scholarly journals Control Rod Modeling and Worth Calculation for a Typical 1100 MWe Nuclear Power Plant Using WIMS/D4 and CITATION

2022 ◽  
Vol 2022 ◽  
pp. 1-13
Author(s):  
Izza Shahid ◽  
Nadeem Shaukat ◽  
Amjad Ali ◽  
Meer Bacha ◽  
Ammar Ahmad ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Power Distribution ◽  
Nuclear Power ◽  
Operating Conditions ◽  
Pressurized Water ◽  
Control Rod ◽  
Operational Conditions ◽  
The Core ◽  
Safety Parameters

A typical 1100 MWe pressurized water reactor (PWR) is a second unit installed at the coastal site of Pakistan. In this paper, verification analysis of reactivity control worth by means of rod cluster control assemblies (RCCAs) for startup and operational conditions of this typical nuclear power plant (CNPP) has been performed. Neutronics analysis of fresh core is carried out at beginning of life (BOL) to determine the effect of grey and black control rod clusters on the core reactivity for startup and operating conditions. The combination of WIMS/D4 and CITATION computer codes equipped with JENDL-3.3 data library is used for the first time for core physics calculations of neutronic safety parameters. The differential and integral worth of control banks is derived from the computed results. The effect of control bank clusters on core radial power distribution is studied precisely. Radial power distribution in the core is evaluated for numerous configurations of control banks fully inserted and withdrawn. The accuracy of computed results is validated against the reference values of Nuclear Design Report (NDR) of 1100 MWe typical CNPP. It has been observed that WIMS-D4/CITATION shows its capability to effectively calculate the reactor physics parameters.

A New Safety Analysis Method of Control Rod Ejection Accident in PWR NPP Based on the Failure of Causal Relationship

2018 ◽  
Author(s):  
Shiyu Yan ◽  
Hua Liu ◽  
Zhaohui Liu ◽  
Xiaohua Yang ◽  
Meng Li ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Safety Analysis ◽  
Pressurized Water ◽  
Control Rod ◽  
Drive Mechanism ◽  
Cause And Effect ◽  
Simulation Calculation ◽  
Causal Logic

In view of control rod ejection accident of the traditional pressurized water reactor, the safety thought of the design phase is to validate reliability and availability of DCS I&C in the severe accidents. Now the most important and effective means is simulation calculation and analysis. It is applied for the imaginary accident of the nuclear power plant by using computer software. The new safety analysis steps based on the analysis of cause-and-effect logic failure: firstly, the composition and working principle of control rod drive mechanism is analyzed; secondly, a list of factors-the dynamics and structure, environmental reasons, the function of the control rod drive mechanism and status analysis-are all taken into account, the initial cause of failure modes with causal logic analysis is carried out; thirdly, based on cause-and-effect logic failure, the prevention and improvement measures of accidents, the new criterion of design are put forward. The advantages of cause-and-effect logic failure safety analysis: 1.be based on causal logic. 2. the system aspects is added compared with the past method that is only based on simulation calculation and analysis of the hypothetical accident, the accident the transient process of the key security parameters as the acceptance criteria. 3. The verification and audit of the lack of safety design criteria, completeness of design content, sufficiency problem are performed before the simulated calculation and analysis. 4. The coverage of safety analysis is expanded. Some good advices are provided for the design, operation and maintenance of nuclear power plant.

14C Distribution in Atmospheric and Aquatic Environments Around Qinshan Nuclear Power Plant, China

Radiocarbon ◽  
2014 ◽  
Vol 56 (3) ◽  
pp. 1107-1114 ◽  
Author(s):  
Zhongtang Wang ◽  
Dan Hu ◽  
Hong Xu ◽  
Qiuju Guo
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Water Samples ◽  
Nuclear Power ◽  
Dissolved Inorganic Carbon ◽  
Tap Water ◽  
Surface Seawater ◽  
Pressurized Water ◽  
Specific Activities ◽  
Water Reactors

Atmospheric CO2 and aquatic water samples were analyzed to evaluate the environmental 14C enrichment due to operation of the Qinshan nuclear power plant (NPP), where two heavy-water reactors and five pressurized-water reactors are employed. Elevated 14C-specific activities (2–26.7 Bq/kg C) were observed in the short-term air samples collected within a 5-km radius, while samples over 5 km were close to background levels. The 14C-specific activities of dissolved inorganic carbon (DIC) in the surface seawater samples ranged from 196.8 to 206.5 Bq/kg C (average 203.4 Bq/kg C), which are close to the background value. No elevated 14C level in surface seawater was found after 20 years of operation of Qinshan NPP, indicating that the 14C discharged was well diffused. The results of the freshwater samples show that excess 14C-specific activity (average 17.1 Bq/kg C) was found in surface water and well water samples, while no obvious 14C increase was found in drinking water (groundwater and tap water) compared to the background level.

scholarly journals 14C Release in Various Chemical Forms With Gaseous Effluents From the Paks Nuclear Power Plant

Radiocarbon ◽  
1989 ◽  
Vol 31 (03) ◽  
pp. 754-761 ◽  
Author(s):  
Ede Hertelendi ◽  
György Uchrin ◽  
Peter Ormai
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Liquid Scintillation ◽  
Mean Value ◽  
Scintillation Counting ◽  
Pressurized Water ◽  
The Mean ◽  
Water Reactors

We present results of airborne 14C emission measurements from the Paks PWR nuclear power plant. Long-term release of 14C in the form of carbon dioxide or carbon monoxide and hydrocarbons were simultaneously measured. The results of internal gas-proportional and liquid scintillation counting agree well with theoretical assessments of 14C releases from pressurized water reactors. The mean value of the 14C concentration in discharged air is 130Bqm-3 and the normalized release is equal to 740GBq/GWe · yr. > 95% of 14C released is in the form of hydrocarbons, ca 4% is apportioned to CO2, and <1% to CO. Tree-ring measurements were also made and indicated a minute increase of 14C content in the vicinity of the nuclear power plant.

scholarly journals Neutronic safety analysis of proposed reactor technologies for Ghana’s nuclear power plant using the MCNP code

2019 ◽  
Vol 34 (3) ◽  
pp. 238-242
Author(s):  
Rex Abrefah ◽  
Prince Atsu ◽  
Robert Sogbadji
Keyword(s):  
High Pressure ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Atomic Energy Commission ◽  
Clean Energy ◽  
Safety Analysis ◽  
Operating Conditions ◽  
The Government

In pursuance of sufficient, stable and clean energy to solve the ever-looming power crisis in Ghana, the Nuclear Power Institute of the Ghana Atomic Energy Commission has on the agenda to advise the government on the nuclear power to include in the country's energy mix. After consideration of several proposed nuclear reactor technologies, the Nuclear Power Institute considered a high pressure reactor or vodo-vodyanoi energetichesky reactor as the nuclear power technologies for Ghana's first nuclear power plant. As part of technology assessments, neutronic safety parameters of both reactors are investigated. The MCNP neutronic code was employed as a computational tool to analyze the reactivity temperature coefficients, moderator void coefficient, criticality and neutron behavior at various operating conditions. The high pressure reactor which is still under construction and theoretical safety analysis, showed good inherent safety features which are comparable to the already existing European pressurized reactor technology.

On the Feasibility of Nuclear Versus Combustion Gas Turbine Propulsion for High-Speed Cargo Ships

2004 ◽  
Author(s):  
H. Boonstra ◽  
A. C. Groot ◽  
C. A. Prins
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Gas Turbine ◽  
High Speed ◽  
Nuclear Power ◽  
Parametric Design ◽  
Pebble Bed ◽  
Pressurized Water ◽  
Combustion Gas ◽  
The Impact

This paper presents the outcome of a study on the feasibility of a nuclear powered High-Speed Pentamaran, initiated by Nigel Gee and Associates and the Delft University of Technology. It explores the competitiveness of a nuclear power plant for the critical characteristics of a marine propulsion plant. Three nuclear reactor types are selected: the Pressurized Water Reactor (PWR), the Pebble-bed and Prismatic-block HTGR. Their characteristics are estimated for a power range from 100 MWth to 1000 MWth in a parametric design, providing a level base for comparison with conventional gas turbine technology. The reactor scaling is based on reference reactors with an emphasis on marine application. This implies that preference is given to passive safety and simplicity, as they are key-factors for a marine power plant. A case study for a 60-knot Pentamaran shows the impact of a nuclear power plant on a ship designed with combustion gas turbine propulsion. The Prismatic-block HTGR is chosen as most suitable because of its low weight compared to the PWR, in spite of the proven technology of a PWR. The Pebble-bed HTGR is considered too voluminous for High-Speed craft. Conservative data and priority to simple systems and high safety leads to an unfavorable high weight of the nuclear plant in competition with the original gas turbine driven Pentamaran. The nuclear powered ship has some clear advantages at high sailing ranges.

Centrifugal Pump Leak-Free Technology Introduction and Its Applications in EPR Nuclear Power Plant

2010 ◽  
Author(s):  
Guohui Cong ◽  
Ling Zhang
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Centrifugal Pump ◽  
Nuclear Power ◽  
Operating Conditions ◽  
Critical Conditions ◽  
Mechanical Seal ◽  
Centrifugal Pumps ◽  
Canned Motor ◽  
Magnetic Drive

Environmental protection requirement is more and more critical now, and it increases the request to prevent dangerous liquid to leak outside in nuclear power plant too. Centrifugal pumps are the most important active equipments in nuclear power plant, but there is a shaft clearance between rotor and stator of centrifugal pump. The shaft clearance can lead pumped fluid to the outside, so the environment may be polluted by the leakage. In some critical conditions such as transferring high radioactive fluid in the pump, the leakage shall be totally forbidden. So solutions have to be found to make centrifugal pumps totally leak-free for applications in nuclear power plant. Normally there are three leak-free technologies for centrifugal pumps: mechanical seal with auxiliary system, canned motor and magnetic drive. In this paper, all the three leak-free technologies and some of their applications in EPR 3rd generation PWR nuclear power plants are presented and discussed. The results show that in EPR nuclear power plant, canned motor pumps can be preferably used for strict environmental requirement of leak-free if the pump power and operating conditions are applicable. For other conditions, pumps with double mechanical seal can also be used with additional sealing water system support. For centrifugal pumps with magnetic drive are not so applicable in high pressure condition, and the safety aspect is weaker than canned motor pumps, generally they are not used in EPR nuclear power plant at present.

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