scholarly journals A comparative study of kinetics of nuclear reactors

2009 ◽  
Vol 24 (3) ◽  
pp. 167-176 ◽  
Author(s):  
Khalilurrahman Obaidurrahman ◽  
Om Singh
Keyword(s):  
Heavy Water ◽  
Computer Code ◽  
Accurate Method ◽  
Operating Conditions ◽  
Pressurized Water ◽  
Safe Limits ◽  
Water Reactors ◽  
Breeder Reactors ◽  
Point Kinetics ◽  
Reactivity Insertion

The paper deals with the study of reactivity initiated transients to investigate major differences in the kinetics behavior of various reactor systems under different operating conditions. The article also states guidelines to determine the safety limits on reactivity insertion rates. Three systems, light water reactors (pressurized water reactors), heavy water reactors (pressurized heavy water reactors), and fast breeder reactors are considered for the sake of analysis. The upper safe limits for reactivity insertion rate in these reactor systems are determined. The analyses of transients are performed by a point kinetics computer code, PKOK. A simple but accurate method for accounting total reactivity feedback in kinetics calculations is suggested and used. Parameters governing the kinetics behavior of the core are studied under different core states. A few guidelines are discussed to project the possible kinetics trends in the next generation reactors.

Effects of Weld Overlays on Leak-Before-Break Margins

2011 ◽  
Author(s):  
G. Angah Miessi ◽  
Peter C. Riccardella ◽  
Peihua Jing
Keyword(s):  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Corrosion Cracking ◽  
Operating Conditions ◽  
Intergranular Stress Corrosion ◽  
Pressurized Water ◽  
Weld Overlay ◽  
Water Reactors ◽  
Weld Overlays ◽  
Leak Before Break

Weld overlays have been used to remedy intergranular stress corrosion cracking (IGSCC) in boiling water reactors (BWRs) since the 1980s. Overlays have also been applied in the last few years in pressurized water reactors (PWRs) where primary water stress corrosion cracking (PWSCC) has developed. The weld overlay provides a structural reinforcement with SCC resistant material and favorable residual stresses at the ID of the overlaid component. Leak-before-break (LBB) had been applied to several piping systems in PWRs prior to recognizing the PWSCC susceptibility of Alloy 82/182 welds. The application of the weld overlay changes the geometric configuration of the component and as such, the original LBB evaluation is updated to reflect the new configuration at the susceptible weld. This paper describes a generic leak-before-break (LBB) analysis program which demonstrates that the application of weld overlays always improves LBB margins, relative to un-overlaid, PWSCC susceptible welds when all the other parameters or variables of the analyses (loads, geometry, operating conditions, analysis method, etc…) are kept equal. Analyses are performed using LBB methodology previously approved by the US NRC for weld overlaid components. The analyses are performed for a range of nozzle sizes (from 6″ to 34″) spanning the nominal pipe sizes to which LBB has been commonly applied, using associated representative loads and operating conditions. The analyses are performed for both overlaid and un-overlaid configurations of the same nozzles, and using both fatigue and PWSCC crack morphologies in the leakage rate calculations and the LBB margins are compared to show the benefit of the weld overlays.

AP1000 Plant CILC/CIPS Risk Assessment Using Advanced TH Methodology

2012 ◽  
Author(s):  
G. Wang ◽  
P. Sapienza ◽  
R. J. Fetterman ◽  
M. Y. Young ◽  
J. R. Secker ◽  
...  
Keyword(s):  
Risk Assessment ◽  
Nucleate Boiling ◽  
Operating Conditions ◽  
Localized Corrosion ◽  
Pressurized Water ◽  
Assembly Design ◽  
Power Shift ◽  
Fuel Assemblies ◽  
Water Reactors ◽  
The Relationship

Similar to many existing Pressurized Water Reactors (PWR), the AP1000® cores will undergo sub-cooled nucleate boiling in the upper grid spans of some fuel assemblies at normal operating conditions. Sub-cooled nucleate boiling may increase crud deposits on the fuel cladding surface which may increase the risk of Crud Induced Power Shift (CIPS) and/or Crud Induced Localized Corrosion (CILC). A CIPS/CILC risk assessment has been performed to support the AP1000 fuel assembly design finalization. In this paper, the advanced thermal-hydraulic (TH) methodology used in the AP1000 plant CIPS/CILC risk assessments are summarized and discussed, and the relationship between the CIPS/CILC mechanisms, fuel reliability, and plant operating conditions is also presented. Finally, acceptable AP1000 core CIPS/CILC risk assessment results are summarized and suggestions that specifically target reducing CIPS/CILC risks for AP1000 plants are described.

Parallel Computation of the Point Neutron Kinetic Equations Using Parallel Revisionist Integral Deferred Correction

2018 ◽  
Author(s):  
Yun Cai ◽  
Xingjie Peng ◽  
Qing Li ◽  
Zhizhu Zhang ◽  
Zhumin Jiang ◽  
...  
Keyword(s):  
Time Integration ◽  
Kinetic Equations ◽  
Accurate Method ◽  
Small Time ◽  
Time Step ◽  
Deferred Correction ◽  
Time Integration Method ◽  
Speed Up ◽  
Point Kinetics ◽  
Reactivity Insertion

The point kinetics is very important to the safety of the reactor operation. However, these equations are stiff and usually solved with very small time step. These equations are solved by Revisionist integral deferred correction (RIDC), which is a parallel time integration method. RIDC is a highly accurate method, and it reduces the error by iteration. Based on C++ and MPI, a four-core fourth-order RIDC is implemented and tested by several cases, such as step, ramp, and sinusoidal reactivity insertion. Compared with other methods, the time step of RIDC in the step reactivity insertion case is smaller, but it’s larger in the case of the sinusoidal reactivity insertion. RIDC can keep high accuracy while the time step is appropriately large. The numerical results also show that the speed-up ratio can achieve 2 when 4 processors are used.

scholarly journals Nuclear viewpoint in India

2017 ◽  
Vol 4 ◽  
Author(s):  
Anshu Bharadwaj ◽  
Lakshminarayana Venkat Krishnan ◽  
Subramaniam Rajagopal
Keyword(s):  
Heavy Water ◽  
Nuclear Power ◽  
Clean Energy ◽  
Pressurized Water ◽  
Heavy Water Reactor ◽  
Water Reactors ◽  
Near Term ◽  
Enriched Uranium ◽  
Installed Capacity

ABSTRACTNuclear power is a crucial source of clean energy for India. In the near-term, India is focusing on thermal reactors using natural and enriched uranium. In the long-term, India is exploring various options to use its large thorium reserves.India’s present nuclear installed capacity is 5680 MW, which contributes to about 3.4% of the annual electricity generation. However, nuclear power is an important source of energy in India’s aspirations for energy security and also in achieving its Intended Nationally Determined Contributions (INDC), of 40% fossil free electricity, by 2030. India has limited uranium reserves, but abundant thorium reserves. The Nuclear Suppliers Group (NSG) lifted restrictions on trade with India, in 2008, enabling India to import uranium (natural and enriched) and nuclear reactors. In the near–term (2030), the nuclear capacity could increase to about 42,000 MW. This would be from a combination of domestic Pressurized Heavy Water Reactors (PHWR) and imported Pressurized Water Reactors (PWR). For the long–term (2050), India is exploring various options for utilising its vast thorium reserves. This includes Advanced Heavy Water Reactor and Molten Salt Breeder Reactor. However, generating public acceptance will be crucial to the expansion of the nuclear power program.

Multiphase Flow Simulation of Subcooled Boiling Using the N-Phase Approach in TransAT

2013 ◽  
Author(s):  
S. Thomas ◽  
C. Narayanan ◽  
D. Lakehal ◽  
Y. Gong
Keyword(s):  
Phase Change ◽  
Flow Simulation ◽  
Operating Conditions ◽  
Navier Stokes ◽  
Operating Pressure ◽  
Subcooled Boiling ◽  
Convective Boiling ◽  
Pressurized Water ◽  
Recent Developments ◽  
Water Reactors

In Pressurized Water Reactors (PWR), convective boiling occurs at the heated walls, which are superheated while liquid bulk is subcooled at a given operating pressure. The physical modeling of the phenomenon is complex, as it requires consideration of phase change and turbulence. The combinations of Reynolds-Averaged Navier Stokes (RANS) and wall boiling models have found moderate success in the past. Succesful modeling depends on the nature of the problem and its operating conditions, the manner of the model implementation and the quality of the computational platform. In the present work recent developments executed in the TransAT code for subcooled boiling are demonstrated. To better capture the phase change, the modified mixture approach is employed where the temperature of the N-phases are resolved separately. This method has been more suitable for phase-change problems, when compared to a single mixture temperature. To check the validity of the modeling, results are evaluated against the DEBORA experiments carried out at CEA Grenoble.

Computational Analysis of Transient 3D-Melt-Water Interactions

2000 ◽  
Author(s):  
Nikolay Ivanov Kolev
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Computational Analysis ◽  
Computer Code ◽  
3D Analysis ◽  
Damage Potential ◽  
Pressurized Water ◽  
Modeling Technology ◽  
Melt Water ◽  
Water Reactors

Abstract This work is part of the Siemens effort to estimate the damage potential of melt water interaction in future nuclear power plants with pressurized water reactors. After creating a modeling technology, the IVA5 computer code, verifying it by comparison with numerous separated effect tests, system tests and analytical benchmarks, performing many 2D computational analysis we present in this work complete 3D analysis of melt water interactions. Interesting conclusions for the systems analyzed are drown. Some of the limitations of the technology are also demonstrated.

scholarly journals Study of Fast Transient Pressure Drop in VVER-1000 Nuclear Reactor Using Acoustic Phenomenon

2018 ◽  
Vol 2018 ◽  
pp. 1-11
Author(s):  
Soroush Heidari Sangestani ◽  
Mohammad Rahgoshay ◽  
Naser Vosoughi ◽  
Mitra Athari Allaf
Keyword(s):  
Pressure Drop ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Computer Code ◽  
Accurate Method ◽  
Transient Simulation ◽  
Transient Pressure ◽  
Pressurized Water ◽  
Thermal Feedback ◽  
Acoustic Phenomenon

This article aims to simulate the sudden and fast pressure drop of VVER-1000 reactor core coolant, regarding acoustic phenomenon. It is used to acquire a more accurate method in order to simulate the various accidents of reactor core. Neutronic equations should be solved concurrently by means of DRAGON 4 and DONJON 4 coupling codes. The results of the developed package are compared with WIMS/CITATION and final safety analysis report of Bushehr VVER-1000 reactor (FSAR). Afterwards, time dependent thermal-hydraulic equations are answered by employing Single Heated Channel by Sectionalized Compressible Fluid method. Then, the obtained results were validated by the same transient simulation in a pressurized water reactor core. Then, thermal-hydraulic and neutronic modules are coupled concurrently by use of producing group constants regarding the thermal feedback effect. Results were compared to the mentioned transient simulation in RELAP5 computer code, which show that mass flux drop is sensed at the end of channel in several milliseconds which causes heat flux drop too. The thermal feedback resulted in production of some perturbations in the changes of these parameters. The achieved results for this very fast pressure drop represent accurate calculations of thermoneutronic parameters fast changes.

Insights Into Operating a Low Pressure Feed Pump for Loss of All AC and DC Power

2008 ◽  
Author(s):  
Barbara R. Baron ◽  
Robert J. Lutz
Keyword(s):  
Operating Experience ◽  
Heat Removal ◽  
Computer Code ◽  
Low Pressure ◽  
Feed Pump ◽  
Pressurized Water ◽  
Local Flow ◽  
Dc Power ◽  
Technical Issues ◽  
Water Reactors

Generic guidance for Pressurized Water Reactors (PWRs) has been developed to address the beyond design basis event of coincident loss of all Alternating Current (AC) and Direct Current (DC) power. The generic guidance included a strategy to use a low pressure feed pump to provide adequate secondary side heat removal via the Steam Generators (SGs) to delay or prevent core uncovery following loss of all AC power with battery depletion, loss of all DC power, seismic initiated events, and/or terrorist initiated events. The purpose of the project was to use thermal hydraulic analyses, operating experience, and other engineering analyses to identify and evaluate technical issues associated with the implementation of the low pressure feed pump strategy at Westinghouse and Combustion Engineering (CE) designed plants. The technical issues that were evaluated are those issues typically addressed in the development of a plant’s Emergency Operating Procedures (EOPs) and Off-Normal Operating Procedures (ONOPs). The thermal hydraulic analyses were performed using the computer code MAAP 4.0.5 and a plant model of a 4-loop Westinghouse designed PWR. The results of the analyses are also applicable to 2-Loop and 3-Loop Westinghouse and CE designed PWRs. The results of the evaluation indicated that the key technical issue potentially impacting the prevention of core uncovery for the implementation of the low pressure feed pump strategy is the potential and consequences of injecting nitrogen into the Reactor Coolant System (RCS) from the cold leg accumulators/Safety Injection Tanks (SITs). The results of the evaluation were used to develop sample instructions for implementing the low pressure feed pump strategy for Westinghouse and CE designed PWRs. The sample instructions were developed for two categories of low pressure feed pumps: (1) low pressure feed pumps with shutoff heads greater than the pressure that nitrogen injects into the RCS and (2) low pressure feed pumps with shutoff heads less than the pressure that nitrogen injects into the RCS. The usefulness of the sample instructions is maximized when the low pressure feed pump has local flow indication, throttling capabilities, deadhead protection, and local SG pressure indication is available. The results demonstrated that the design characteristics of the low pressure feed pump are important to prolonging/preventing the time of core uncovery.

Scaling Assessment for the Integral Effect Tests: 2 — Open System Depressurization

2013 ◽  
Author(s):  
Xiaoyu Cai
Keyword(s):  
Open System ◽  
Computer Code ◽  
Passive Safety ◽  
Prototype System ◽  
Integral Test ◽  
Pressurized Water ◽  
Passive Safety Systems ◽  
Loss Of Coolant Accident ◽  
Water Reactors ◽  
Westinghouse Electric

In this paper a method is described for using NOTRUMP models to corroborate the Hierarchical Two-Tiered Scaling (H2TS) methodology that has been used for design of the APEX and SPES-2 test facilities. These facilities were built for the Westinghouse Electric Corporation to obtain data on the performance of the passive safety systems of the advanced pressurized water reactors. Similarity between the prototype system and the scaled test facilities is investigated for the open system depressurization phenomena in the postulated small break loss of coolant accident transient. The objective of this analysis is to provide a basis that an integral test for the passive safety system will provide valid experimental data for the high-ranked phenomena that may occur during the hypothetical SBLOCA transients. In this way, the experiment will capture the phenomena that would be expected to occur in the plant, and hence provide data that can be used to validate computer code.

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