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.

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.

Effect of Nitrogen Release From Accumulators on PWR LOCA Analysis

2002 ◽  
Author(s):  
Woon-Shing Yeung ◽  
Ramu K. Sundaram
Keyword(s):  
Coming Out ◽  
Computer Code ◽  
Nitrogen Release ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Concentration Difference ◽  
Dissolved Nitrogen ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Release Model

The accumulator in a Pressurized Water Reactor (PWR) is generally pressurized with inert nitrogen cover gas, and the accumulator water will be saturated with nitrogen. Nitrogen released due to system depressurization during a Loss-of-Coolant Accident (LOCA) transient, consists of the nitrogen that is in the gas phase as well as nitrogen coming out of the liquid from a dissolved state. The effect of nitrogen release from the accumulator on the accident sequence is generally not explicitly addressed in typical LOCA analyses. This paper presents an analytical nitrogen release model and its incorporation into the RELAP5/MOD3 computer code. The model predicts the amount of nitrogen release as a function of concentration difference between the actual and equilibrium conditions, and can track its subsequent transport through the downstream reactor coolant system in a LOCA transient. The model is compared to data from discharge tests with a refrigerant type fluid, pressurized with nitrogen. The results demonstrate that the model is able to calculate the release of the dissolved nitrogen as designed. The modified computer code has been applied to analyze the discharge from a typical PWR accumulator. The results are compared to those obtained without the nitrogen release model. The effect of nitrogen release on major system parameters, including accumulator level, accumulator flow rate, and accumulator pressure, is discussed.

Performance of a Spray Pond for Nuclear Power Plant Ultimate Heat Sink

1977 ◽  
Vol 99 (4) ◽  
pp. 650-656
Author(s):  
V. E. Schrock ◽  
G. J. Trezek ◽  
L. R. Keilman
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Thermal Performance ◽  
Heat Sink ◽  
Nuclear Power ◽  
Field Tests ◽  
Computer Code ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Transient Thermal

Spray ponds have become an attractive method of providing the “ultimate heat sink”, i.e., the assured means of dissipating heat from a nuclear power plant. Two redundant spray ponds were the choice for this service in the Rancho Seco Nuclear Generating Station owned by Sacramento Municipal Utility District. This paper describes the results of full scale field tests of the Rancho Seco ponds which were conducted to verify the thermal performance, drift loss characteristics, and the capability to sustain the cooling requirements for a period of 30 days following a loss-of-coolant accident (LOCA). Correlations of local and average nozzle efficiency and of the drift loss are presented. A computer code was developed for the transient thermal performance of the pond. After verification the code was used to predict performance following LOCA under adverse meteorological conditions based on weather records.

Methodology for the Analysis of Fuel Behavior During LOCA and RIA for Licensing Purposes

2012 ◽  
Author(s):  
Martina Adorni ◽  
Alessandro Del Nevo ◽  
Francesco D’Auria
Keyword(s):  
Hydraulic System ◽  
Nuclear Power ◽  
Computer Code ◽  
Safety Analysis ◽  
Fuel Rod ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Fuel Pin ◽  
Computer Codes ◽  
Flow Blockage

Licensing requirements vary by country in terms of their scope, range of applicability and numerical values and may imply the use of system thermal hydraulic computer codes. Depending on the specific event scenario and on the purpose of the analysis, it might be required the availability of calculation methods that are not implemented in the standard system thermal hydraulic codes, as for burst temperature, burst strain and flow blockage calculations. This may imply the use of a dedicated fuel rod thermo-mechanical computer code, which can be coupled with thermal-hydraulic system and neutron kinetic codes to be used for the safety analysis. This paper describes the development and the application of a methodology for the analysis of the Large Break Loss of Coolant Accident (LB-LOCA) scenario in Atucha-2 Nuclear Power Plant (NPP), focusing on the procedure adopted for the use of the fuel rod thermo-mechanical code and its application for the safety analysis (Chapter 15 Final Safety Analysis Report, FSAR). The methodology implies the application of best estimate thermal-hydraulic, neutron physics and fuel pin performance computer codes, with the objective to verify the compliance with the specific acceptance criteria. The fuel pin performance code is applied with the main objective to evaluate the extent of cladding failures during the transient. A strong effort has been performed in order to enhance the fuel behaviour code capabilities and to improve the reliability of the code results.

Analysis of ABWR Containment Pressure-Temperature Loads and Suppression Pool Hydrodynamics Following Loss of Coolant Accidents

2011 ◽  
Author(s):  
P. Sawant ◽  
M. Khatib-Rahbar ◽  
F. Moody ◽  
A. Drozd
Keyword(s):  
Computer Code ◽  
Behavior Modeling ◽  
Model Parameters ◽  
Design Basis ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Design Basis Accident ◽  
Temperature Loads ◽  
Suppression Pool ◽  
Containment Pressure

This paper focuses on the assessment of Advanced Boiling Water Reactor (ABWR) containment pressure-temperature and suppression pool hydrodynamics under design-basis, loss-of-coolant accident (LOCA) conditions. The paper presents a phenomena identification and ranking table (PIRT) applicable to the ABWR containment response behavior, modeling of pressure-temperature loads using the MELCOR computer code, and analysis of suppression pool hydrodynamics parameters based on a mechanistic one-dimensional hydrodynamics model. A MELCOR 1.8.6 model with detailed nodalization of the ABWR containment is used to perform the containment pressure-temperature calculations following a design basis accident. The best estimate and several sensitivity calculations are performed for the ABWR containment using the suppression pool swell model. The sensitivity calculations demonstrate the influence of key model parameters and assumptions on the suppression pool hydrodynamics response. The comparison of containment pressure-temperature and the suppression pool swell analyses results to those reported in the ABWR licensing calculations showed reasonable agreement.

scholarly journals A proposal for a new U-D2O criticality benchmark: RB reactor core 39/1978

2012 ◽  
Vol 27 (1) ◽  
pp. 75-83
Author(s):  
Milan Pesic
Keyword(s):  
Heavy Water ◽  
Uranium Dioxide ◽  
Reactor Core ◽  
Nuclear Fission ◽  
Natural Uranium ◽  
Single Type ◽  
Uranium Metal ◽  
The Core ◽  
Enriched Uranium ◽  
Fuel Elements

In 1958, the experimental RB reactor was designed as a heavy water critical assembly with natural uranium metal rods. It was the first nuclear fission critical facility at the Boris Kidric (now Vinca) Institute of Nuclear Sciences in the former Yugoslavia. The first non-reflected, unshielded core was assembled in an aluminium tank, at a distance of around 4 m from all adjacent surfaces, so as to achieve as low as possible neutron back reflection to the core. The 2% enriched uranium metal and 80% enriched uranium dioxide (dispersed in aluminum) fuel elements (known as slugs) were obtained from the USSR in 1960 and 1976, respectively. The so-called ?clean? cores of the RB reactor were assembled from a single type of fuel elements. The ?mixed? cores of the RB reactor, assembled from two or three types of different fuel elements, were also positioned in heavy water. Both types of cores can be composed as square lattices with different pitches, covering a range of 7 cm to 24 cm. A radial heavy water reflector of various thicknesses usually surrounds the cores. Up to 2006, four sets of clean cores (44 core configurations) have been accepted as criticality benchmarks and included into the OECD ICSBEP Handbook. The RB mixed core 39/1978 was made of 31 natural uranium metal rods positioned in heavy water, in a lattice with a pitch of 8?2 cm and 78

The Development and Application of SCDAP-3D©

2002 ◽  
Author(s):  
E. W. Coryell ◽  
E. A. Harvego ◽  
L. J. Siefken
Keyword(s):  
Computer Code ◽  
Alternative Fuel ◽  
Severe Accident ◽  
Initial Development ◽  
Standard Problem ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Environmental Laboratory ◽  
Reactor Accidents ◽  
Thermal Hydraulic Analysis

The SCDAP-3D© computer code (Coryell 2001) has been developed at the Idaho National Engineering & Environmental Laboratory (INEEL) for the analysis of severe reactor accidents. A prominent feature of SCDAP-3D© relative to other versions of the code is its linkage to the state-of-the-art thermal/hydraulic analysis capabilities of RELAP5-3D©. Enhancements to the severe accident models include the ability to simulate high burnup and alternative fuel, as well as modifications to support advanced reactor analyses, such as those described by the Department of Energy’s Generation IV (GenIV) initiative. Initial development of SCDAP-3D© is complete and two widely varying but successful applications of the code are summarized. The first application is to large break loss of coolant accident analysis performed for a reactor with alternative fuel, and the second is a calculation of International Standard Problem 45 (ISP-45) or the QUENCH 6 experiment.

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