Thermal Analysis of Severe Channel Damage Caused by a Stagnation Channel Break in a PHWR

2002 ◽  
Vol 124 (2) ◽  
pp. 161-167 ◽  
Author(s):  
D. Mukhopadhyay ◽  
P. Majumdar ◽  
G. Behera ◽  
S. K. Gupta ◽  
V. Venkat Raj
Keyword(s):  
Thermal Analysis ◽  
Heavy Water ◽  
Heat Generation ◽  
Temperature Rise ◽  
Heat Removal ◽  
Reactor Inlet ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Rapid Temperature ◽  
Water Reactors

The reactor channel of the horizontal core of pressurized heavy water reactors experiences very low sustained flow during loss of coolant accident (LOCA) at the reactor inlet feeders caused by certain breaks known as critical channel breaks. In this type of accident the reactor trip is delayed causing a gross mismatch of the heat generation and heat removal in the channel, thus leading to rapid temperature rise in the affected channel. A study has been carried out to identify the phenomena and the break size leading to such a situation. Severe fuel damage is predicted in the channel.

A Preliminary Safety Analysis of the Whole Assembly Seed-Blanket Heterogeneous PWR Core

2004 ◽  
Author(s):  
Jeongik Lee ◽  
Pradip Saha ◽  
Mujid S. Kazimi ◽  
Won-Jae Lee
Keyword(s):  
Temperature Rise ◽  
Safety Analysis ◽  
Complete Loss ◽  
Pressurized Water Reactors ◽  
Primary Flow ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Water Reactors ◽  
Safety Limit

The “Whole Assembly Seed and Blanket” (WASB) design, which utilizes mostly thorium in the blanket, consists of 84 seed and 109 blanket assemblies which may be backfitted into existing Pressurized Water Reactors (PWRs). Since the seed assemblies produce significantly more power than the blanket assemblies, a preliminary safety analysis of this design has been performed. Three accidents/transients (Large Break Loss of Coolant Accident (LBLOCA), Complete Loss of Primary Flow (LOPF) and Loss of Off-site Power (LOSP)), have been analyzed for both the WASB design and a typical all UO2 design for a typical 4-Loop Westinghouse PWR plant. LBLOCA results show that the peak cladding temperature (PCT) for the WASB design is approximately 260 K higher than that for a typical PWR design. However, this higher PCT for the WASB design is still about 200 K lower than the present regulatory safety limit. The response of the WASB and all UO2 core for LOPF and LOSP transients are very similar, and no post-DNB type rapid cladding temperature rise was observed in either of the two calculations.

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.

Best-Estimate Analysis of Loss-of-Coolant Accident in Pressurized Heavy Water Reactors Using Microcomputers

1992 ◽  
Author(s):  
P. Saha ◽  
T. K. Das ◽  
A. Chanda ◽  
S. Ray
Keyword(s):  
Heavy Water ◽  
Computer Software ◽  
The Other ◽  
Water Reactor ◽  
Canadian French ◽  
Best Estimate ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Heavy Water Reactor ◽  
Water Reactors

Abstract The present paper discusses the development of a computer software or code for best-estimate analysis of Loss-of-Coolant Accident (LOCA) in Pressurized Heavy Water Reactor (PHWR) systems. The formulation is comparable to U.S., Canadian, French LOCA codes, namely, TRAC, RELAPS, ATHENA, CATHARE, etc. However, the present software has been developed on Microcomputers, namely, PC-XT and AT, whereas the other softwares were developed and are being used primarily on Mainframes such as CDC-7600, CYBER-176, CRAY, etc.

Ex-Vessel Loss of Coolant Accident Analysis of ITER Divertor Cooling System Using Modified RELAP/SCADAPSIM/Mod 4.0

2017 ◽  
Vol 3 (4) ◽  
Author(s):  
S. P. Saraswat ◽  
P. Munshi ◽  
A. Khanna ◽  
C. Allison
Keyword(s):  
Heat Flux ◽  
Cooling System ◽  
Heat Removal ◽  
Hydraulic Modeling ◽  
Carbon Fiber Composites ◽  
High Heat Flux ◽  
Vacuum Vessel ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Accident Scenarios

The initial design of ITER incorporated the use of carbon fiber composites in high heat flux regions and tungsten was used for low heat flux regions. The current design includes tungsten for both these regions. The present work includes thermal hydraulic modeling and analysis of ex-vessel loss of coolant accident (LOCA) for the divertor (DIV) cooling system. The purpose of this study is to show that the new concept of full tungsten divertor is able to withstand in the accident scenarios. The code used in this study is RELAP/SCADAPSIM/MOD 4.0. A parametric study is also carried out with different in-vessel break sizes and ex-vessel break locations. The analysis discusses a number of safety concerns that may result from the accident scenarios. These concerns include vacuum vessel (VV) pressurization, divertor temperature profile, passive decay heat removal capability of structure, and pressurization of tokamak cooling water system. The results show that the pressures and temperatures are kept below design limits prescribed by ITER organization.

Analysis of Pressure Suppression Pool Swell Using Microcomputers

1992 ◽  
Author(s):  
P. Saha ◽  
B. K. Rakshit ◽  
P. Mukhopadhyay
Keyword(s):  
Operating System ◽  
Heat Transport ◽  
Transport System ◽  
Heavy Water ◽  
Computer Software ◽  
Water Reactor ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Heavy Water Reactor ◽  
Suppression Pool

Abstract The present paper discusses the development of a computer software or code for a best-estimate analysis of Pressure Suppression Pool Hydrodynamics in a Pressurized Heavy Water Reactor (PHWR) system during a Loss-of-Coolant Accident (LOCA) at the primary heat transport system. The software has been developed on Microcomputers, namely, PC-XT or AT (286) under MS-DOS operating system.

Large-Break Loss-of-Coolant Accident Phenomena Identification and Ranking Table (PIRT) for the Advanced CANDU Reactor

2004 ◽  
Author(s):  
N. Popov ◽  
H. E. Sills ◽  
V. G. Snell ◽  
B. Boyack ◽  
V. J. Langman
Keyword(s):  
Heavy Water ◽  
Computer Code ◽  
Natural Uranium ◽  
Summary Table ◽  
Regulatory Review ◽  
Loss Of Coolant Accident ◽  
Unit Energy ◽  
Wide Range ◽  
Loss Of Coolant ◽  
Enriched Uranium

The Advanced CANDU Reactor (ACR™)* is an evolutionary advancement of the current CANDU 6® reactor, aimed at producing electrical power for a capital cost and unit-energy cost significantly less than that of current reactor designs. The ACR retains the modular concept of horizontal fuel channels surrounded by heavy water moderator, as with all CANDU reactors. However, ACR uses slightly enriched uranium (SEU) fuel, compared to the natural uranium used in CANDU 6. This achieves the twin goals of improved economics (e.g., via reductions in the heavy water requirements and the use of a light water coolant), as well as improved safety. This paper is focused on the double-ended guillotine critical inlet header break (CRIHB) loss-of-coolant accident (LOCA) in an ACR reactor, which is considered as a large break LOCA. Large Break LOCA in water-cooled reactors has been used historically as a design basis event by regulators, and it has attracted a very large share of safety analysis and regulatory review. The LBLOCA event covers a wide range of system behaviours and fundamental phenomena. The Phenomena Identification and Ranking Table (PIRT) for LBLOCA therefore provides a good understanding of many of the safety characteristics of the ACR design. The paper outlines the design characteristics of the ACR reactor that impact the PIRT process and computer code applicability. It also describes the LOCA phenomena, lists all components and systems that have an important role during the event, discusses the PIRT process and results, and presents the final PIRT summary table.

scholarly journals Fuel R&D Needs and Strategy towards a Revision of Acceptance Criteria

2010 ◽  
Vol 2010 ◽  
pp. 1-7 ◽  
Author(s):  
François Barré ◽  
Claude Grandjean ◽  
Marc Petit ◽  
Jean-Claude Micaelli
Keyword(s):  
Loss Of Coolant Accident ◽  
Safety Margins ◽  
Power Load ◽  
Higher Power ◽  
Loss Of Coolant ◽  
Advanced Fuel ◽  
Water Reactors ◽  
Improved Performance ◽  
Irradiated Fuel ◽  
Cladding Material

The study of fuel behaviour under accidental conditions is a major concern in the safety analysis of the Pressurised Water Reactors. The consequences of Design Basis Accidents, such as Loss of Coolant Accident and Reactivity Initiated Accident, have to be quantified in comparison to the safety criteria. Those criteria have been established in the 1970s on the basis of experiments performed with fresh or low irradiated fuel. Starting in the 1990s, the increased industrial competition and constraints led utilities to use fuel in more and more aggressive conditions (higher discharge burnup, higher power, load follow, etc.) and create incentive conditions for the development of advanced fuel designs with improved performance (new fuel types with additives, cladding material with better resistance to corrosion, etc.). These long anticipated developments involved the need for new investigations of irradiated fuel behaviour in order to check the adequacy of the current criteria, evaluate the safety margins, provide new technical bases for modelling and allow an evolution of these criteria. Such an evolution is presently under discussion in France and several other countries, in view of a revision in the next coming years. For this purpose, a R&D strategy has been defined at IRSN.

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