scholarly journals Investigation of Non-cooldown SG Secondary Condition on the Natural Circulation Cooling Procedure

2021 ◽  
Vol 7 (4) ◽  
pp. 1-8
Author(s):  
Eisaku TATSUMI ◽  
Wataru SAKUMA ◽  
Shinya MIYATA ◽  
Manabu MARUYAMA ◽  
Junto OGAWA
Keyword(s):  
Natural Circulation ◽  
Sensitivity Analyses ◽  
Coolant Temperature ◽  
Pressurized Water Reactor ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Coolant Pump ◽  
Dry Out ◽  
The Impact ◽  
Secondary Condition

In typical pressurized water reactor (PWR), in case that one steam generator (SG) cannot be credited for the primary cooldown, it is necessary to homogenize primary coolant temperature among loops using at least one reactor coolant pump (RCP) for the plant cooldown. If the natural circulation condition is established due to unavailability of all the RCPs, the continuous cooldown using intact SGs causes to disturb the smooth depressurization because it leads to void generation in the top of the non-cooldown SG tube where the high temperature coolant is remained. For this purpose, W.Sakuma, et al.[1] suggested the outline of asymmetric cooldown procedure without any RCPs restart. Since the suggested procedure is based on only one secondary condition (SG dry-out) of non-cooldown SG, and hence the impact of difference of the secondary condition should be investigated. In this paper, the sensitivity analyses were performed to confirm the impact on the asymmetric cooldown procedure, and consequently, it was confirmed that the coolable range used in the procedure was expanded if the water inventory exists in non-cooldown SG. Therefore it was concluded that the coolable range which was defined with the SG dry-out condition in non-cooldown SG can be conservatively applied for the operating procedure.

Feasibility Study on the Natural Circulation Cooling Procedure at the Loop Unbalanced Condition

2017 ◽  
Author(s):  
Wataru Sakuma ◽  
Shinya Miyata ◽  
Manabu Maruyama ◽  
Junto Ogawa
Keyword(s):  
Large Scale ◽  
Natural Circulation ◽  
Test Facility ◽  
Coolant Temperature ◽  
Pressurized Water Reactor ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Coolant Pump ◽  
Scale Test ◽  
Safety Enhancement

In typical pressurized water reactor (PWR) plant, in case that one steam generator (SG) is dried out and cannot be credited for the primary cooldown, at least one reactor coolant pump (RCP) has to be operated in order to homogenize the primary coolant temperature distribution among loops when the plant is cooled down to the cold shutdown state. For example, an accident such as steam line break (SLB) and feedwater line break (FLB) leads to this situation. If the natural circulation condition is established due to unavailability of all the RCPs, the natural circulation in the primary loop connected to the affected SG would be interrupted in the plant cooldown phase. In this situation, the continuous cooldown disturbs the smooth depressurization because it leads to void generation at the top of the affected SG tube where the high temperature coolant is left. In addition, there is a possibility that all RCPs cannot be operated in case of the earthquake or the fire if the RCPs are not earthquake-proof and fire-resistant. Therefore the establishment of the cooldown procedure without RCPs operation under the temperature unbalanced condition among the primary loops can contribute to the safety enhancement for typical PWR plants. The several experiments have been already performed to observe the natural circulation phenomena under the temperature unbalanced condition. It has been reported that the plant can be continuously cooled down with smooth depressurization by stepwise cooling manner using MSRVs of the intact SGs. In this study, Mitsubishi Heavy Industries, Ltd. (MHI) performed the transient analyses to simulate the natural circulation cooldown test under the temperature unbalanced condition among loops performed by Large Scale Test Facility (JAEA ROSA/LSTF) using M-RELAP5, which was a modified plant system transient code by MHI based on RELAP5-3D. Based on the analysis results, the thermal hydraulic phenomena of natural circulation cooldown under the temperature unbalanced condition were investigated. As a result, the mechanism of natural circulation interruption was clarified, and this paper shows the outline of the cooldown procedure under the temperature unbalanced condition which could be applied to the PWR plants.

Analysis of PRHRS for an Integral Pressurized Water Reactor Using RELAP5/MOD3.4

2010 ◽  
Author(s):  
Xuhua Ye ◽  
Minjun Peng ◽  
Jiange Liu
Keyword(s):  
Natural Circulation ◽  
Heat Removal ◽  
Parameter Study ◽  
Pressurized Water Reactor ◽  
Steam Generators ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Water Reactor ◽  
Pressure Peak ◽  
Residual Heat

An investigation on the thermal hydraulic characteristics of the passive residual heat removal system (PRHRS) which is used in an integral pressurized water reactor (INSURE-100) is presented in this paper. The main components of primary coolant system are enclosed in reactor vessel. Primary fluid flow circle is natural circulation. The PRHRS can remove the energy from the primary side as long as the residual heat exchanger (RHE) is submerged in the emergency cooldown tank (ECT). The parameter study is performed by considering the effects of an effective height between the steam generators and the RHE and a valve actuation time, which are useful for the design of the PRHRS. The mass flow in the PRHRS has been affected by the height difference between the steam generators and the RHE. The pressure peak of the primary side and PRHRS has been affected by the valve action time.

The Integral PWR SIR Transients: Comparisons Between CATHARE and RELAP Codes

2002 ◽  
Author(s):  
Jean-Franc¸ois Pignatel
Keyword(s):  
Natural Circulation ◽  
Atomic Energy Commission ◽  
Light Water Reactors ◽  
Pressurized Water Reactor ◽  
Large Area ◽  
Primary Coolant ◽  
Pressurized Water ◽  
The Subject ◽  
Water Reactors ◽  
Predictive Study

Within the framework of the research program on innovative light water reactors, the SERI (Service of Studies on Innovative Reactors) of the French Atomic Energy Commission (CEA), is presenting a predictive study on the modeling of a low-power integral Pressurized Water Reactor, using the CATHARE thermalhydraulic code. The concept selected for this study is that of the SIR reactor project, developed by AEA-T and ABB consortium. This very interesting concept is no doubt that which is the most complete to this date, and on which most information in the literature can be obtained. Many safety calculations made with the RELAP code are also available and represent a highly interesting base for comparison purposes, in order to improve the approach on the results obtained with CATHARE. A comparison of the behavior of the two codes is thus presented in this article. This study therefore shows that CATHARE finely models this type of new PWR concept. The transients studied cover a large area, ranging from natural circulation to loss of primary coolant accidents. The ATWS and a power transient have also been calculated. The comparison made between the CATHARE and RELAP results shows a very good agreement between the two codes, and leads to a very positive conclusion on the pertinence of simulating an integral PWR. Moreover, even though this study is a thorough investigation on the subject, it confirms the potentially safe nature of the SIR reactor.

Construction of an Elementary Model for the Dynamic Analysis of a Pressurized Water Reactor

2013 ◽  
Vol 284-287 ◽  
pp. 652-656 ◽  
Author(s):  
Chiung Wen Tsai ◽  
Chun Kuan Shih ◽  
Jong Rong Wang
Keyword(s):  
Transfer Functions ◽  
Feedback Loops ◽  
Dynamic Responses ◽  
Coolant Temperature ◽  
Inlet Temperature ◽  
Pressurized Water Reactor ◽  
Elementary Model ◽  
Pressurized Water ◽  
Water Reactor ◽  
Neutron Kinetics

A lumped-parameter numerical model was constructed based on the conservation laws of mass and energy and the point neutron kinetics with 6 groups of delayed neutron to represent the dynamics of primary loop of a pressurized water reactor (PWR) core. On the viewpoint of control theory, the coupled phenomenon of neutron kinetics and thermohydraulics can be recognized as a dynamic system with feedback loops which is caused by the Doppler effect and the coolant temperature difference. Scilab was implemented to representing the equivalent transfer functions and associated feedback loops of a PWR core. The dynamic responses were performed by the perturbations of coolant inlet flow, coolant inlet temperature, and reactivity insertion.

Development of a Thermal Fatigue Screening and Evaluation Methodology for Pressurized Water Reactor Plants

2004 ◽  
Author(s):  
J. D. Keller ◽  
A. J. Bilanin ◽  
S. T. Rosinski
Keyword(s):  
Thermal Cycling ◽  
Development Program ◽  
Fatigue Cracking ◽  
Evaluation Methodology ◽  
Experimental Investigations ◽  
Pressurized Water Reactor ◽  
Primary Coolant ◽  
Modeling Tools ◽  
Pressurized Water ◽  
Water Reactor

Thermal cycling has been identified as a mechanism that can potentially lead to fatigue cracking in un-isolable branch lines attached to pressurized water reactor (PWR) primary coolant piping. A significant research and development program has been undertaken to understand the mechanisms causing thermal cycling and to develop models for predicting the thermal-hydraulic boundary conditions for use in piping structural and fatigue analysis. A combination of first-principles engineering modeling and scaled experimental investigations has been used to formulate improved thermal cycling modeling tools. This paper will provide an overview of the model development program, a summary of the supporting test program, and a description of the thermal cycling model structure. Benchmarking of the thermal cycling model against several PWR plant configurations is presented, demonstrating favorable comparison with cases where thermal stratification and cycling has been previously observed.

scholarly journals Contribution to the study of fission products release from nuclear fuels in severe accident conditions: effect of the pO2 on Cs, Mo and Ba speciation

2020 ◽  
Vol 6 ◽  
pp. 2 ◽  
Author(s):  
Claire Le Gall ◽  
Fabienne Audubert ◽  
Jacques Léchelle ◽  
Yves Pontillon ◽  
Jean-Louis Hazemann
Keyword(s):  
Oxygen Potential ◽  
Fission Products ◽  
Severe Accident ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Mox Fuel ◽  
Chemical Behaviour ◽  
Accident Conditions ◽  
The Impact

The objective of this work is to experimentally investigate the effect of the oxygen potential on the fuel and FP chemical behaviour in conditions representative of a severe accident. More specifically, the speciation of Cs, Mo and Ba is investigated. These three highly reactive FP are among the most abundant elements produced through 235U and 239Pu thermal fission and may have a significant impact on human health and environmental contamination in case of a light water reactor severe accident. This work has set out to contribute to the following three fields: providing experimental data on Pressurized Water Reactor (PWR) MOX fuel behaviour submitted to severe accident conditions and related FP speciation; going further in the understanding of FP speciation mechanisms at different stages of a severe accident; developing a method to study volatile FP behaviour, involving the investigation of SIMFuel samples manufactured at low temperature through SPS. In this paper, a focus is made on the impact of the oxygen potential towards the interaction between irradiated MOX fuels and the cladding, the interaction between Mo and Ba under oxidizing conditions and the assessment of the oxygen potential during sintering.

Assessment of RELAP/SCDAPSIM for Turbine Trip Transient in NuScale-SMR

2018 ◽  
Author(s):  
Katarzyna Skolik ◽  
Anuj Trivedi ◽  
Marina Perez-Ferragut ◽  
Chris Allison
Keyword(s):  
Natural Circulation ◽  
System Modeling ◽  
Thermal Power ◽  
Reactor Core ◽  
Operating Conditions ◽  
Severe Accident ◽  
Circulation System ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Secondary Steam

The NuScale Small Modular Reactor (SMR) is an integrated Pressurized Water Reactor (iPWR) with the coolant flow based on the natural circulation. The reactor core consists of 37 fuel assemblies similar to those used in typical PWRs, but only half of their length to generate 160MW thermal power (50 MWe). Current study involves the development of a NuScale-SMR model based on its Design Certification Application (DCA) data (from NRC) using RELAP/SCDAPSIM. The turbine trip transient (TTT) was simulated and analysed. The objective was to assess this version of the code for natural circulation system modeling capabilities and also to verify the input model against the publicly available TTT results obtained using NRELAP5. This successful benchmark confirms the reliability of the thermal hydraulic model and allows authors to use it for further safety and severe accident analyses. The reactor core channels, pressurizer, riser and downcomer pipes as well as the secondary steam generator tubes and the containment were modeled with RELAP5 components. SCDAP core and control components were used for the fuel elements in the core. The final input deck achieved the steady state with the operating conditions comparable to those reported in the DCA. RELAP/SCDAPSIM predictions are found to be satisfactory and comparable to the reference study. It confirms the code code capabilities for natural circulation system transients.

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