Benchmarking Severe Accident Computer Codes for Heavy Water Reactor Applications

2017 ◽  
Vol 3 (2) ◽  
Author(s):  
Thambiayah Nitheanandan ◽  
X. Cao ◽  
J.-H. Choi ◽  
D. Dupleac ◽  
D.-H. Kim ◽  
...  
Keyword(s):  
Boundary Conditions ◽  
International Atomic Energy Agency ◽  
Heavy Water ◽  
Severe Accident ◽  
Energy Agency ◽  
Heavy Water Reactor ◽  
Computer Codes ◽  
Water Reactors ◽  
Definition Of ◽  
Selection Of

The International Atomic Energy Agency (IAEA) organized a coordinated research project (CRP) on “Benchmarking Severe Accident Computer Codes for Heavy Water Reactors (HWR) Applications,” (IAEA TECDOC Series No. 1727), and the activity was completed in 2012. This paper summarizes the results from the CRP: the selection of a severe accident sequence, definition of appropriate geometrical and boundary conditions, benchmarking code analyses, comparison of the code results, evaluation of the capabilities of existing computer codes to predict important severe accident phenomena, and suggestions for code improvements and/or new experiments to reduce uncertainties.

Evaluation of Dump Tank Coolability in PHWRs During Late-Phase Severe Accident

2019 ◽  
Vol 5 (4) ◽  
Author(s):  
Pradeep Pandey ◽  
Parimal P. Kulkarni ◽  
Arun Nayak ◽  
Sumit V. Prasad
Keyword(s):  
Heavy Water ◽  
Late Phase ◽  
Atomic Power Station ◽  
Severe Accident ◽  
Large Temperature ◽  
Tank Wall ◽  
Power Station ◽  
Melt Temperature ◽  
Decay Heat ◽  
Water Reactors

In some of the older design of pressurized heavy water reactors (PHWRs), such as in Rajasthan Atomic Power Station (RAPS) and Madras Atomic Power Station (MAPS), in case of a severe accident, the debris/corium may cause failure of the dump port of calandria and relocate into the dump tank. The sensible and decay heat of debris/corium makes the heavy water in dump tank to boil off leaving the dry debris in dump tank. The dry debris remelt with time and the molten corium, thus, formed has the potential to breach the dump tank and move into the containment cavity, which is highly undesirable. Hence, as an accident management strategy, water is being flooded outside the dump tank using fire water hook-up lines to remove the heat from corium to cool and stabilize it and terminate the accident progression, similar to in vessel retention. However, the question is “is the molten corium coolable by this technique.” The coolability of the molten corium in dump tank as in the reactor is assessed by conducting experiments in a scaled facility using a simulant material having comparable thermophysical properties with that of corium. Melting of dry debris resting on dump tank bottom marks the beginning of the experimental investigation for present analysis. Decay heat is simulated by a set of immersed heaters inside the melt. Temperature profiles at different locations in dump tank and in the melt pool are obtained as a function of time to demonstrate the coolability with decay heat. Large temperature gradient is observed within the corium, involving high melt center temperature and low tank wall temperature suggesting formation of crust which insulates the dump tank wall from hot corium.

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.

SCDAP/RELAP5 Analysis of Progression of Severe Accident in CANDU 6 Plant

2009 ◽  
Author(s):  
Daniel Dupleac ◽  
Mirea Mladin ◽  
Ilie Prisecaru
Keyword(s):  
Heat Sinks ◽  
Severe Accident ◽  
Light Water Reactors ◽  
Joint Project ◽  
Common Features ◽  
Severe Accidents ◽  
Computer Codes ◽  
Subsequent Loss ◽  
Water Reactors ◽  
Relap5 Code

The CANDU system design has specific features which are important to severe accidents progression and require selective consideration of models, methods and techniques of severe accident evaluation. Moreover, it should be noted that the mechanistic models for severe accident in CANDU system are largely less validated and, as a consequence, the level of uncertainty remains high in many instances. Unlike the light water reactors, for which several different computer codes to analyze severe accidents exist, for CANDU severe accidents analysis only two codes were developed: MAAP4-CANDU and ISAAC. However, both codes started by using MAAP4/PWR as reference code and implemented CANDU 6 specific models. Thus, the two codes have many common features. Recently, a joint project involving Romanian nuclear organizations and coordinated by Politehnica University of Bucharest has been started. The purpose of the project is the assessment and adaptation of the SCDAP/RELAP5 code to CANDU 6 severe accidents analysis. The present work investigates the progression of a severe accident in CANDU 6 reactor starting from a LBLOCA and a subsequent loss of all heat sinks. The paper provides details concerning the methodology and nodalisation used, and interprets the results obtained. Comparisons of the SCDAP/RELAP5 simulations with the MAAP4-CANDU and the ISAAC codes reported results are presented. Also, some insights are given on possible reasons for the discrepancies between the SCDAP/RELAP5, MAAP4-CANDU and ISAAC codes predictions.

scholarly journals POWER LEVEL EFFECTS ON THORIUM-BASED FUELS IN PRESSURE-TUBE HEAVY WATER REACTORS

2016 ◽  
Vol 5 (1) ◽  
pp. 107-119 ◽  
Author(s):  
Blair Patrick Bromley ◽  
Geoffrey W.R. Edwards ◽  
Pranavan Sambavalingam
Keyword(s):  
Heavy Water ◽  
Power Level ◽  
Pressure Tube ◽  
Specific Power ◽  
Power Flux ◽  
Heavy Water Reactor ◽  
Water Reactors ◽  
Physics Modeling ◽  
Reactor Operating ◽  
The Impact

Lattice and core physics modeling and calculations have been performed to quantify the impact of power/flux levels on the reactivity and achievable burnup for 35-element fuel bundles made with Pu/Th or U-233/Th. The fissile content in these bundles has been adjusted to produce on the order of 20 MWd/kg burnup in homogeneous cores in a 700 MWe-class pressure-tube heavy water reactor, operating on a once-through thorium cycle. Results demonstrate that the impact of the power/flux level is modest for Pu/Th fuels but significant for U-233/Th fuels. In particular, high power/flux reduces the breeding and burnup potential of U-233/Th fuels. Thus, there may be an incentive to operate reactors with U-233/Th fuels at a lower power density or to develop alternative refueling schemes that will lower the time-average specific power, thereby increasing burnup.

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.

Selection of Deformation Modes for Flexible Multibody Dynamics

1987 ◽  
Author(s):  
S.-S. Kim ◽  
E. J. Haug
Keyword(s):  
Boundary Conditions ◽  
Gaussian Elimination ◽  
Flexible Multibody Dynamics ◽  
Coefficient Matrix ◽  
Flexible Beam ◽  
Equilibrium Equations ◽  
Definition Of ◽  
Flexible Component ◽  
Deformation Modes ◽  
Selection Of

Abstract This paper presents a method of selecting boundary conditions and deformation modes for redundantly constrained flexible components in mechanical system dynamics. Gaussian elimination is used to partition the coefficient matrix in equilibrium equations for each flexible component, leading to definition of a retained statically determinate set and a redundant set of boundary conditions. Methods for selection of deformation modes is presented, to account for deformation due to constraint reaction forces. A door closing mechanism and a moving flexible beam illustrate the method of selecting boundary conditions and the effectiveness of constraint modes for approximation for system dynamic response.

scholarly journals The Feasibility of Multidimensional CFD Applied to Calandria System in the Moderator of CANDU-6 PHWR Using Commercial and Open-Source Codes

2016 ◽  
Vol 2016 ◽  
pp. 1-10 ◽  
Author(s):  
Hyoung Tae Kim ◽  
Se-Myong Chang ◽  
Jong-Hyeon Shin ◽  
Yong Gwon Kim
Keyword(s):  
Open Source ◽  
Heavy Water ◽  
Numerical Models ◽  
Three Dimensional ◽  
Hydrothermal Systems ◽  
Candu Reactors ◽  
Source Codes ◽  
Ansys Cfx ◽  
Heavy Water Reactor ◽  
Water Reactors

The moderator system of CANDU, a prototype of PHWR (pressurized heavy-water reactor), has been modeled in multidimension for the computation based on CFD (computational fluid dynamics) technique. Three CFD codes are tested in modeled hydrothermal systems of heavy-water reactors. Commercial codes, COMSOL Multiphysics and ANSYS-CFX with OpenFOAM, an open-source code, are introduced for the various simplified and practical problems. All the implemented computational codes are tested for a benchmark problem of STERN laboratory experiment with a precise modeling of tubes, compared with each other as well as the measured data and a porous model based on the experimental correlation of pressure drop. Also the effect of turbulence model is discussed for these low Reynolds number flows. As a result, they are shown to be successful for the analysis of three-dimensional numerical models related to the calandria system of CANDU reactors.

Analysis of Possibility of Pressure Tube Cold Pressurization During ECCS Injection for a Small Break LOCA

2002 ◽  
Vol 124 (4) ◽  
pp. 483-486 ◽  
Author(s):  
D. Mukhopadhyay ◽  
S. K. Gupta ◽  
V. Venkat Raj
Keyword(s):  
Control System ◽  
System Design ◽  
Heavy Water ◽  
Pressure Control ◽  
Severe Accident ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Pressure Control System ◽  
Safe Limits ◽  
Water Reactors

ECCS is designed to keep the reactor fuel temperatures within safe limits. The paper describes an additional criterion for Indian pressurized heavy water reactors (IPHWRs) evolving from the need to avoid a small break loss of cooling accident (LOCA) developing into a more severe accident. During a small break loss of coolant accident (LOCA) in PHWRs, the hydro-accumulators ride on the system and inject emergency coolant. The atmospheric steam discharge valves (ASDVs) open and cool the system due to energy discharge. In addition, the pressure control system tends to maintain the pressure. Depending on the system design, this could lead to cold pressurization of the system. This paper examines this issue.

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