External Hazard Coinciding With Small Break LOCA Thermohydraulic Calculation With System Code ATHLET

2018 ◽  
Author(s):  
Kai Kosowski ◽  
Marcus Seidl
Keyword(s):  
Boundary Conditions ◽  
Diesel Engines ◽  
Power Supply ◽  
Heat Removal ◽  
Design Feature ◽  
Primary Circuit ◽  
External Event ◽  
Grid Connection ◽  
Design Basis ◽  
Loss Of Coolant Accident

The safety behaviors of a NPP after an external hazard-initiated event, as well as after a small break LOCA are already well known as part of the analyses done for standard licensing. In this paper, a combination of both event categories is investigated by means of the thermohydraulic system code ATHLET (analysis of the thermohydraulics of leaks and transients) developed by the Gesellschaft für Anlagen- und Reaktorsicherheit (GRS, Germany). The scenario assumes an external event with a loss of coolant accident caused by induced vibrations on a small pipe attached to the primary circuit, although all pipes are designed to withstand the loads created by such an external event. The coincidence of both events leads to a beyond-design basis consideration. Furthermore, in the context of both robustness and enveloping analyses, the unavailability of both the internal power supply via the external grid connection and the emergency diesel power supply due to the external impact is postulated. For emergency cooling, the NPP in the scenario considered has only access to the passive accumulators and to systems supplied via the safeguard emergency diesel engines (2nd quartet of emergency diesel engines) — a design feature of the Siemens KWU type PWRs —, which are housed in the bunkered emergency feed building and operate with lower voltage levels. The scenario is exemplarily modeled and simulated with ATHLET. The manner of the external event itself is not in focus, but rather the thermohydraulic behavior of the NPP is considered in the reported analysis. Beside the model assumptions and the boundary conditions, the accident sequence is explained in detail. It turns out that the remaining systems for emergency cooling are able to handle the LOCA under such demanding boundary conditions. Most importantly, the primary pressure can be lowered below the zero flow pump head of the pool cooling pumps. The long-term cooling can thereby be ensured. Furthermore, the heat removal out of the core is always sufficient. Overall it can be concluded that all safety protection objectives have been complied for this beyond-design basis scenario.

External Hazard Coinciding With Small Break LOCA—Thermohydraulic Calculation With System Code ATHLET

2020 ◽  
Vol 6 (2) ◽  
Author(s):  
Kai Kosowski ◽  
Marcus Seidl
Keyword(s):  
Diesel Engines ◽  
Nuclear Power ◽  
Heat Removal ◽  
Primary Circuit ◽  
External Event ◽  
Design Basis ◽  
Loss Of Coolant Accident ◽  
License Application ◽  
Thermohydraulic Calculation ◽  
Core Cooling

Abstract The safety behaviors of a nuclear power plant (NPP) after an external hazard-initiated event, as well as after a small break (SB) loss of coolant accident (LOCA), are already well known as part of the analyses made for standard license application. The coincidence of both events leads to a beyond-design basis consideration. Such a combination of both event categories is investigated by means of the thermohydraulic system code ATHLET. The scenario assumes an external event with a LOCA caused by induced vibrations on a small pipe attached to the primary circuit, although all pipes are designed to withstand the loads created by such an external event. Furthermore, in the context of both robustness and enveloping analyses, both a loss of offsite power (LOOP) and an unavailability of the emergency diesel power supply are postulated. The NPP in the scenario considered only has access to the passive accumulators and to systems supplied via the safeguard emergency diesel engines (second quartet of emergency diesel engines), which are housed in the bunkered emergency feed building. The dedicated type of external event itself is not in focus, but rather the thermohydraulic behavior of the NPP is considered. Apart from the model's assumptions, the accident sequence is explained in detail. The remaining systems for emergency core cooling are capable of handling the LOCA under such demanding boundary conditions. Long-term cooling can be ensured. Furthermore, heat removal out of the core is always sufficient. Eventually, all safety protection objectives have been complied for this beyond-design basis scenario.

scholarly journals Analysis and Estimation of Core Damage Frequency of Flow Blockage and Loss of Coolant Accident: A Case Study of a 10 MW Water-Water Research Reactor-PSA Level 1

2021 ◽  
Vol 2021 ◽  
pp. 1-17
Author(s):  
F. Ameyaw ◽  
R. Abrefah ◽  
S. Yamoah ◽  
S. Birikorang
Keyword(s):  
Research Reactor ◽  
Heat Removal ◽  
Primary Circuit ◽  
Water Research ◽  
Core Damage ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Level 1 ◽  
Safety Systems ◽  
Flow Blockage

Fault trees (FT) and event trees (ET) are widely used in industry to model and evaluate the reliability of safety systems. This work seeks to analyze and estimate the core damage frequency (CDF) due to flow blockage (FB) and loss of coolant accident (LOCA) due to large rupture of primary circuit pipe with respect to a specific 10 MW Water-Water Research Reactor in Ghana using the FT and ET technique. Using FT, the following reactor safety systems: reactor protection system, primary heat removal system, isolation of the reactor pool, emergency core cooling system (ECCS), natural circulation heat removal, and isolation of the containment were evaluated for their dependability. The probabilistic safety assessment (PSA) Level 1 was conducted using a commercial computational tool, system analysis program for practical coherent reliability assessment (SAPHIRE) 7.0. The frequency of an accident resulting in severe core damage for the internal initiating event was estimated to be 2.51e − 4/yr for the large LOCA as well as 1.45e − 4/yr for FB, culminating in a total core damage frequency of 3.96e − 4/yr. The estimated values for the frequencies of core damage were within the expected margins of 1.0e − 5/yr to 1.0e − 4/yr and of identical sequence of the extent as found for similar reactors.

Calculation studies of coated particles performance in sodium-cooled fast reactor

Kerntechnik ◽  
2021 ◽  
Vol 86 (1) ◽  
pp. 45-49
Author(s):  
N. V. Maslov ◽  
E. I. Grishanin ◽  
P. N. Alekseev
Keyword(s):  
Fast Reactor ◽  
Reactor Core ◽  
Heat Removal ◽  
Design Solution ◽  
Fast Neutrons ◽  
Steel Shell ◽  
Primary Circuit ◽  
Coated Particles ◽  
The Core ◽  
Fuel Assemblies

Abstract This paper presents results of calculation studies of the viability of coated particles in the conditions of the reactor core on fast neutrons with sodium cooling, justifying the development of the concept of the reactor BN with microspherical fuel. Traditional rod fuel assemblies with pellet MOX fuel in the core of a fast sodium reactor are directly replaced by fuel assemblies with micro-spherical mixed (U,Pu)C-fuel. Due to the fact that the micro-spherical (U, Pu)C fuel has a developed heat removal surface and that the design solution for the fuel assembly with coated particles is horizontal cooling of the microspherical fuel, the core has additional possibilities of increasing inherent (passive) safety and improve the competitiveness of BN type of reactors. It is obvious from obtained results that the microspherical (U, Pu)C fuel is limited with the maximal burn-up depth of ∼11% of heavy atoms in conditions of the sodium-cooled fast reactor core at the conservative approach; it gives the possibility of reaching stated thermal-hydraulic and neutron-physical characteristics. Such a tolerant fuel makes it less likely that fission products will enter the primary circuit in case of accidents with loss of coolant and the introduction of positive reactivity, since the coating of microspherical fuel withstands higher temperatures than the steel shell of traditional rod-type fuel elements.

Radiation Dose Evaluation of Typical Design Basis Accident for Advanced PWR in China

2021 ◽  
Author(s):  
Haiying Chen ◽  
Shaowei Wang ◽  
Xinlu Tian ◽  
Fudong Liu
Keyword(s):  
Radiation Dose ◽  
Effective Dose ◽  
Half Life ◽  
Nuclear Power ◽  
Outer Boundary ◽  
Design Basis ◽  
Loss Of Coolant Accident ◽  
Design Basis Accident ◽  
Typical Design ◽  
Total Effective Dose

Abstract The loss of coolant accident (LOCA) is one of the typical design basis accidents for nuclear power plant. Radionuclides leak to the environment and cause harm to the public in LOCA. Accurate evaluation of radioactivity and radiation dose in accident is crucial. The radioactivity and radiation dose model in LOCA were established, and used to analyze the radiological consequence at exclusion area boundary (EAB) and the outer boundary of low population zone (LPZ) for Hualong 1. The results indicated that the long half-life nuclides, such as 131I, 133I, 135I, 85Kr, 131mXe, 133mXe and 133Xe, released to environment continuously, while the short half-life nuclides, such as 132I, 134I, 83mKr, 85mKr, 87Kr, 88Kr, 135mXe and 138Xe, no longer released to environment after a few hours in LOCA. 133Xe may release the largest radioactivity to environment, more than 1015Bq. Inhalation dose was the major contribution to the total effective dose. The total effective dose and thyroid dose of Hualong 1 at EAB and the outer boundary of LPZ fully met the requirements of Chinese GB6249.

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.

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.

Micro-Porous Coatings and Enhanced CHF for Downward Facing Boiling During Passive Emergency Reactor Cooling

2016 ◽  
Author(s):  
Albert E. Segall ◽  
Faruk A. Sohag ◽  
Faith R. Beck ◽  
Lokanath Mohanta ◽  
Fan-Bill Cheung ◽  
...  
Keyword(s):  
Cold Spray ◽  
Heat Removal ◽  
Porous Coating ◽  
Porous Coatings ◽  
Spray Coatings ◽  
Spray Process ◽  
External Cooling ◽  
Loss Of Coolant Accident ◽  
Lower Head ◽  
Viable Approach

During a Reaction Initiated Accident (RIA) or Loss of Coolant Accident (LOCA), passive external-cooling of the reactor lower head is a viable approach for the in-vessel retention of Corium; while this concept can certainly be applied to new constructions, it may also be viable for operational systems with existing cavities below the reactor. However, a boiling crisis will inevitably develop on the reactor lower head owing to the occurrence of Critical Heat Flux or CHF that could reduce the decay heat removal capability as the vapor phase impedes continuous boiling. Fortunately, this effect can be minimized for both new and existing reactors through the use of a Cold-Spray delivered, micro-porous coating that facilitates the formation of vapor micro-jets from the reactor surface. The micro-porous coatings were created by first spraying a binary mixture with the sacrificial material then removed via etching. Subsequent quenching experiments on uncoated and coated hemispherical surfaces showed that local CHF values for the coated vessel were consistently higher relative to the bare surface. Moreover, it was observed for both coated and uncoated surfaces that the local rate of boiling and local CHF limit varied appreciably along the outer surface. Nevertheless, the results of this intriguing study clearly show that the use of Cold Spray coatings could enhance the local CHF limit for downward facing boiling by more than 88%. Moreover, the Cold-Spray process is amenable to coating the lower heads of operating reactors.

Analysis of Passive Residual Heat Removal System at Primary Side when Station Blackout Occurs

2014 ◽  
Vol 986-987 ◽  
pp. 231-234
Author(s):  
Jun Teng Liu ◽  
Qi Cai ◽  
Xia Xin Cao
Keyword(s):  
Nuclear Reactor ◽  
Natural Circulation ◽  
Reactor Core ◽  
Saturation Temperature ◽  
Heat Removal ◽  
Primary Circuit ◽  
Station Blackout ◽  
Heat Removal System ◽  
Removal System ◽  
Residual Heat

This paper regarded CNP1000 power plant system as the research object, which is the second-generation half Nuclear Reactor System in our country, and tried to set Westinghouse AP1000 passive residual heat removal system to the primary circuit of CNP1000. Then set up a simulation model based on RELAP5/MOD3.2 program to calculate and analyze the response and operating characteristic of passive residual heat removal system on assumption that Station Blackout occurs. The calculation has the following conclusions: natural circulation was quickly established after accident, which removes core residual heat effectively and keep the core safe. The residual heat can be quickly removed, and during this process the actual temperature was lower than saturation temperature in reactor core.

scholarly journals Edge Cooling of a Fuel Cell during Aerial Missions by Ambient Air

Micromachines ◽  
2021 ◽  
Vol 12 (11) ◽  
pp. 1432
Author(s):  
Lev Zakhvatkin ◽  
Alex Schechter ◽  
Eilam Buri ◽  
Idit Avrahami
Keyword(s):  
Fuel Cell ◽  
Numerical Model ◽  
Wind Tunnel ◽  
Boundary Conditions ◽  
Air Temperature ◽  
Numerical Models ◽  
Heat Removal ◽  
Ambient Air ◽  
Experimental Setup ◽  
Good Agreement

During aerial missions of fuel-cell (FC) powered drones, the option of FC edge cooling may improve FC performance and durability. Here we describe an edge cooling approach for fixed-wing FC-powered drones by removing FC heat using the ambient air during flight. A set of experiments in a wind tunnel and numerical simulations were performed to examine the efficiency of FC edge cooling at various flight altitudes and cruise speeds. The experiments were used to validate the numerical model and prove the feasibility of the proposed method. The first simulation duplicated the geometry of the experimental setup and boundary conditions. The calculated temperatures of the stack were in good agreement with those of the experiments (within ±2 °C error). After validation, numerical models of a drone’s fuselage in ambient air with different radiator locations and at different flight speeds (10–30 m/s) and altitudes (up to 5 km) were examined. It was concluded that onboard FC edge cooling by ambient air may be applicable for velocities higher than 10 m/s. Despite the low pressure, density, and Cp of air at high altitudes, heat removal is significantly increased with altitude at all power and velocity conditions due to lower air temperature.

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