressurized Water ◽  
Impact Phenomena ◽  
The Impact

In Japan, nuclear power plants must be protected from tornado missiles that are prescribed by Nuclear Regular Authority (NRA). When evaluating the structural integrity of steel structures in the plant with impact analysis by numerical code, strain-based criteria are appropriate because the tornado missiles have huge impact energy and may cause large deformation of the structures. As one of the strain-based criteria, the Japan Society of Mechanical Engineers (JSME) prescribes limiting triaxial strain for severe accident of Pressurized Water Reactor (PWR) steel containment. To confirm whether or not this criterion is appropriate to the evaluation of the impact phenomena between the steel structures and the tornado missiles, a free drop impact experiment to steel plates (carbon steel and austenitic stainless steel) was carried out with heavy weights imitated on one of the tornado missiles, followed by an impact analysis of the experiment with AUTODYN code and the JSME strain-based criterion. Consequently, it was confirmed that the strain-based criterion of JSME standard was for evaluating the fracture of steel structures caused by tornado missiles.

Comparison of Analytical Methods to Study the Impact of a Postulated Closure Head Drop Accident on the Structural Integrity of the Bottom-Mounted Instrumentation System

2010 ◽  
Author(s):  
William C. Castillo ◽  
Geoffrey M. Loy ◽  
Joseph M. Remic ◽  
David P. Molitoris ◽  
George J. Demetri ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
The United States ◽  
Reactor Vessel ◽  
Primary Coolant ◽  
Heavy Load ◽  
Pressurized Water ◽  
System Pressure ◽  
The Impact

During typical nuclear power plant refueling activities for a pressurized water reactor (PWR), the reactor vessel closure head assembly must be removed from the reactor vessel (RV), transported for storage, and returned to the RV after refueling. This is categorized as a critical heavy load lift in NUREG-0612 [1] because a drop accident could result in damage to the components required to cool the fuel in the RV core. In order to mitigate the potentially severe consequences of a closure head drop, the United States Nuclear Regulatory Commission (USNRC) has mandated that nuclear power plants upgrade to a single failure-proof crane, show single failure-proof crane equivalence, or perform a head drop analysis to demonstrate that the core remains covered with coolant and sufficient cooling is available after the head drop accident. The primary coolant-retaining components associated with the RV are the inlet and outlet nozzles and the hot and cold leg main loop piping. Typical head drop analyses have considered these components to ensure that their structural integrity is maintained. One coolant-retaining component that has not been included in head drop evaluations on a consistent basis is the bottom-mounted instrumentation (BMI) system. In a typical Westinghouse PWR, 50 to 60 BMI nozzles are connected through the bottom hemisphere of the RV to one-inch diameter guide tubes which run under the vessel to a seal table above. Failure of the BMI system has the potential to adversely affect core coolability, especially if multiple failures are postulated within the system. A study was performed to compare static and dynamic methods of analyzing the effects of a head drop accident on the structural integrity of the BMI system. This paper presents the results of that study and assesses the adequacy of each method. Acceptability of the BMI system pressure boundary is based on the Nuclear Energy Institute Initiative (NEI 08–05 [2]) criteria for coolant-retaining components, which are based on Section III, Appendix F of the ASME Code [3].

Analysis of Pressure Relief Valve Dynamics During Opening

2016 ◽  
Author(s):  
John Bossard ◽  
Alton Reich ◽  
Alex DiMeo
Keyword(s):  
High Pressures ◽  
Experimental Testing ◽  
Pressure Relief Valve ◽  
Primary System ◽  
Relief Valve ◽  
Predictive Tool ◽  
Pressure Relief ◽  
System Pressure ◽  
Valve Dynamics ◽  
Valve Operation

In nuclear power plants power actuated pressure relief valves serve several purposes. They act as safety valves and open automatically in response to unusually high pressures in the primary system. They also act as power operated valves and are used to relieve steam in response to automatic or manually initiated control signals. These valves are required to lift completely over a short duration from the time that they receive an actuation signal, or the system pressure exceeds the set point. This short lift time results in the valve disk moving at high velocities, and can result in high impact forces on the piston and stem when the valve fully opens. To quantitatively evaluate the dynamic performance of the Target Rocket Pressure Relief Valve, an analysis effort was undertaken which would accommodate both the fluid dynamic features of the valve operation, as well as the kinematic characteristics of the valve, during pressure relief valve operation. To execute the analysis, the Generalized Fluid System Simulation Program (GFSSP) was used. GFSSP is a network flow solver CFD code developed by NASA that has the ability to analyze transient, multi-phase flows, and conjugate heat transfer, along with the inclusion of custom user subroutines developed by the user which can accommodate other simulation requirements. In this paper we present the GFSSP model developed, and the computed results that could be compared with corresponding parameters as measured from experimental testing for the pressure relief valve. Adjustments to GFSSP input parameters allow the anchoring of the GFSSP valve model to test data. This makes it possible to use the GFSSP model as a predictive tool for understanding valve dynamics, as well as evaluating proposed pressure relief valve modifications for performance improvements.

scholarly journals Impact Analysis of NPP Construction on National Economy Using Empower Model

2020 ◽  
Vol 22 (1) ◽  
pp. 55
Author(s):  
Suparman Suparman ◽  
Nuryanti Nuryanti ◽  
Elok Satiti Amitayani
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
National Economy ◽  
Nuclear Power ◽  
Power Plants ◽  
Government Regulation ◽  
Impact Analysis ◽  
Input Output ◽  
Construction Period ◽  
The Impact

The Nuclear Power Plant (NPP) could be one of the generation technology options to fulfill the mandate of Government Regulation No. 79 of 2014 which targeted the New Renewable Energy (NRE) portion in the national energy mix amounted to 23% by 2025 and 31% by 2050, while the realization of NRE until year 2019 is 12,6%. Any implementation of a new project or industry will have an impact on both national and region economy, and NPP project is no exception. This study aims to analyze the impact of nuclear power plant development on the national economy sector. The economic parameters analyzed in this study focused on gross domestic product (GDP) and employment. The analysis was done by using Input Output model with EMPOWER(An Extended Input-Output Model for Impact Assessment of Nuclear Power Plants) model released by IAEA as a tool. Construction period for 2 units of NPP 1000 MWe is assumed 10 years including site preparation. The results of the analysis showed that NPP construction has a significant impact on GDP and employment absorption. Each of module (A, AB, ABC and ABCD) had an impact of GDP increase of 0.021%, 0.033%, 0.040% and 0.040% respectively when compared to the GDP gained without any NPP construction. As for the amount of employment creation in module A, AB, ABC and ABCD respectively equal to 66,083, 107,693,86,081 and 85,449.It is can be concluded that according to the analysis provided by the EMPOWER, the construction of a NPP has positive impacts on the national economy.

Overview of EAF Screening Results on the 900 MWe NPP French Fleet

2016 ◽  
Author(s):  
Thomas Métais ◽  
Nicolas Robert ◽  
Pierre Genette ◽  
Nicolas Etchegaray
Keyword(s):  
Power Plants ◽  
Screening Method ◽  
Experimental Tests ◽  
Degradation Process ◽  
Primary System ◽  
Transfer Coefficients ◽  
Screening Process ◽  
The Usa ◽  
License Renewal ◽  
The Impact

In the wake of numerous experimental tests carried out in air and also in a PWR environment, both abroad and in France, an update of the current thermal fatigue codification is underway in France. Proposals are currently being integrated in the RCC-M code [1]. In parallel, it is necessary to evaluate the impact of codification evolution on the RCS components. In the USA, such evaluations have already been implemented for license renewal to operate power plants beyond their initial 40 years of operation. In order to reduce the scope of the calculations to perform, a preliminary screening was carried out on the various areas of the primary system components: this screening is detailed in an EPRI report [2]. The output of this screening process is a list of locations that are most prone to EAF degradation process and it is on these zones only that detailed EAF calculations are carried out. In France, a similar approach was defined in the perspective of the fourth ten-year visit of the 900 MWe plants (VD4 900 MWe) so as to map out all the locations that are most impacted by EAF and hence concentrate the calculation effort on these specific areas for the VD4 900 MWe. In that respect, a specific methodology to evaluate the factor to account for environmental effects or Fen [3] based on correlations [4] for hot and cold shocks was established. These correlations use data that is readily accessible in transient description documents and stress reports such as temperature change, heat transfer coefficients, ramp duration and geometry. The need for these correlations is specific to the French context due to a need for a preliminary and yet precise idea of the overall impact of the modifications brought to the RCC-M code in fatigue before the VD4 900 MWe. This paper presents the results of the screening method that was applied to the whole RCS of the 900 MWe NPP fleet.

scholarly journals A Conceptual Approach to Eliminate Bypass Release of Fission Products by In-Containment Relief Valve under SGTR Accident

2018 ◽  
Vol 2018 ◽  
pp. 1-12
Author(s):  
Taeseok Kim ◽  
Wonjun Choi ◽  
Joongoo Jeon ◽  
Nam Kyung Kim ◽  
Hoichul Jung ◽  
...  
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Fission Products ◽  
Mitigation Strategies ◽  
Severe Accident ◽  
Base Case ◽  
Conceptual Approach ◽  
Relief Valve ◽  
Radioactive Nuclides

During a hypothesized severe accident, a containment building is designed to act as a final barrier to prevent release of fission products to the environment in nuclear power plants. However, in a bypass scenario of steam generator tube rupture (SGTR), radioactive nuclides can be released to environment even if the containment is not ruptured. Thus, thorough mitigation strategies are needed to prevent such unfiltered release of the radioactive nuclides during SGTR accidents. To mitigate the consequence of the SGTR accident, this study was conducted to devise a conceptual approach of installing In-Containment Relief Valve (ICRV) from steam generator (SG) to the free space in the containment building and it was simulated by MELCOR code for numerical analysis. Simulation results show that the radioactive nuclides were not released to the environment in the ICRV case. However, the containment pressure increased more than the base case, which is a disadvantage of the ICRV. To minimize the negative effects of the ICRV, the ICRV linked to Reactor Drain Tank (RDT) and cavity flooding was performed. Because the overpressurization of containment is due to heat of ex-vessel corium, only cavity flooding was effective for depressurization. The conceptual design of the ICRV is effective in mitigating the SGTR accident.

scholarly journals Comparison and Evaluation of Domestic and Foreign Radiation Environmental Standards for Nuclear Power Plants

2021 ◽  
Vol 2083 (2) ◽  
pp. 022020
Author(s):  
Jiahuan Yu ◽  
Xiaofeng Zhang
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
The United States ◽  
Discharge System ◽  
Radioactive Effluent ◽  
Effluent Discharge ◽  
The Impact

Abstract With the development of the nuclear energy industry and the increasing demand for environmental protection, the impact of nuclear power plant radiation on the environment has gradually entered the public view. This article combs the nuclear power plant radiation environmental management systems of several countries, takes the domestic and foreign management of radioactive effluent discharge from nuclear power plants as a starting point, analyses and compares the laws and standards related to radioactive effluents from nuclear power plants in France, the United States, China, and South Korea. In this paper, the management improvement of radioactive effluent discharge system of Chinese nuclear power plants has been discussed.

Sign in / Sign up

Export Citation Format

Share Document