scholarly journals Universal system of passive heat removal from the core of a research reactor

2019 ◽  
Vol 34 (2) ◽  
pp. 107-121
Author(s):  
Vitaly Uzikov ◽  
Irina Uzikova
Keyword(s):  
Research Reactor ◽  
Cooling System ◽  
Heat Removal ◽  
Check Valve ◽  
Universal System ◽  
Cooling Circuit ◽  
The Core ◽  
Wide Range ◽  
Reactor Installation ◽  
Moving Parts

This paper presents the results of an analysis of a universal cooling system for the core of re- search reactors built on the passive principle of natural convection. A 3-D model, technologi- cal and design diagrams of the reactor installation are provided, along with examples of nu- merical evaluation of transients during the operation of the cooling circuit in normal and emergency modes to substantiate the possibility of using such a cooling system in research re- actors of small and medium power. The principal feature of the described passive system is the absence of not only active elements, such as circulation pumps and shut-off and control valves from the cooling circuit, but also of passive elements with moving parts, such as a check valve. The cooling circuit includes only vessels, piping and a heat exchanger. The absence of elements with mechanical moving parts can significantly reduce the likelihood of equipment failures and improve the reliability of such a cooling system while also reducing its cost. The versatility of the proposed system allows it to be used for a wide range of research reactor plants with various capacities, which are nowadays being developed designed to carry out programs in various areas of research and applied usages related to nuclear technologies.

Wearable Engine-Driven Vapor-Compression Cooling System for Elevated Ambients

2009 ◽  
Vol 1 (2) ◽  
Author(s):  
Timothy C. Ernst ◽  
Srinivas Garimella
Keyword(s):  
Cooling System ◽  
Average Rate ◽  
Heat Removal ◽  
Activity Level ◽  
Working Environment ◽  
Gear Train ◽  
Reduction Gear ◽  
Small Scale ◽  
Vapor Compression ◽  
Wide Range

The development of a wearable cooling system for use in elevated-temperature environments by military, fire-fighting, chemical-response, and other hazardous duty personnel is underway. Such a system is expected to reduce heat-related stresses, increasing productivity and allowable mission duration, to reduce fatigue, and to lead to a safer working environment. The cooling system consists of an engine-driven vapor-compression system assembled in a backpack configuration, to be coupled with a cooling garment containing refrigerant lines worn in close proximity to the skin. A 2.0 l fuel tank powers a small-scale engine that runs a compressor modified from the original air compression application to the refrigerant compression application here. A centrifugal clutch and reduction gear train system was designed and fabricated to couple the engine output to the refrigerant compressor and heat rejection fan. The system measured 0.318×0.273×0.152 m3 and weighed 4.46 kg. Testing was conducted in a controlled environment to determine performance over a wide range of expected ambient temperatures (37.7–47.5°C), evaporator refrigerant temperatures (22.2–26.1°C), and engine speeds (10,500–13,300 rpm). Heat removal rates of up to 300 W, which is the cooling rate for maintaining comfort at an activity level comparable to moderate exercise, were demonstrated at a nominal ambient temperature of 43.3°C. The system consumes fuel at an average rate of 0.316 kg/h to provide nominal cooling of 178 W for 5.7 h between refueling.

The Specifics of RBMK Core Cooling at Overheated Conditions

2006 ◽  
Author(s):  
Eugenijus Uspuras ◽  
Egidijus Urbonavicius ◽  
Algirdas Kaliatka
Keyword(s):  
Pressure Vessel ◽  
Cooling System ◽  
Heat Removal ◽  
Niobium Alloy ◽  
Point Of View ◽  
The Core ◽  
Flow Length ◽  
On Line ◽  
Core Cooling ◽  
Graphite Moderator

Reactor RBMK-1500 of Ignalina NPP is a boiling light water reactor with graphite moderator. Several important design features of RBMK-1500 are unique and extremely complex in respect to western reactors: the fuel assemblies are loaded into individual channels rather than a single pressure vessel; the plant can be refueled on-line; the neutron spectrum is thermalized by a massive graphite moderator. The reactor coolant system consists of two loops, each having flow length of more than 200 m. There are 1661 of vertical parallel fuel channels and numerous components, such as headers, pumps, valves, etc. The fuel channels and fuel claddings are made of Zirconium-Niobium alloy. From the point of view of safety barriers, each fuel channel in RBMK-type reactor corresponds to a pressure vessel of vessel-type reactors. Thus, the fuel channels are the most important element in reactor cooling system. However, in case of beyond design basis accidents with loss heat removal from the core the integrity of fuel channels could be challenged as they are not so strong as the pressure vessel. The paper presents the analysis of different possibilities to cooldown the core of RBMK-type reactors. Injection of water to RCS is considered as main strategy. Such “bleed and feed” procedure is used for vessel type reactors, but at present is not considered at RBMK-1500.

Wearable Engine-Driven Vapor-Compression Cooling System for Elevated Ambients

2007 ◽  
Author(s):  
Timothy C. Ernst ◽  
Srinivas Garimella
Keyword(s):  
Cooling System ◽  
Average Rate ◽  
Heat Removal ◽  
Activity Level ◽  
Working Environment ◽  
Gear Train ◽  
Reduction Gear ◽  
Small Scale ◽  
Vapor Compression ◽  
Wide Range

The development of a wearable cooling system for use in elevated temperature environments by military, fire-fighting, chemical-response, and other hazardous duty personnel is underway. Such a system is expected to reduce heat-related stresses, increasing productivity and allowable mission duration, reduce fatigue, and lead to a safer working environment. The cooling system consists of an engine-driven vapor compression system assembled in a backpack configuration, to be coupled with a cooling garment containing refrigerant lines worn in close proximity to the skin. A 2.0 L fuel tank powers a small-scale engine that runs a compressor modified from the original air compression application to the refrigerant compression application here. A centrifugal clutch and reduction gear train system was designed and fabricated to couple the engine output to the refrigerant compressor and heat rejection fan. The system measured 0.318×0.273×0.152 m and weighed 4.46 kg. Testing was conducted in a controlled environment to determine performance over a wide range of expected ambient temperatures (37.7–47.5°C), evaporator refrigerant temperatures (22.2–26.1°C), and engine speeds (10,500–13,300 RPM). Heat removal rates of up to 300 W, which is the cooling rate for maintaining comfort at an activity level comparable to moderate exercise, were demonstrated at a nominal ambient temperature of 43.3°C. The system consumes fuel at an average rate of 0.316 kg/hr to provide nominal cooling of 178 W for 5.7 hrs between refueling.

The Analysis of AP1000 Beyond Design Basis DEDVI Accident

2013 ◽  
Author(s):  
Sheng Zhu
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Cooling System ◽  
Heat Removal ◽  
The Core ◽  
Design Basis ◽  
Injection Line ◽  
Design Basis Accident ◽  
Core Cooling

Double ended break of direct vessel injection line (DEDVI) is the most typical small-break lost of coolant accident (LOCA) in AP 1000 nuclear power plant. This study simulated the DEDVI (without actuation of automatic depressurization system 1–3 stage valves, accumulators and passive residual heat removal heat exchanger) beyond design basis accident (BDBA) to validate the safety capability of AP1000 under such conditions. The results show that the core will be uncovered for about 863 seconds and then recovered by water after gravity injection from IRWST into the pressure vessel. The peak cladding temperature (PCT) goes up to 838.08°C, much lower than the limiting value 1204°C. This study confirms that in the DEDVI beyond design basis accident, the passive core cooling system (PXS) can effectually cool the core and preserve it integrate, and ensure the safety of AP 1000 nuclear power plant.

scholarly journals Evaluation of Heat Removal from RBMK-1500 Core Using Control Rods Cooling Circuit

2008 ◽  
Vol 2008 ◽  
pp. 1-8
Author(s):  
A. Kaliatka ◽  
E. Uspuras ◽  
M. Vaisnoras
Keyword(s):  
Nuclear Power ◽  
Cooling System ◽  
Reactor Core ◽  
Heat Removal ◽  
Thermal Inertia ◽  
Severe Accident ◽  
Cooling Circuit ◽  
High Thermal Inertia ◽  
The Moment ◽  
Control Rods

The Ignalina nuclear power plant is a twin unit with two RBMK-1500, graphite moderated, boiling water, multichannel reactors. After the decision was made to decommission the Ignalina NPP, Unit 1 was shut down on December 31, 2004, and Unit 2 is to be operated until the end of 2009. Despite of this fact, severe accident management guidelines for RBMK-1500 reactor at Ignalina NPP are prepared. In case of beyond design basis accidents, it can occur that no water sources are available at the moment for heat removal from fuel channels. Specificity of RBMK reactor is such that the channels with control rods are cooled with water supplied by the system totally independent from the reactor cooling system. Therefore, the heat removal from RBMK-1500 reactor core using circuit for cooling of rods in control and protection system can be used as nonregular mean for reactor cooldown in case of BDBA. The heat from fuel channels, where heat is generated, through graphite bricks is transferred in radial direction to cooled CPS channels. This article presents the analysis of possibility to remove heat from reactor core in case of large LOCA by employing CPS channels cooling circuit. The analysis was performed for Ignalina NPP with RBMK-1500 reactor using RELAP5-3D and RELAP5 codes. Results of the analysis have shown that, in spite of high thermal inertia of graphite, this heat removal from CPS channels allows to slow down effectively the core heat-up process.

Generation of Pressure Distribution Inside a Decay Tank in a Research Reactor Using CFD

2013 ◽  
Author(s):  
Seong Hoon Kim ◽  
Kyoungwoo Seo ◽  
Dae-young Chi ◽  
Juhyeon Yoon
Keyword(s):  
Research Reactor ◽  
Gamma Ray ◽  
Cooling System ◽  
Perforated Plate ◽  
Reactor Core ◽  
Perforated Plates ◽  
Pump Operation ◽  
The Core ◽  
Cooling Loop ◽  
Irradiation Device

The Primary Cooling System (PCS) of a research reactor circulates coolant to remove the heat produced in the fuel or irradiation device. The core outlet coolant contains many kinds of radionuclides because it passes the reactor core [1]. As N-16 among them emits a very hard gamma ray, it not only causes radiation damage to some components but also requires very heavy shielding of the primary cooling loop. Since its half-life is 7.13s, its level can be effectively lowered by installing a decay tank including an internal structure to provide a transit time [2]. To ensure that the N-16 activity decreases enough before the coolant leaves the heavily shielded decay tank room, perforated plates are installed inside the decay tank. The perforated plates are designed to disturb and delay the PCS flow. Normally, when a flow from a relative narrow inlet nozzle goes out to an enlarged tank, it becomes a complex turbulent flow inside the tank. In addition, the PCS flow is frequently changed from zero to a normal flow rate owing to the research reactor characteristics. Thus, the integrity of the perforated plate shall be verified with the pump operation and shutdown condition.

Experimental Study on the Performance of IIST Passive Core Cooling System

2002 ◽  
Author(s):  
Chin-Jang Chang ◽  
Chien-Hsiung Lee ◽  
Wen-Tan Hong ◽  
Lance L. C. Wang
Keyword(s):  
Nuclear Power ◽  
Cooling System ◽  
Heat Removal ◽  
General System ◽  
System Response ◽  
Water Storage Tank ◽  
The Core ◽  
Third Stage ◽  
Core Cooling

The purpose of this study is to conduct the experiments at the Institute of Nuclear Energy Research (INER) Integral System Test (IIST) facility for evaluation of the performance of the passive core cooling system (PCCS) during the cold-leg small break loss-of-coolant accidents (SBLOCAs). Five experiments were performed with (1) three different break sizes, 2%, 0.5%, and 0.2% (approximately corresponding to 1 1/4”, 2”, and 4” breaks for Maanshan nuclear power plant), and (2) 0.2% and 0.5% without actuation of the first-stage and third-stage automatic depressurization valve (ADS-1 and ADS-3) to initiate PCCS for assessing its capacity in accident management. The detailed descriptions of general system response and the interactions of core makeup tanks (CMTs), accumulators (ACCs), automatic depressurization system (ADS), passive residual heat Removal (PRHR), and in-containment refueling water storage tank (IRWST) on the core heat removal are included. The results show: (1) core long term cooling can be maintained for all cases following the PCCS procedures, (2) the core can be covered for the cases of the 0.2% and 0.5% breaks without actuation of ADS-1 and ADS-3.

Simulation of Performance of Centrifugal Circulators With Vaneless Diffuser for GCR Applications

2008 ◽  
Author(s):  
N. Tauveron
Keyword(s):  
Nuclear Reactor ◽  
Heat Removal ◽  
Vaneless Diffuser ◽  
International Forum ◽  
Dimensional Approach ◽  
Present Contribution ◽  
One Dimensional ◽  
The Core ◽  
Whole Plant ◽  
Wide Range

In the frame of the international forum GenIV, CEA has selected various innovative concepts of Gas cooled Nuclear Reactor. Among them, an indirect-cycle gas reactor is under consideration. Thermal hydraulic performances are a key issue for the design. For transient conditions and decay heat removal situations, the thermal hydraulic performance must remain as high as possible. In this context, all the transient situations, the incidental and accidental scenarii must be evaluated by a validated system code able to correctly describe, in particular, the thermalhydraulics of the whole plant. As concepts use a helium compressor to maintain the flow in the core, a special emphasis must be laid on compressor modelling. Centrifugal circulators with a vaneless diffuser have significant properties in term of simplicity, cost, ability to operate over a wide range of conditions. The objective of this paper is to present a dedicated description of centrifugal compressor, based on a one dimensional approach. This type of model requires various correlations as input data. The present contribution consists in establishing and validating the numerical simulations (including different sets of correlations) by comparison with representative experimental data. The results obtained show a qualitatively correct behaviour of the model compared to open literature cases of the gas turbine aircraft community and helium circulators of High Temperature Gas Reactors. Further work on modelling and validation are nevertheless needed to have a better confidence in the simulation predictions.

Operational Map and Thermal Performance of a Thermosyphon Cooling System for Compact Servers

2021 ◽  
Author(s):  
Raffaele L. Amalfi ◽  
Cong H. Hoang ◽  
Ryan Enright ◽  
Filippo Cataldo ◽  
Jackson B. Marcinichen ◽  
...  
Keyword(s):  
Heat Flux ◽  
Thermal Performance ◽  
Cooling System ◽  
Heat Removal ◽  
Total Power ◽  
Two Phase ◽  
Maximum Heat ◽  
Wide Range ◽  
Experimental Apparatus ◽  
Secondary Side

Abstract This paper advances the state-of-the-art in novel passive two-phase systems for more efficient cooling of datacenters and telecom central offices compared to the traditional air-based cooling solutions (e.g. aisle-based containment systems). The proposed passive two-phase technology uses numerous server-level thermosyphons to dissipate the heat generated by critical components, such as central processing units, accelerators, etc., with the flexibility of selecting the rack-level and room-level cooling elements depending on the deployment scenarios. The main goal of this paper is to experimentally investigate the thermal performance and maximum heat removal capability of a server-level thermosyphon for cooling compact servers. The experimental apparatus, built at Nokia Bell Labs, incorporates a single 7-cm high liquid-cooled thermosyphon that fits within a 2U server (smaller form factors can be achieved by a proper design that would further reduce the thermosyphon’s height). The heat source is represented by a pseudo-chip, composed of six parallel cartridge heaters installed in a copper block that incorporates local temperature measurements and is able of dissipating a total power of ≈ 500 W over a footprint area of 3.5 cm × 3.5 cm (corresponding heat flux of ≈ 41 W/cm2). Steady-state experiments were carried out at various heat loads up to 240 W (corresponding heat flux of ≈ 20 W/cm2), filling ratios and secondary side inlet conditions (coolant temperatures and mass flow rates), using R1234ze(E) and deionized water as the working fluids on the primary and secondary side, respectively. Test results demonstrate high heat transfer performance of the server-level thermosyphon over a wide range of conditions, and operating points are identified and classified into an operational map. Thermosyphon-based cooling systems across multiple length scales can significantly improve operation in terms of lowering energy consumption, allowing for higher hardware density, increased processing speed and reliability.

Passive Safety Features in Sodium Cooled Super-Safe, Small and Simple Reactor

2002 ◽  
Author(s):  
N. Ueda ◽  
I. Kinoshita ◽  
Y. Nishi ◽  
A. Minato ◽  
H. Matsumiya ◽  
...  
Keyword(s):  
Cooling System ◽  
Heat Removal ◽  
Reactor Vessel ◽  
Passive Safety ◽  
Primary Coolant ◽  
Decay Heat ◽  
The Core ◽  
Design Requirements ◽  
Safety Features ◽  
Dynamics Analyses

This paper describes the passive safety features utilized in the updated sodium cooled Super-Safe, Small and Simple fast reactor, which is the improved 4S reactor. This reactor can operate up to ten years without refueling and neutron reflector regulates the reactivity. One of the design requirements is to secure the core against all anticipated transients without reactor scram. Therefore, the reactor concept is to design to enhance the passive safety features. All temperature reactivity feedback coefficients including whole core sodium void worth are negative. Also, introducing of RVACS (Reactor Vessel Auxiliary Cooling System) can enhance the passive decay heat removal capability. Safety analyses are carried out to simulate various transient sequences, which are loss of flow events, transient overpower events and loss of heat sink events, in order to evaluate the passive safety capabilities. A calculation tool for plant dynamics analyses for fast reactors has been modified to model the 4S including the unique plant system, which are reflector control system, circulation pumps and RVACS. The analytical results predict that the designed passive features improve the safety in which temperature variation in transients are satisfied with the safety criteria for the fuel element and the structure of the primary coolant boundary.

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