Numerical Analysis of the Heat Transfer in a Pressurizer Water Reactor Core

2010 ◽  
Author(s):  
Soo W. Jo ◽  
Yong K. Lee ◽  
Jong C. Jo
Keyword(s):  
Core Temperature ◽  
Nuclear Reactors ◽  
Severe Accident ◽  
Fuel Cladding ◽  
Water Reactor ◽  
Operating Modes ◽  
Fuel Rod ◽  
The Core ◽  
Accident Management ◽  
Emergency Operating

For the initiation of emergency operating procedures and severe accident management of most commercial nuclear reactors worldwide, monitoring of the core temperature is required. Currently, it is not practicable to directly measure the temperature of fuel cladding surface temperature due to some technical limitations. Thus, measurement of the coolant temperature by using thermocouples at the core exit locations is widely used. However, the core exit temperature (CET) may not represent the core temperature properly because the measurement locations are somewhat distant from the heat generating part of fuel rod assembly. In this regard, it is important to assess the difference between the fuel cladding temperature and the CET. The objective of this study is to get the general insight and understanding of the boiling-induced multiphase flow inside fuel rod bundle during an abnormal operation mode following a loss-of coolant accident by comparing the calculation results of the CET deviation from the fuel cladding (or in-core) temperature for the two different cases of the present analysis model subjected to a coolant flowrate of either 100% or 50% of the nominal value. To do this, three-dimensional multi-phase computational fluid dynamics (CFD) calculations of a simplified pressurizer water reactor (PWR) core model were performed for both reactor operating modes. As a result, it was found that the calculated CETs are much lower than the maximum fuel rod cladding temperatures during both operating modes. Consequently, it is considered that the temperature deviation should be taken into account carefully to use the measured CETs for the initiation of emergency operating procedures and severe accident management of commercial nuclear reactors.

scholarly journals Assessment of Severe Accident Management for Small IPWR under an ESBO Scenario

2019 ◽  
Vol 2019 ◽  
pp. 1-10 ◽  
Author(s):  
Hao Yu ◽  
Minjun Peng
Keyword(s):  
High Pressure ◽  
Low Pressure ◽  
Lower Pressure ◽  
Severe Accident ◽  
Maximum Temperature ◽  
Pressurized Water ◽  
The Core ◽  
Severe Accidents ◽  
Accident Management ◽  
Degradation Studies

Interest in evaluation of severe accidents induced by extended station blackout (ESBO) has significantly increased after Fukushima. In this paper, the severe accident process under the high and low pressure induced by an ESBO for a small integrated pressurized water reactor (IPWR)-IP200 is simulated with the SCDAP/RELAP5 code. For both types of selected scenarios, the IP200 thermal hydraulic behavior and core meltdown are analyzed without operator actions. Core degradation studies firstly focus on the changes in the core water level and temperature. Then, the inhibition of natural circulation in the reactor pressure vessel (RPV) on core temperature rise is studied. In addition, the phenomena of core oxidation and hydrogen generation and the reaction mechanism of zirconium with the water and steam during core degradation are analyzed. The temperature distribution and time point of the core melting process are obtained. And the IP200 severe accident management guideline (SAMG) entry condition is determined. Finally, it is compared with other core degradation studies of large distributed reactors to discuss the influence of the inherent design characteristics of IP200. Furthermore, through the comparison of four sets of scenarios, the effects of the passive safety system (PSS) on the mitigation of severe accidents are evaluated. Detailed results show that, for the quantitative conclusions, the low coolant storage of IP200 makes the core degradation very fast. The duration from core oxidation to corium relocation in the lower-pressure scenario is 53% faster than that of in the high-pressure scenario. The maximum temperature of liquid corium in the lower-pressure scenario is 134 K higher than that of the high-pressure scenario. Besides, the core forms a molten pool 2.8 h earlier in the lower-pressure scenario. The hydrogen generated in the high-pressure scenario is higher when compared to the low-pressure scenario due to the slower degradation of the core. After the reactor reaches the SAMG entry conditions, the PSS input can effectively alleviate the accident and prevent the core from being damaged and melted. There is more time to alleviate the accident. This study is aimed at providing a reference to improve the existing IPWR SAMGs.

Ex-Vessel Coolability Analysis for a Belgian NPP

2009 ◽  
Author(s):  
Jarne R. Verpoorten ◽  
Miche`le Auglaire ◽  
Frank Bertels
Keyword(s):  
State Of The Art ◽  
The State ◽  
Severe Accident ◽  
Research Projects ◽  
Specific Data ◽  
The Core ◽  
Core Damage ◽  
Accident Management ◽  
Lower Head ◽  
Core Cooling

During a hypothetical Severe Accident (SA), core damage is to be expected due to insufficient core cooling. If the lack of core cooling persists, the degradation of the core can continue and could lead to the presence of corium in the lower plenum. There, the thermo-mechanical attack of the lower head by the corium could eventually lead to vessel failure and corium release to the reactor cavity pit. In this paper, it is described how the international state-of-the-art knowledge has been applied in combination with plant-specific data in order to obtain a custom Severe Accident Management (SAM) approach and hardware adaptations for existing NPPs. Also the interest of Tractebel Engineering in future SA research projects related to this topic will be addressed from the viewpoint of keeping the analysis up-to-date with the state-of-the art knowledge.

Fretting Wear Behavior of APMT Steel at 350°C for Reactor Fuel Cladding Application

MRS Advances ◽  
2016 ◽  
Vol 1 (35) ◽  
pp. 2495-2500
Author(s):  
Thomas Winter ◽  
James Huggins ◽  
Richard Neu ◽  
Preet Singh ◽  
Chaitanya S. Deo
Keyword(s):  
Nuclear Reactor ◽  
Wear Behavior ◽  
Sem Analysis ◽  
Fretting Wear ◽  
Pressurized Water Reactor ◽  
Fuel Cladding ◽  
Reactor Accident ◽  
Pressurized Water ◽  
Water Reactor ◽  
Fuel Rod

ABSTRACTIn support of a recent surge in research to develop an accident tolerant reactor, accident tolerant fuels and cladding candidates are being investigated. Relative motion between the fuel rods and fuel assembly spacer grids can lead to excessive fuel rod wear and, in some cases, to fuel rod failure. Based on industry data, grid-to-rod-fretting (GTRF) has been the number one cause of fuel failures within the U.S. pressurized water reactor (PWR) fleet, accounting for more than 70% of all PWR leaking fuel assemblies. APMT, an Fe-Cr-Al steel alloy, is being examined for the I2S-LWR project as a possible alternative to conventional fuel cladding in a nuclear reactor due to its favorable performance under LOCA conditions. Tests were performed to examine the reliability of the cladding candidate under simulated fretting conditions of a pressurized water reactor (PWR). The contact is simulated with a rectangular and a cylindrical specimen over a line contact area. A combination of SEM analysis and wear & work rate calculations are performed on the samples to determine their performance and wear under fretting. While APMT can perform favorably in loss of coolant accident scenarios, it also needs to perform well when compared to Zircaloy-4 with respect to fretting wear.

scholarly journals FISSION-PRODUCT RELEASES FROM A PHWR TERMINAL DEBRIS BED

2016 ◽  
Vol 5 (1) ◽  
pp. 95-105 ◽  
Author(s):  
M.J. Brown ◽  
D.G. Bailey
Keyword(s):  
Fission Product ◽  
Heavy Water ◽  
Fission Products ◽  
Severe Accident ◽  
Water Reactor ◽  
Decay Heat ◽  
The Core ◽  
Heavy Water Reactor ◽  
Release Model ◽  
Tank Water

During an unmitigated severe accident in a pressurized heavy water reactor (PHWR) with horizontal fuel channels, the core may disassemble and relocate to the bottom of the calandria vessel. The resulting heterogeneous in-vessel terminal debris bed (TDB) would likely be quenched by any remaining moderator, and some of the decay heat would be conducted through the calandria vessel shell to the surrounding reactor vault or shield tank water. As the moderator boiled off, the solid debris bed would transform into a more homogeneous molten corium pool located between top and bottom crusts. Until recently, the severe accident code MAAP-CANDU assumed that unreleased volatile and semi-volatile fission products remained in the TDB until after calandria vessel failure, due to low diffusivity through the top crust and the lack of gases or steam to flush released fission products from the debris. However, national and international experimental results indicate this assumption is unlikely; instead, high- and medium-volatility fission products would be released from a molten debris pool, and their volatility and transport should be taken into account in TDB modelling. The resulting change in the distribution of fission products within the reactor and containment, and the associated decay heat, can have significant effects upon the progression of the accident and fission-product releases to the environment. This article describes a postulated PHWR severe accident progression to generate a TDB and the effects of fission-product releases from the terminal debris, using the simple release model in the MAAP-CANDU severe accident code. It also provides insights from various experimental programs related to fission-product releases from core debris, and their applicability to the MAAP-CANDU TDB model.

The CANDLE burnup strategy applied to small modular pressurized water reactor loading with fully ceramic microencapsulated fuel

Kerntechnik ◽  
2022 ◽  
Vol 0 (0) ◽  
Author(s):  
Jinfeng Huang ◽  
Jiaming Jiang
Keyword(s):  
Equilibrium State ◽  
Power Plants ◽  
Reactor Core ◽  
Severe Accident ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
The Core ◽  
Core Height ◽  
Optimization Schemes

Abstract For post-Fukushima nuclear power plants, there has been interested in accident-tolerant fuel (ATF) since it has better tolerant in the event of a severe accident. The fully ceramic microencapsulated (FCM) fuel is one kind of the ATF materials. In this study, the small modular pressurized water reactor (PWR) loading with FCM fuels was investigated, and the modified Constant Axial shape of Neutron flux, nuclide number densities and power shape During Life of Energy producing reactor (CANDLE) burnup strategy was successfully applied to such compact reactor core. To obtain ideal CANDLE shape, it’s necessary to set the infinity or enough length of the core height, but that is impossible for small compact core setting infinity or enough length of the core height. Due to the compact and finite core, the equilibrium state can only be maintained short periods and is not obvious, other than infinitely long active core to reach the long equilibrium state for ideal CANDLE. Consequently, the modified CANDLE shape would be presented. The approximate characteristics of CANDLE burnup are observed in the finite and compact core, and the power density and fuel burnup are selected as main characteristic of modified CANDLE burnup. In this study, firstly, lots of optimization schemes were discussed, and one of optimization schemes was chosen at last to demonstrate the modified CANDLE burnup strategy. Secondly, for chosen compact small rector core, the modified CANDLE burnup strategy is applied and presented. Consequently, the new characteristics of this reactor core can be discovered both in ignition region and in fertile region. The results show that application of CANDLE burnup strategy to small modular PWR loading with FCM fuels suppresses the excess reactivity effectively and reduces the risk of small PWR reactivity-induced accidents during the whole core life, which makes the reactor control more safety and simple.

Development of Approaches to Simulate Fuel Rod Destruction With Different Fuel Type

2020 ◽  
Author(s):  
Eduard Usov ◽  
Nikolay Pribaturin ◽  
Vladimir Chukhno ◽  
Ilya Klimonov ◽  
Anton Butov ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Liquid Metals ◽  
Numerical Algorithms ◽  
Severe Accident ◽  
Theoretical Research ◽  
Detailed Calculation ◽  
Fuel Rods ◽  
Fuel Rod ◽  
The Core

Abstract Due to the revival of interest to the development of fast reactors cooled by liquid metals, the problem of carrying out theoretical research in support of their safety is actual. A detailed calculation of all stages of the accident from the beginning to the end requires knowledge of the laws for modeling physical processes occurring in the reactor in an emergency. The most serious are accidents with the destruction of the core. Simulation of severe accident in nuclear reactor is the key element in safety analysis of nuclear power plants. Destruction of fuel rods is one of the most important processes that should be calculated during core degradation. For different type of fuels the mechanism of the degradation are different too. For example, oxide and metallic fuels usually melt congruently at high temperature, but nitride fuel dissociates. The main objective of the proposed research is developing of models and numerical algorithms for calculation fuel rods destruction with oxide, metallic and nitride fuels. The models of the destruction processes and some calculation results are presented in the paper. The processes are investigated for the first phase of severe accidents covering the period from the onset of fuel-rod melting to the melt escape from the core center.

Improvement of SAMPSON Code for BWR Severe Accident Analysis

2013 ◽  
Author(s):  
Maolong Liu ◽  
Yuki Ishiwatari ◽  
Koji Okamoto
Keyword(s):  
Stainless Steel ◽  
Oxidation Reaction ◽  
Reactor Core ◽  
Severe Accident ◽  
Accident Analysis ◽  
Light Water Reactors ◽  
Fuel Cladding ◽  
Fuel Rod ◽  
Water Reactors ◽  
The Impact

The SAMPSON code has been developed in the IMPACT project in Japan to investigate severe accident phenomena for light water reactors. It integrates various analysis modules into a single code. The authors improved the fuel rod heat-up module of SAMPSON code by modeling the oxidation reaction of various core structures, including Zircaloy, stainless steel and B4C. And the creep failures of the Zircaloy fuel cladding and stainless steel monitoring guide tubes of the source range monitor (SRM) in the reactor core was also modeled for severe accident analysis.

Generalized Thermohydraulics Module GENFLO for Combining With the PWR Core Melting Model, BWR Recriticality Neutronics Model and Fuel Performance Model

2002 ◽  
Author(s):  
Jaakko Miettinen ◽  
Anitta Ha¨ma¨la¨inen ◽  
Esko Pekkarinen
Keyword(s):  
Performance Model ◽  
Severe Accident ◽  
Fuel Pellet ◽  
Fuel Performance ◽  
Basic Field ◽  
Drift Flux Model ◽  
Fuel Rod ◽  
The Core ◽  
Fraction Mixture ◽  
Accident Conditions

Thermal hydraulic simulation capability for accident conditions is needed at present in VTT in several programs. Traditional thermal hydraulic models are too heavy for simulation in the analysis tasks, where the main emphasis is the rapid neutron dynamics or the core melting. The GENFLO thermal hydraulic model has been developed at VTT for special applications in the combined codes. The basic field equations in GENFLO are for the phase mass, the mixture momentum and phase energy conservation equations. The phase separation is solved with the drift flux model. The basic variables to be solved are the pressure, void fraction, mixture velocity, gas enthalpy, liquid enthalpy, and concentration of non-condensable gas fractions. The validation of the thermohydraulic solution alone includes large break LOCA reflooding experiments and in specific for the severe accident conditions QUENCH tests. In the recriticality analysis the core neutronics is simulated with a two-dimensional transient neutronics code TWODIM. The recriticality with one rapid prompt peak is expected during a severe accident scenario, where the control rods have been melted and ECCS reflooding is started after the depressurization. The GENFLO module simulates the BWR thermohydraulics in this application. The core melting module has been developed for the real time operator training by using the APROS engineering simulators. The core heatup, oxidation, metal and fuel pellet relocation and corium pool formation into the lower plenum are calculated. In this application the GENFLO model simulates the PWR vessel thermohydraulics. In the fuel performance analysis the fuel rod transient behavior is simulated with the FRAPTRAN code. GENFLO simulates the subchannel around a single fuel rod and delivers the heat transfer on the cladding surface for the FRAPTRAN. The transient boundary conditions for the subchannel are transmitted from the system code for operational transient, loss of coolant accidents and anticipated transients without scram. Experiences with the GENFLO code have pointed out that the efforts needed for the code development and its validation can be minimized with this kind of the generalized solution.

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