Stability analysis of Tesla valve based natural circulation loop for decay heat removal in nuclear power plants

The Korea Atomic Energy Research Institute (KAERI) is developing a Generation IV sodium-cooled fast reactor design equipped with a passive decay heat removal circuit (PDRC), which is a unique safety system in the design. The performance of the PDRC system is quite important for the safety in a simple system transient and also in an accident condition. In those situations, the heat generated in the core is transported to the ambient atmosphere by natural circulation of the PDRC loop. It is essential to investigate the performance of its heat removal capability through experiments for various operational conditions. Before the main experiments, KAERI is performing numerical studies for an evaluation of the performance of the PDRC system. First, the formation of a stable natural circulation is numerically simulated in a sodium test loop. Further, the performance of its heat removal at a steady state condition and at a transient condition is evaluated with the real design configuration in the KALIMER-600. The MARS-LMR code, which is developed for the system analysis of a liquid metal-cooled fast reactor, is applied to the analysis. In the present study, it is validated that the performance of natural circulation loop is enough to achieve the required passive heat removal for the PDRC. The most optimized modeling methodology is also searched for using various modeling approaches.

Download Full-text

In-Calandria Retention of Corium in PHWR: Experimental Investigation and Remaining Issues

Journal of Nuclear Engineering and Radiation Science ◽

10.1115/1.4035691 ◽

2017 ◽

Vol 3 (2) ◽

Cited By ~ 1

Author(s):

Sumit V. Prasad ◽

A. K. Nayak

Keyword(s):

Experimental Investigation ◽

Nuclear Power ◽

Power Plants ◽

Nuclear Power Plants ◽

Heat Removal ◽

Public Domain ◽

Severe Accident ◽

Fukushima Accident ◽

Decay Heat ◽

Critical Issues

After the Fukushima accident, the public has expressed concern regarding the safety of nuclear power plants. This accident has strengthened the necessity for further improvement of safety in the design of existing and future nuclear power plants. Pressurized heavy water reactors (PHWRs) have a high level of defense-in-depth (DiD) philosophy to achieve the safety goal. It is necessary for designers to demonstrate the capability of decay heat removal and integrity of containment in a PHWR reactor for prolonged station blackout to avoid any release of radioactivity in public domain. As the design of PHWRs is distinct, its calandria vessel (CV) and vault cooling water offer passive heat sinks for such accident scenarios and submerged calandria vessel offers inherent in-calandria retention (ICR) features. Study shows that, in case of severe accident in PHWR, ICR is the only option to contain the corium inside the calandria vessel by cooling it from outside using the calandria vault water to avoid the release of radioactivity to public domain. There are critical issues on ICR of corium that have to be resolved for successful demonstration of ICR strategy and regulatory acceptance. This paper tries to investigate some of the critical issues of ICR of corium. The present study focuses on experimental investigation of the coolability of molten corium with and without simulated decay heat and thermal behavior of calandria vessel performed in scaled facilities of an Indian PHWR.

Download Full-text

Mitigation of Temperature Gradients in the Buffer Tank of a Natural Circulation Loop Simulating Passive Decay Heat Removal in a Pool Type Reactor

Heat Transfer Engineering ◽

10.1080/01457632.2019.1628478 ◽

2019 ◽

Vol 41 (15-16) ◽

pp. 1315-1340

Author(s):

Chiara Spaccapaniccia ◽

Philippe Planquart ◽

Jean-Marie Buchlin

Keyword(s):

Natural Circulation ◽

Heat Removal ◽

Temperature Gradients ◽

Circulation Loop ◽

Decay Heat ◽

Type Reactor ◽

Natural Circulation Loop ◽

Buffer Tank ◽

Pool Type

Download Full-text

Recent Reactor Head Drop Analysis for Complying NUREG-0612 Appendix A Guidelines

Volume 7: Operations, Applications and Components ◽

10.1115/pvp2008-61880 ◽

2008 ◽

Author(s):

Phuong H. Hoang ◽

Mohammad Amin ◽

Chee W. Mak

Keyword(s):

Nuclear Power ◽

Power Plants ◽

Nuclear Power Plants ◽

Heat Removal ◽

Federal Regulations ◽

Load Drop ◽

Safe Handling ◽

Decay Heat ◽

Parameters Analysis ◽

Heavy Loads

The USNRC staff provided guidelines in NUREG-0612 for the control of heavy loads to meet the requirements of the Code of Federal Regulations (10 CFR) Sections 50.59 and 50.71(e) relate to the safe handling of heavy loads and load drop analyses to assure fuel assembly integrity and to permit continued decay heat removal in nuclear power plants. The USNRC staff considered plants that had installed single-failure-proof cranes or had completed load drop analyses conforming with the guidelines of Appendix A to NUREG-0612 would remain in conformance with the safe handling guidelines for lifting reactor head during the plant refueling operation. Reactor head drop analyses have been performed recently for a number of US nuclear power plants and have been reviewed by the USNRC staff for complying with the guidelines of Appendix A to NUREG-0612. This paper provides an overview of the methodology, parameters, analysis results and acceptance criteria used in these recent works.

Download Full-text

INVESTIGATION OF THE DEPOSITION OF INSOLUBLE AEROSOLS ON THE HEAT EXCHANGE SURFACES OF THE SYSTEM OF PASSIVE HEAT REMOVAL FROM THE CONTAINMENT DURING ACCIDENTS AT NUCLEAR POWER PLANTS

Technology of technosphere safety ◽

10.25257/tts.2019.1.83.24-33 ◽

2019 ◽

Vol 83 ◽

pp. 24-33 ◽

Cited By ~ 2

Author(s):

A. A. Fiskov ◽

◽

V. V. Bezlepkin ◽

S. E. Semashko. ◽

V. M. Pogrebenkov ◽

...

Keyword(s):

Heat Exchange ◽

Nuclear Power ◽

Power Plants ◽

Nuclear Power Plants ◽

Heat Removal

Download Full-text

Seismic safety evaluation during site selection for the nuclear power plants in Bangladesh

Nuclear Energy and Technology ◽

10.3897/nucet.4.31873 ◽

2018 ◽

Vol 4 (4) ◽

pp. 251-256 ◽

Cited By ~ 1

Author(s):

Sergey Shcheklein ◽

Ismail Hossain ◽

Mohammad Akbar ◽

Vladimir Velkin

Keyword(s):

Power Plant ◽

Nuclear Power Plant ◽

Nuclear Power ◽

Power Plants ◽

Nuclear Power Plants ◽

Spent Nuclear Fuel ◽

Heat Removal ◽

Good Time ◽

Diagnostics System ◽

Seismic Impacts

Bangladesh lies in a tectonically active zone. Earlier geological studies show that Bangladesh and its adjoining areas are exposed to a threat of severe earthquakes. Earthquakes may have disastrous consequences for a densely populated country. This dictates the need for a detailed analysis of the situation prior to the construction of nuclear power plant as required by the IAEA standards. This study reveals the correlation between seismic acceleration and potential damage. Procedures are presented for investigating the seismic hazard within the future NPP construction area. It has been shown that the obtained values of the earthquake’s peak ground acceleration are at the level below the design basis earthquake (DBE) level and will not lead to nuclear power plant malfunctions. For the most severe among the recorded and closely located earthquake centers (Madhupur) the intensity of seismic impacts on the nuclear power plant site does not exceed eight points on the MSK-64 scale. The existing predictions as to the possibility of a super-earthquake with magnitude in excess of nine points on the Richter scale to take place on the territory of the country indicate the necessity to develop an additional efficient seismic diagnostics system and to switch nuclear power plants in good time to passive heat removal mode as stipulated by the WWER 3+ design. A conclusion is made that accounting for the predicted seismic impacts in excess of the historically recorded levels should be achieved by the establishment of an additional efficient seismic diagnostics system and by timely switching the nuclear power plants to passive heat removal mode with reliable isolation of the reactor core and spent nuclear fuel pools.

Download Full-text

Simulation of Steady States of Helium Loop at Long Time Scenarios

MATEC Web of Conferences ◽

10.1051/matecconf/202032801009 ◽

2020 ◽

Vol 328 ◽

pp. 01009

Author(s):

František Világi ◽

Branislav Knížat ◽

Róbert Olšiak ◽

Marek Mlkvik ◽

Peter Mlynár ◽

...

Keyword(s):

Nuclear Power ◽

Power Plants ◽

Nuclear Reactor ◽

Natural Circulation ◽

Heat Removal ◽

Iteration Process ◽

Heat Accumulation ◽

Proposed Model ◽

Long Time ◽

Convection Cooling

The natural circulation helium loop is an experimental facility designed for the research of the possibility of utilizing natural convection cooling for the case of decay heat removal from a fast nuclear reactor. This concept would bring an improved automated safety system for future nuclear power plants operating a gas-cooled reactor. The article presents a new possibility of direct use of energy conservation laws in a 1D simulation of natural circulation loops. The calculation is performed by a triple iteration process, nested into each other. The results of the calculations showed good agreement with the measurements at steady state. A calculation with the proposed model at unsteady state is not yet possible, especially because of the exclusion of heat accumulation into the material.

Download Full-text

Effect of Coolant Inventories and Parallel Loop Interconnections on the Natural Circulation in Various Heat Transport Systems of a Nuclear Power Plant during Station Blackout

Science and Technology of Nuclear Installations ◽

10.1155/2008/458316 ◽

2008 ◽

Vol 2008 ◽

pp. 1-11 ◽

Cited By ~ 6

Author(s):

Avinash J. Gaikwad ◽

P. K. Vijayan ◽

Sharad Bhartya ◽

Kannan Iyer ◽

Rajesh Kumar ◽

...

Keyword(s):

Power Plant ◽

Nuclear Power Plant ◽

Heat Transport ◽

Nuclear Power ◽

Natural Circulation ◽

Heat Removal ◽

Transport Systems ◽

Critical Examination ◽

Decay Heat ◽

Station Blackout

Provision of passive means to reactor core decay heat removal enhances the nuclear power plant (NPP) safety and availability. In the earlier Indian pressurised heavy water reactors (IPHWRs), like the 220 MWe and the 540 MWe, crash cooldown from the steam generators (SGs) is resorted to mitigate consequences of station blackout (SBO). In the 700 MWe PHWR currently being designed an additional passive decay heat removal (PDHR) system is also incorporated to condense the steam generated in the boilers during a SBO. The sustainability of natural circulation in the various heat transport systems (i.e., primary heat transport (PHT), SGs, and PDHRs) under station blackout depends on the corresponding system's coolant inventories and the coolant circuit configurations (i.e., parallel paths and interconnections). On the primary side, the interconnection between the two primary loops plays an important role to sustain the natural circulation heat removal. On the secondary side, the steam lines interconnections and the initial inventory in the SGs prior to cooldown, that is, hooking up of the PDHRs are very important. This paper attempts to open up discussions on the concept and the core issues associated with passive systems which can provide continued heat sink during such accident scenarios. The discussions would include the criteria for design, and performance of such concepts already implemented and proposes schemes to be implemented in the proposed 700 MWe IPHWR. The designer feedbacks generated, and critical examination of performance analysis results for the added passive system to the existing generation II & III reactors will help ascertaining that these safety systems/inventories in fact perform in sustaining decay heat removal and augmenting safety.

Download Full-text