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.

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.

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.

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.

Test and Evaluation of a Filtering System for Retention of Fibres in a Coolant Flow After Loss-of-Coolant Accident

2012 ◽  
Author(s):  
T. Gocht ◽  
W. Kästner ◽  
A. Kratzsch ◽  
M. Strasser
Keyword(s):  
Cooling System ◽  
Reactor Core ◽  
Heat Removal ◽  
Test Facility ◽  
Experimental Investigations ◽  
Time Interval ◽  
Insulation Material ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Core Cooling

In case of an accident the safe heat removal from the reactor core with the installed emergency core cooling system (ECCS) is one of the main features in reactor safety. During a loss-of-coolant accident (LOCA) the release of insulation material fragments in the reactor containment can lead to malfunctions of ECCS. Therefore, the retention of particles by strainers or filtering systems in the ECCS is one of the major tasks. The aim of the presented experimental investigations was the evaluation of a filtering system for the retention of fiber-shaped particles in a fluid flow. The filtering system consists of a filter case with a special lamellar filter unit. The tests were carried out at a test facility with filtering units of different mesh sizes. Insulation material (mineral rock wool) was fragmented to fiber-shaped particles. To simulate the distribution of particle concentration at real plants with large volumes the material was divided into single portions and introduced into the loop with a defined time interval. Material was transported to the filter by the fluid and agglomerated there. The assessment of functionality of the filtering system was made by differential pressure between inlet and outlet of the filtering system and by mass of penetrated particles. It can be concluded that for the tested filtering system no penetration of insulation particles occurred.

Optimizing Maintenance Strategy of a Reactor Pressure Vessel Using 3D-CFD and FEM Based Probabilistic Pressurized Thermal Shock Analysis

2017 ◽  
Author(s):  
Xiaoyong Ruan ◽  
Toshiki Nakasuji ◽  
Kazunori Morishita
Keyword(s):  
Spatial Distribution ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Cooling System ◽  
Reactor Pressure Vessel ◽  
Pressurized Thermal Shock ◽  
Core Cooling ◽  
Reactor Pressure

The structural integrity of a reactor pressure vessel (RPV) is important for the safety of a nuclear power plant. When the emergency core cooling system (ECCS) is operated and the coolant water is injected into the RPV due to a loss-of-coolant accident (LOCA), the pressurized thermal shock (PTS) loading takes place. With the neutron irradiation, PTS loading may lead a RPV to fracture. Therefore, it is necessary to evaluate the performance of RPV during PTS loading to keep the reactor safety. In the present study, optimization of RPV maintenance is considered, where two different attempts are made to investigate the RPV integrity during PTS loading by employing the deterministic and probabilistic methodologies. For the deterministic integrity evaluation, 3D-CFD and finite element method (FEM) simulations are performed, and stress intensity factors (SIFs) are obtained as a function of crack position inside the RPV. As to the probabilistic integrity evaluation, on the other hand, a more accurate spatial distribution of SIF on the RPV is calculated. By comparing the distribution thus obtained with the fracture toughness included as a part of the master curve, the dependence of fracture probabilities on the position inside the RPV is obtained. Using the spatial distribution of fracture probabilities in RPV, the priority of the inspection and maintenance is finally discussed.

Conceptual Design and Feasibility Study for FAil-Safe Simple Economical SMR (FASES) System

2014 ◽  
Author(s):  
Ho Sik Kim ◽  
Hee Cheon No
Keyword(s):  
Conceptual Design ◽  
Feasibility Study ◽  
Cooling System ◽  
Heat Removal ◽  
Operating Conditions ◽  
Design Concepts ◽  
Decay Heat ◽  
Design Requirements ◽  
Core Cooling ◽  
Fail Safe

Because of high marketability of SMR, although many countries are trying to develop SMR, the SMR market is not formed yet. For early dominance of SMR market, the SMR system should be fail-safe, simple and economical. In order to develop FAil-safe Simple Economical SMR (FASES) system we applied the design characteristics of HTGRs into water-cooled reactors. In this study, we performed conceptual design and feasibility study for the FASES system. The feasibility study is focused on a thermal-hydraulic aspect in normal operating conditions and several accident conditions. The key design characteristic of the FASES system obtained from design concepts is to have sufficient decay heat removal capability even in the accidents involving extended station blackout (SBO), failure of passive decay heat removal system (PDHRS), and failure of emergency core cooling system (ECCS). Based on the design concepts, we could define several thermal-hydraulic design requirements. Then, we performed thermal-hydraulic analysis for feasibility study and proposed the specific design of the FASES system satisfying several design requirements.

scholarly journals Effect of Boric Acid Solubility in Steam on the Process of Mass Transfer during Emergency Cooling of VVER-1200 Nuclear Reactor

2021 ◽  
Vol 9 ◽  
pp. 07-15
Author(s):  
Md Rezouanul Kabir ◽  
Morozov A.V. ◽  
Md Saif Kabir
Keyword(s):  
Mass Transfer ◽  
Boric Acid ◽  
Nuclear Reactor ◽  
Water Saturation ◽  
Reactor Core ◽  
Saturation Temperature ◽  
Heat Removal ◽  
Solubility Limit ◽  
The Core ◽  
Core Cooling

The mechanisms of boric acid mass transfer in a VVER-1200 reactor core are studied in this work in the event of a major circulatory pipeline rupture and loss of all AC power. The VVER-1200's passive core cooling technology is made up of two levels of hydro accumulators. They use boric acid solution with a concentration of 16 g H3BO3/kg H2O to control the reactivity. Because of the long duration of the accident process, the coolant with high boron content starts boiling and steam with low concentration of boric acid departs the core. So, conditions could arise in the reactor for possible accumulation and subsequent crystallization of boric acid, causing the core heat removal process to deteriorate. Calculations were carried out to estimate the likelihood of H3BO3 build-up and subsequent crystallization in the core of the VVER reactor. According to the calculations, during emergency the boric acid concentration in the reactor core is 0.153 kg/ kg and 0.158 kg/kg in both the events of solubility of steam and without solubility of steam respectively and it does not exceed the solubility limit which is about 0.415 kg/kg at water saturation temperature. No precipitation of boric acid occurs within this time during the whole emergency process. Therefore, findings of the study can be used to verify whether the process of decay heat removal is affected or not.

Analysis of IP200 Severe Accident Process Response to SBO and Emergency Power Failure

2021 ◽  
Author(s):  
Zhenhang Zheng ◽  
Minjun Peng ◽  
Hao Yu ◽  
Yang Yang
Keyword(s):  
Cooling System ◽  
Cooling Capacity ◽  
Severe Accident ◽  
Mitigation Measures ◽  
Pressurized Water ◽  
Power Failure ◽  
The Core ◽  
Emergency Power ◽  
Core Cooling ◽  
High Level

Abstract Advanced SMRs such as the integrated pressurized water reactor IP200 use different design in the systems, structures, components from large reactors for achieving a high level of safety and reliability. In this thesis, the IP200 severe accident induced by the SBO and emergency power failure was modeled and analyzed using RELAP5 / SCDAP / MOD3.4 code. Based on the steady state calculation, which agrees well with designed values, the SBO accident for transient calculation is carried out. First, the case of the SBO accident without the passive core cooling system was calculated. The progression and scenario in the RPV was simulated and analyzed, including the transient response, cooling capacity and thermal-hydraulic characteristics and so on. Then, mitigation measures PRHRS and CMT were put in at four different time points when the core is began to uncovered, the core is completely uncovered, hydrogen is began to produced, and the molten pool is formed. The results show that putting in mitigation measures before the accident progresses to the point where the core starts to produce hydrogen can ensure that the core does not melt and avoid hydrogen risk.

Review and Status of the Gen-IV CANDU-SCWR Passive Moderator Core Cooling System

2008 ◽  
Author(s):  
H. F. Khartabil
Keyword(s):  
Heat Sink ◽  
Cooling System ◽  
Natural Circulation ◽  
Power Level ◽  
Heat Removal ◽  
Circulation System ◽  
The Past ◽  
Test Loop ◽  
Flow Oscillations ◽  
Core Cooling

Enhanced safety is an important priority in the development of Generation IV reactors, which can be accomplished through the use of improved passive heat removal systems. In CANDU® reactors, the separation between the low-pressure moderator and high-pressure coolant provides a unique passive heat sink for decay heat removal during accident scenarios. Methods for enhancing this passive heat sink for the GenIV CANDU-SCWR (supercritical water cooled reactor) have been under investigation for the past several years to support a “no core melt” reactor design concept (1, 2). Initially, to test feasibility, tests and analysis at AECL studied a full-height passive cooling loop and showed that a flashing-driven natural circulation system was possible in principle. However, flow oscillations were observed at low powers and could not be readily explained through analysis. While these oscillations were not considered to be detrimental to the heat removal capability, additional separate-effects experiments were conducted and causal mechanisms proposed for the oscillations. In addition, these separate effects tests suggested that oscillations could be avoided at any power level by suitable design. A new test loop with a more representative geometry was recently constructed and commissioned. Preliminary commissioning tests confirmed conclusions from the separate effects tests. In this paper, the new tests are compared to the past tests to explain the improved and more stable loop operation. This comparison suggests that a complete system coupled to an ultimate heat sink has the potential to improve loop operation even more by eliminating or significantly reducing flow oscillations at low powers. Plans for validating this conclusion will be provided.

Modification of the Core Cooling System of TRIGA 2000 Reactor

2010 ◽  
Author(s):  
Efrizon Umar ◽  
Rosalina Fiantini ◽  
Zaki Su’ud ◽  
A. Waris
Keyword(s):  
Cooling System ◽  
The Core ◽  
Core Cooling
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