Steady-State Safety Analysis of Ghana Research Reactor-1 With Low-Enriched-Uranium Core

2020 ◽  
Vol 12 (5) ◽  
Author(s):  
Edward Shitsi ◽  
Prince Amoah ◽  
Emmanuel Ampomah-Amoako ◽  
Henry Cecil Odoi
Keyword(s):  
Steady State ◽  
Research Reactor ◽  
Flow Instability ◽  
Argonne National Laboratory ◽  
Safety Analysis ◽  
Nucleate Boiling ◽  
Pressurized Water ◽  
Safety Margins ◽  
Water Reactors ◽  
Enriched Uranium

Abstract Research reactors all over the world are expected to operate within certain safety margins just like pressurized water reactors and boiling water reactors. These safety margins mainly include onset of nucleate boiling ratio (ONBR), departure from nucleate boiling ratio (DNBR), and flow instability ratio (FIR) in addition to the maximum clad or fuel temperature and saturation temperature or boing point of the coolant inside the core of the reactor. This study carried out steady-state safety analysis of the Ghana Research Reactor-1 (GHARR-1) with low enriched uranium (LEU) core. Monte Carlo N-particle (MCNP) code was used to obtain radial and axial power peaking factors used as inputs in the preparation of the input file of plate temperature code of Argonne National Laboratory (PLTEMP/ANL code), which was then used to obtain the mentioned safety parameters of GHARR-1 with LEU core in this study. The data obtained on the ONBR were used to obtain the initiation of nucleate boiling boundary data with respect to the active length of the reactor core for various reactor powers. The obtained results for LEU core were also compared with that of the high enriched uranium (HEU) core. The results obtained show that the 34 kW GHARR-1 with LEU core is safe to operate just as the previous 30 kW HEU core was safe to operate.

Modeling of Porous Crud Layer Based on Effects of Diffusion and Water Radiolysis on Anomalous Porous Crud Deposition on Fuel Pin Surfaces in PWRs

2002 ◽  
Author(s):  
Hyun-Jong Joe ◽  
Barclay G. Jones
Keyword(s):  
Negative Impact ◽  
Nucleate Boiling ◽  
Water Radiolysis ◽  
Trace Impurities ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Fuel Pin ◽  
Negative Effect ◽  
Water Reactors ◽  
Coolant System

Many studies have been undertaken to understand crud formation on the upper spans of fuel pin clad surfaces, which is called axial offset anomaly (AOA), is observed in pressurized water reactors (PWR) as a result of sub-cooled nucleate boiling. Separately, researchers have considered the effect of water radiolysis in the primary coolant of PWR. This study examines the effects of radiolysis of liquid water, which aggressively participate in general cladding corrosion and solutes within the primary coolant system, in the terms of pH, temperature, and Linear Energy Transfer (LET). It also discusses the effect of mass transfer, especially diffusion, on the concentration distribution of the radiolytic products, H2 and O2, in the porous crud layer. Finally it covers the effects of chemical reactions of boric acid (H3BO3), which has a negative impact on the mechanisms of water recombination with hydrogen, lithium hydroxide (LiOH), which has a negative effect on water decomposition, dissolved hydrogen (DH), and some trace impurities.

AP1000 Plant CILC/CIPS Risk Assessment Using Advanced TH Methodology

2012 ◽  
Author(s):  
G. Wang ◽  
P. Sapienza ◽  
R. J. Fetterman ◽  
M. Y. Young ◽  
J. R. Secker ◽  
...  
Keyword(s):  
Risk Assessment ◽  
Nucleate Boiling ◽  
Operating Conditions ◽  
Localized Corrosion ◽  
Pressurized Water ◽  
Assembly Design ◽  
Power Shift ◽  
Fuel Assemblies ◽  
Water Reactors ◽  
The Relationship

Similar to many existing Pressurized Water Reactors (PWR), the AP1000® cores will undergo sub-cooled nucleate boiling in the upper grid spans of some fuel assemblies at normal operating conditions. Sub-cooled nucleate boiling may increase crud deposits on the fuel cladding surface which may increase the risk of Crud Induced Power Shift (CIPS) and/or Crud Induced Localized Corrosion (CILC). A CIPS/CILC risk assessment has been performed to support the AP1000 fuel assembly design finalization. In this paper, the advanced thermal-hydraulic (TH) methodology used in the AP1000 plant CIPS/CILC risk assessments are summarized and discussed, and the relationship between the CIPS/CILC mechanisms, fuel reliability, and plant operating conditions is also presented. Finally, acceptable AP1000 core CIPS/CILC risk assessment results are summarized and suggestions that specifically target reducing CIPS/CILC risks for AP1000 plants are described.

scholarly journals ANALYSIS ON THE PERFORMANCE OF THE BANDUNG CONVERSION FUEL-PLATE TRIGA REACTOR IN STEADY STATE WITH CONSTANT COOLANT FLOW RATE

2020 ◽  
Vol 22 (2) ◽  
pp. 41
Author(s):  
Endiah Puji Hastuti ◽  
Sudjatmi K. Alfa ◽  
Sudarmono Sudarmono
Keyword(s):  
Steady State ◽  
Fuel Element ◽  
Flow Rate ◽  
Flow Instability ◽  
Safety Margin ◽  
Nucleate Boiling ◽  
Coolant Flow ◽  
Coolant Flow Rate ◽  
Steady State Condition ◽  
Triga Reactor

Bandung TRIGA2000 Reactor, a General Atomic (GA)-made research reactor used for training, research andiIsotope production, has been upgraded to operate at power of 2000 kW using TRIGA fuel rod type. Recently, the TRIGA reactor fuel element producers are going to discontinue the production of TRIGA fuel element. To overcome the unavailability of TRIGA fuel element, BATAN planned to modify TRIGA2000 fuel type from rod-type to U3Si2-Al plate-type fuel with 19.75% enrichment, similar to the domestically fabricated one used in RSG-GAS. The carried out design emphasized on the determination of operation condition limits for setting the reactor protection system in accordance to the reactor safety calculation results. The conceptual design of the innovative fuel plate TRIGA reactor cooling system is expected to remove heat generated by fuels with nominal power of 1 MW up to 2 MW. The design is developed through modelling and safety analysis using COOLOD-N2 validated code. The safety margin is set to its flow instability at transient condition of the fuel plate, which is ≥ 2.38; departure from nucleate boiling ratio ≥1.50; and no onset of nucleate boiling, ΔTONB ≥ 0oC. The primary coolant flow rate accommodating the existing Bandung TRIGA reactor capability is as high as 50 kg/s. The analysis results show that at power of 1 MW, the reactor can safely operate, while at power of 2 MW the safety margin is exceeded. In other words, the plate TRIGA reactor that employs forced convection mode operates safely at 1 MW with excess power 120% of its nominal power.Keywords: 1 MW, Thermalhydraulic design, Steady state condition, TRIGA plate, Constant flowrate

Development of a Sub-Channel Analysis Code “MIDAC” for Core Thermal-Hydraulic Design and Safety Analysis in PWR Plants

2017 ◽  
Author(s):  
Tadakatsu Yodo ◽  
Naohiro Takeda ◽  
Naoko Iida ◽  
Motoko Kawachi
Keyword(s):  
Two Phase Flow ◽  
Safety Analysis ◽  
Nucleate Boiling ◽  
Phase Flow ◽  
Two Phase ◽  
Drift Flux Model ◽  
Drift Flux ◽  
Hydraulic Design ◽  
Safety Margins ◽  
Channel Analysis

In PWR, a Departure from Nucleate Boiling (DNB) is one of criteria for the thermal-hydraulic design and safety analysis. A sub-channel analysis code calculates local coolant conditions to evaluate the PWR safety margins such as a DNB Ratio (DNBR). Mitsubishi Heavy Industries, LTD (MHI) has developed Mitsubishi Three Dimensional Drift flux Code for Analysis of Core Two-Phase Flow (MIDAC) that began the development since the 1990s which is a sub-channel analysis code for DNBR and Peak Cladding Temperature (PCT) evaluations. The code design is based on a drift flux model for the two-phase flow and a radial heat conduction model for the fuel rod temperatures. MIDAC has been verified by comparisons with exact solutions and other codes, and validated by comparisons with test data based on a Phenomena Identification and Ranking Table (PIRT) under the core thermal-hydraulic design and safety analysis conditions. As a result, MHI confirmed the applicability of MIDAC to PWR conditions in the thermal-hydraulic design and Non-LOCA.

Thermal Analysis of the FSP-1 Experiment in the Advanced Test Reactor

2016 ◽  
Author(s):  
Grant L. Hawkes ◽  
Nicolas E. Woolstenhulme
Keyword(s):  
Thermal Analysis ◽  
Flow Instability ◽  
National Laboratory ◽  
Safety Evaluation ◽  
Thermal Safety ◽  
Primary Coolant ◽  
Coolant Pump ◽  
Safety Margins ◽  
Enriched Uranium ◽  
Test Reactor

The U.S. High Performance Research Reactor Conversions fuel development team is focused on developing and qualifying the uranium-molybdenum (U-Mo) alloy monolithic fuel to support conversion of domestic research reactors to low enriched uranium. Several previous irradiations have demonstrated the favorable behavior of the monolithic fuel. The Full Scale Plate 1 (FSP-1) fuel plate experiment will be irradiated in the northeast (NE) flux trap of the Advanced Test Reactor (ATR). This fueled experiment contains six aluminum-clad fuel plates consisting of monolithic U-Mo fuel meat. Three different types of fuel plates with matching pairs for a total of six plates were analyzed. These three types of plates are: full burn, intermediate power, and thick meat. A thermal analysis has been performed on the FSP-1 experiment to be irradiated in the ATR at the Idaho National Laboratory (INL). A thermal safety evaluation was performed to demonstrate that the FSP-1 irradiation experiment complies with the thermal-hydraulic safety requirements of the ATR Safety Analysis Report (SAR). The ATR SAR requires that minimum safety margins to critical heat flux and flow instability be met in the case of a loss of commercial power with primary coolant pump coast-down to emergency flow. The thermal safety evaluation was performed at 26 MW NE lobe power to encompass the expected range of operating power during a standard cycle. Additional safety evaluations of reactivity insertion events, loss of coolant event, and free convection cooling in the reactor and in the canal are used to determine the response of the experiment to these events and conditions. This paper reports and shows that each safety evaluation complies with each safety requirement of the ATR SAR.

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.

scholarly journals Assessment of severe accident source terms in pressurized-water reactors with a 40% mixed-oxide and 60% low-enriched uranium core using MELCOR 1.8.5.

2010 ◽  
Author(s):  
Randall O. Gauntt ◽  
Andrew S. Goldmann ◽  
Kenneth C. Wagner ◽  
Dana Auburn Powers ◽  
Scott G. Ashbaugh ◽  
...  
Keyword(s):  
Mixed Oxide ◽  
Severe Accident ◽  
Pressurized Water Reactors ◽  
Source Terms ◽  
Pressurized Water ◽  
Water Reactors ◽  
Enriched Uranium

scholarly journals Startup, transition core, and molybdenum-99 production upgrade analyses for low-enriched uranium fuel conversion at the University of Missouri research reactor

2020 ◽  
Author(s):  
◽  
Wilson Cowherd
Keyword(s):  
Low Power ◽  
Research Reactor ◽  
The United States ◽  
Fuel Conversion ◽  
Reactor Physics ◽  
University Of Missouri ◽  
Safety Margins ◽  
Enriched Uranium ◽  
The University ◽  
The Impact

Under the direction of the United States Department of Energy (DOE) National Nuclear Security Administration (NNSA) Office of Material Management and Minimization (M3) Reactor Conversion Program, the University of Missouri Research Reactor (MURR®) plans to convert from highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Low power physics startup test predictions, transition core planning, and analysis for a proposed fission-based molybdenum-99 production upgrade were done in support of LEU fuel conversion. As a first step to LEU fuel conversion, low-power physics tests will be performed to calculate reactor physics parameters. These parameters include flux distributions, coefficients of reactivity, and critical assembly measurements. To facilitate this test, reactor physics calculations were performed using MCNP5 to predict the values of these parameters. Implications of these predictions and areas of uncertainty in the prediction analysis are also discussed. Once MURR completes the testing of the initial LEU core, MURR will enter into a series of transition cycles until steady-state mixed-burnup operation is reached. A Python program was developed that incorporated the constraints of MURR operation while minimizing the time MURR will have to operate atypically during the transition cycles. The impacts of the transition cycles on experiment performance are reported, as well as the number of fuel elements needed. Finally, preliminary analysis on a proposed molybdenum-99 production device at MURR was performed. This analysis shows the impact on the reactor power distribution with implications to predicted safety margins as a part of the larger scope of the experiment analysis.

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