scholarly journals Status of MELCOR Sodium Models Development

2017 ◽  
Author(s):  
David L. Y. Louie ◽  
Larry L. Humphries
Keyword(s):  
Atmospheric Chemistry ◽  
Equations Of State ◽  
Freezing Point ◽  
Interaction Model ◽  
Nuclear Regulatory Commission ◽  
Pool Fire ◽  
Primary Circuit ◽  
Accident Analysis ◽  
Model Code ◽  
Design Basis

A sodium coolant accident analysis code is necessary to provide regulators with a means of performing confirmatory analyses for future sodium reactor licensing submissions. MELCOR and CONTAIN, which have been employed by the U.S. Nuclear Regulatory Commission for light water reactor licensing, have been traditionally used for Level 2 and Level 3 probabilistic analyses as well as containment design basis accident analysis. To meet future regulatory needs, new models are being added to the MELCOR code for simulation of sodium reactor designs by integrating the existing models developed for separate effects codes into the MELCOR architecture. Sodium properties and equations of state, such as from the SAS4A code, have previously been implemented into MELCOR to replace the water properties and equation of state. Additional specific sodium-related models to address design basis accidents are now being implemented into MELCOR from CONTAIN-LMR. Although the codes are very different in the code architecture, the feasibility fit is being investigated, and the models for the sodium spray fire and the sodium pool fire have been integrated into MELCOR. A new package called Sodium Chemistry (NAC) has been added to MELCOR to handle all sodium related chemistry models for sodium reactor safety applications. Although MELCOR code requires the ambient condition to be above the freezing point of the coolant (e.g., sodium or water), the high relative freezing point of sodium requires MELCOR to handle situations, particularly far from the primary circuit, where the ambient temperatures are usually at room temperature. Because only a single coolant can be modeled in a problem at a time, any presence of water in the problem would be treated as a trace material, an aerosol, in MELCOR. This paper addresses and describe the integration of the sodium models from CONTAIN-LMR, and the testing of the sodium chemistry models in the NAC package of MELCOR that handles sodium type reactor accidents, using available sodium experiments on spray fire and pool fire. In addition, we describe the anticipated sodium models to be completed in this year, such as the atmospheric chemistry model and sodium-concrete interaction model. Code-to-code comparison between MELCOR and CONTAIN-LMR results, in addition to the experiment code validations, will be demonstrated in this year.

scholarly journals A Safety Re-Evaluation of the AVR Pebble Bed Reactor Operation and Its Consequences for Future HTR Concepts

2008 ◽  
Author(s):  
Rainer Moormann
Keyword(s):  
Fission Product ◽  
End Of Life ◽  
Fission Products ◽  
Normal Operation ◽  
Primary Circuit ◽  
Pebble Bed ◽  
Pebble Bed Reactor ◽  
Hot Gas ◽  
Design Basis ◽  
Gas Tight

The AVR pebble bed reactor (46 MWth) was operated 1967–1988 at coolant outlet temperatures up to 990°C. Also because of a lack of other experience the AVR operation is a basis for future HTRs. This paper deals with insufficiently published unresolved safety problems of AVR and of pebble bed HTRs. The AVR primary circuit is heavily contaminated with dust bound and mobile metallic fission products (Sr-90, Cs-137) which create problems in current dismantling. The evaluation of fission product deposition experiments indicates that the end of life contamination reached several percent of a single core inventory. A re-evaluation of the AVR contamination is performed in order to quantify consequences for future HTRs: The AVR contamination was mainly caused by inadmissible high core temperatures, and not — as presumed in the past — by inadequate fuel quality only. The high AVR core temperatures were detected not earlier than one year before final AVR shut-down, because a pebble bed core cannot be equipped with instruments. The maximum core temperatures were more than 200 K higher than precalculated. Further, azimuthal temperature differences at the active core margin were observed, as unpredictable hot gas currents with temperatures > 1100°C. Despite of remarkable effort these problems are not yet understood. Having the black box character of the AVR core in mind it remains uncertain whether convincing explanations can be found without major experimental R&D. After detection of the inadmissible core temperatures, the AVR hot gas temperatures were strongly reduced for safety reasons. Metallic fission products diffuse in fuel kernel, coatings and graphite and their break through takes place in long term normal operation, if fission product specific temperature limits are exceeded. This is an unresolved weak point of HTRs in contrast to other reactors and is particularly problematic in pebble bed systems with their large dust content. Another disadvantage, responsible for the pronounced AVR contamination, lies in the fact that activity released from fuel elements is distributed in HTRs all over the coolant circuit surfaces and on graphitic dust and accumulates there. Consequences of AVR experience on future reactors are discussed. As long as pebble bed intrinsic reasons for the high AVR temperatures cannot be excluded they have to be conservatively considered in operation and design basis accidents. For an HTR of 400 MWth, 900°C hot gas temperature, modern fuel and 32 fpy the contaminations are expected to approach at least the same order as in AVR end of life. This creates major problems in design basis accidents, for maintenance and dismantling. Application of German dose criteria on advanced pebble bed reactors leads to the conclusion that a pebble bed HTR needs a gas tight containment even if inadmissible high temperatures as observed in AVR are not considered. However, a gas tight containment does not diminish the consequences of the primary circuit contamination on maintenance and dismantling. Thus complementary measures are discussed. A reduction of demands on future reactors (hot gas temperatures, fuel burn-up) is one option; another one is an elaborate R&D program for solution of unresolved problems related to operation and design basis accidents. These problems are listed in the paper.

Design basis accident analysis for the Ignitor experiment

2015 ◽  
Vol 98-99 ◽  
pp. 2235-2238 ◽  
Author(s):  
Massimo Zucchetti ◽  
Bruno Coppi ◽  
Maria Teresa Porfiri ◽  
Marco Riva
Keyword(s):  
Accident Analysis ◽  
Design Basis ◽  
Design Basis Accident

scholarly journals Thermal-Hydraulic Safety Analysis of Mixed Core Loads for Ukrainian NPPs with VVER-1000

2016 ◽  
pp. 9-12
Author(s):  
Yu. Vorobyov ◽  
A. Nosovsky ◽  
O. Pohonets ◽  
I. Shevchenko
Keyword(s):  
Nuclear Fuel ◽  
Computer Code ◽  
Safety Analysis ◽  
Accident Analysis ◽  
Fuel Cladding ◽  
Hydraulic Analysis ◽  
Design Basis ◽  
Hydraulic Safety ◽  
Design Basis Accident ◽  
Thermal Hydraulic Analysis

The paper presents thermal-hydraulic analysis of mixed core loads to confirm compliance with safety criteria. The objective is to verify reliability of nuclear fuel cooling in representative events of the design-basis accident analysis. RELAP5/MOD3.2 computer code was applied to show that maximum fuel cladding temperature does not exceed 1200 °C in mixed TVSA-12, TVS-WR and TVSA cores. The analysis led to the conclusion on possible safe implementation of new fuel at Ukrainian NPPs.

scholarly journals Thermal properties of hot and dense matter: Influence of rapid rotation on protoneutron stars, hot neutron stars, and neutron star merger remnants

2021 ◽  
Vol 252 ◽  
pp. 05004
Author(s):  
Polychronis Koliogiannis ◽  
Charalampos Moustakidis
Keyword(s):  
Neutron Star ◽  
Equation Of State ◽  
Neutron Stars ◽  
Nuclear Matter ◽  
Equations Of State ◽  
Interaction Model ◽  
Dense Matter ◽  
Baryon Density ◽  
Dense Nuclear Matter ◽  
Protoneutron Stars

The knowledge of the equation of state is a key ingredient for many dynamical phenomena that depend sensitively on the hot and dense nuclear matter, such as the formation of protoneutron stars and hot neutron stars. In order to accurately describe them, we construct equations of state at FInite temperature and entropy per baryon for matter with varying proton fractions. This procedure is based on the momentum dependent interaction model and state-of-the-art microscopic data. In addition, we investigate the role of thermal and rotation effects on microscopic and macroscopic properties of neutron stars, including the mass and radius, the frequency, the Kerr parameter, the central baryon density, etc. The latter is also connected to the hot and rapidly rotating remnant after neutron star merger. The interplay between these quantities and data from late observations of neutron stars, both isolated and in matter of merging, could provide useful insight and robust constraints on the equation of state of nuclear matter.

Numerical Analysis of Safety Injection System’s Flow Rate

2010 ◽  
Author(s):  
Dengke Sun ◽  
Xiaojiang Wang ◽  
Jun Li ◽  
Jiang Liu ◽  
Changyue Li
Keyword(s):  
Flow Rate ◽  
Optimization Design ◽  
Direct Injection ◽  
Injection System ◽  
Primary Circuit ◽  
Accident Analysis ◽  
Fluid Temperature ◽  
Injection Flow Rate ◽  
Injection Flow ◽  
Injection Pump

In this paper, a hydraulic model for Safety Injection System (SIS) of M310 reactor is extended. The model is checked and calibrated by test results under test conditions. Based on commissioning test criteria, the system’s maximum and minimum pressure drop coefficients are calibrated according to anti-extrapolation method. Considering modifications of various projects, analysis of 41 flow rate curves under different conditions has been performed using this model. These flow rate curves indicate the relationship between injection flow rates and primary circuit pressures. Accident analysis has taken these curves as input data. Also, the strategy for dealing with accident is established based on these curves. Results of accident analysis show that the design of SIS system can satisfy the safety requirements of M310 reactor. The sensitivity analysis of typical conditions illustrates that injection flow rate will increase as the primary circuit pressure decreases. With the same configuration, the injection flow rate during recirculation phase will be smaller than that during direct injection phase, which is mainly caused by the decrement of suction elevation and the increment of fluid temperature. When low head safety injection pump (LHSI PO) is boosting high head safety injection pump (HHSI PO), if the pressure is relatively high, the injection flow rate will not be improved apparently. If the pressure is relatively low, the boosting is necessary. These conclusions can be the basis for the later optimization design.

Used Nuclear Fuel Transportation Package Seal Performance in Beyond Design Basis Thermal Conditions

2011 ◽  
Author(s):  
Felix Gonzalez ◽  
Christopher Bajwa ◽  
Robert Einziger ◽  
Earl Easton ◽  
Jiann Yang ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Experimental Testing ◽  
Thermal Conditions ◽  
Nuclear Regulatory Commission ◽  
Pressure Vessels ◽  
Radioactive Material ◽  
Small Scale ◽  
Thermal Testing ◽  
Design Basis

The U.S. Nuclear Regulatory Commission (NRC) is evaluating the performance of seals in used fuel transportation packages during beyond-design-basis fires, similar to the Baltimore tunnel fire that occurred in 2001. The performance of package seals is important for determining the potential for a release of radioactive material from a package during a beyond-design-basis accident. Seals generally have lower temperature limits than other package components and are often part of the containment barrier between the environment and the cask contents. The NRC’s Office of Nuclear Regulatory Research (RES) funded the National Institute of Standards and Technology (NIST) to conduct small-scale thermal testing to obtain experimental data of the performance of seals during beyond-design basis temperature exposures. The experimental testing consisted of several small-scale pressure vessels fabricated with a modified ASME flange design, using commercial grade metallic seals, similar to those that might be used on an actual spent nuclear fuel transportation package. The vessels were heated in an electrical furnace for exposures up to 9 hours (hrs) at temperatures as high as 800°C (1472°F), which far exceeded the rated temperature of the seals in question. This paper will provide a summary of the testing completed as well as the preliminary results and conclusions of the experiments performed by NIST.

Water-Ingress Accident of the 250MW Pebble-Bed Modular High Temperature Gas-Cooled Reactor

2009 ◽  
Author(s):  
Zheng Yanhua ◽  
Shi Lei
Keyword(s):  
High Temperature ◽  
Safety Valve ◽  
Primary Circuit ◽  
Fuel Temperature ◽  
Pebble Bed ◽  
Water Ingress ◽  
Design Basis ◽  
Design Basis Accident ◽  
Water Gas ◽  
High Temperature Gas

Water-ingress accident, caused by the steam generator heating tube rupture of a high temperature gas-cooled reactor, will introduce a positive reactivity to lead the nuclear power increase rapidly, as well as the chemical reaction of graphite fuel elements and reflector structure material with steam. Increase of the primary circuit pressure may result in the opening of the safety valve, which will cause the release of radioactive isotopes and flammable water gas. The analysis of such an important and particular accident is significant for verifying the inherent safety characteristics of the pebble-bed modular high temperature gas-cooled reactor. Based on the preliminary design of the 250MW Pebble-bed Modular High Temperature Gas-cooled Reactor (HTR-PM), the design basis accident of double-ended guillotine break of a heating tube has been analyzed by using TINTE, which is a special transient analysis program for high temperature gas-cooled reactors. Some safety relevant concerns, such as the fuel temperature and primary loop pressure, the graphite corrosion inventory, the water gas releasing amount, as well as the natural convection influence under the condition of the failure of the blower flaps shut down, have been studied in detail. The calculation result of the design basis accident indicates that, the maximal possible water ingress amount is less than 600 kg and the maximal fuel temperature keeps far below the design limitation of 1620°C. The result also shows that the slight amount of graphite corrosion will not damage the reactor structure and the fuel element, and there is no potential explosive risk caused by the opening of the safety valve.

Spent Nuclear Fuel Transportation Package Seals in Beyond Design Basis Temperature Excursions

2012 ◽  
Author(s):  
Felix Gonzalez ◽  
Christopher Bajwa ◽  
Robert Einziger ◽  
Earl Easton ◽  
Jiann Yang ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Experimental Testing ◽  
Extreme Temperature ◽  
Nuclear Regulatory Commission ◽  
Pressure Vessels ◽  
Radioactive Material ◽  
Small Scale ◽  
Thermal Testing ◽  
Design Basis

The US Nuclear Regulatory Commission (NRC) is studying the performance of seals in spent nuclear fuel (SNF) transportation packages exposed to fires that could exceed the hypothetical accident condition fire described in Title 10 of the Code of Federal Regulations, Part 71, such as the Baltimore Tunnel Fire that occurred in 2001, or the MacArthur Maze fire that occurred in 2007. The performance of package seals is important for determining the potential for release of radioactive material from a package during a beyond-design-basis accident. Seals generally have lower temperature limits than other package components and are the containment barrier between the environment and the radioactive package contents. The NRC Office of Nuclear Regulatory Research contracted the National Institute of Standards and Technology to conduct small-scale thermal testing to obtain experimental data of the performance of seals during extreme temperature exposures. The experimental testing consisted of several small-scale pressure vessels fabricated with a modified ASME flange design and tested metallic and polymeric seals, similar to those that might be used on an actual SNF transportation package. The vessels were heated in an electrical oven to temperatures as high as 800°C (1472°F), exceeding the rated temperatures of the seals in question. This paper will provide a summary of the testing conducted and present test results and conclusions.

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