Evaluation of Dump Tank Coolability in PHWRs During Late-Phase Severe Accident

In some of the older design of pressurized heavy water reactors (PHWRs), such as in Rajasthan Atomic Power Station (RAPS) and Madras Atomic Power Station (MAPS), in case of a severe accident, the debris/corium may cause failure of the dump port of calandria and relocate into the dump tank. The sensible and decay heat of debris/corium makes the heavy water in dump tank to boil off leaving the dry debris in dump tank. The dry debris remelt with time and the molten corium, thus, formed has the potential to breach the dump tank and move into the containment cavity, which is highly undesirable. Hence, as an accident management strategy, water is being flooded outside the dump tank using fire water hook-up lines to remove the heat from corium to cool and stabilize it and terminate the accident progression, similar to in vessel retention. However, the question is “is the molten corium coolable by this technique.” The coolability of the molten corium in dump tank as in the reactor is assessed by conducting experiments in a scaled facility using a simulant material having comparable thermophysical properties with that of corium. Melting of dry debris resting on dump tank bottom marks the beginning of the experimental investigation for present analysis. Decay heat is simulated by a set of immersed heaters inside the melt. Temperature profiles at different locations in dump tank and in the melt pool are obtained as a function of time to demonstrate the coolability with decay heat. Large temperature gradient is observed within the corium, involving high melt center temperature and low tank wall temperature suggesting formation of crust which insulates the dump tank wall from hot corium.

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Experimental Study on Melt Coolability Capability of Calandria Vault Water During Severe Accident in Indian PHWRs for Prolonged Duration

Journal of Nuclear Engineering and Radiation Science ◽

10.1115/1.4039636 ◽

2018 ◽

Vol 4 (3) ◽

Cited By ~ 2

Author(s):

Sumit V. Prasad ◽

A. K. Nayak

Keyword(s):

Heat Transfer ◽

Single Phase ◽

Convection Heat Transfer ◽

Severe Accident ◽

Transfer Coefficients ◽

Decay Heat ◽

Prolonged Duration ◽

Water Reactors ◽

Vessel Integrity ◽

Inner Vessel

The present experimental investigation in a scaled facility of an Indian pressurized heavy water reactors (PHWRs) is focused on the heat transfer behavior from the calandria vessel (CV) to the calandria vault during a prolonged severe accident condition in the presence of decay heat. The transient heat transfer simulates the conditions from single phase to boiling in the calandria vault water, partial uncovery of the CV due to boil off of water in the vault, and refill of calandria vault. Molten borosilicate glass was used as the simulant due to its comparable heat transfer characteristics similar to prototypic material. About 60 kg of the molten material was poured into the test section at about 1100 °C. Decay heat in the melt pool was simulated by using high watt cartridge type heaters. The temperature distributions inside the molten pool across the CV wall thickness and vault water were measured for prolonged period which can be divided into various phases, viz., single phase natural convection heat transfer in calandria vault, boiling heat transfer in calandria vault, partial uncovery of CV, and refilling calandria vault. Experimental results showed that once the crust formed, the inner vessel temperature remained very low and vessel integrity maintained. Even boiling of calandria vault water and uncovery of CV had negligible effect on melt, CV, and vault water temperature. The heat transfer coefficients on outer vessel surface were obtained and compared with various conditions.

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The A-1 station, Czechoslovakia's first atomic power station, with the KS-150 heavy-water reactor (development and construction)

Soviet Atomic Energy ◽

10.1007/bf01117802 ◽

1974 ◽

Vol 36 (2) ◽

pp. 138-150

Author(s):

V. M. Abramov ◽

B. B. Baturov ◽

N. V. Bogdanov ◽

V. F. Zelenskii ◽

V. E. Ivanov ◽

...

Keyword(s):

Heavy Water ◽

Atomic Power Station ◽

Power Station ◽

Water Reactor ◽

Heavy Water Reactor

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Numerical Simulation of Concrete Ablation and Corium Cooling for Molten Corium-Concrete Interaction (MCCI)

Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation ◽

10.1115/icone2020-16388 ◽

2020 ◽

Author(s):

Zhichun Xu ◽

Yapei Zhang ◽

G. H. Su ◽

Wenxi Tian ◽

Suizheng Qiu

Keyword(s):

Water Injection ◽

Reactor Core ◽

Reactor Vessel ◽

Severe Accident ◽

Sensitivity Analyses ◽

Core Material ◽

Melt Temperature ◽

Concrete Type ◽

Depth Study ◽

Water Reactors

Abstract In a postulated severe accident situation in Light Water Reactors (LWRs), if the core fuel cannot be effectively cooled, the reactor core material will be heated and form a molten corium in the lower head. When the lower plenum of the reactor vessel fails, the molten corium may flow into the cavity under the reactor vessel and react with the concrete. This process, known as Molten Corium Concrete Interaction (MCCI), is characterized by concrete ablation and oxidation of metal in the corium, both of which produce a large amount of combustible and non-condensable gases, threatening the integrity of the containment. Thus in-depth study of the characteristics of concrete ablation and corium cooling have great significance. In the present study, an MCCI analysis code, MOQUICO (molten corium concrete interaction and corium cooling code, QUI means quintic) has been developed. The MACE M3b and OECD/MCCI CCI-3 tests were analyzed to validate the developed code. The melt temperature, axial and radial ablation depths, upward heat flux were calculated and were in good agreement with the experimental measurements, which proved that the code is capable of simulating MCCI and related phenomena of LWRs. Sensitivity analyses on the factors of decay heat, concrete type and water injection moment were performed and analyzed.

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Benchmarking Severe Accident Computer Codes for Heavy Water Reactor Applications

Journal of Nuclear Engineering and Radiation Science ◽

10.1115/1.4035726 ◽

2017 ◽

Vol 3 (2) ◽

Cited By ~ 2

Author(s):

Thambiayah Nitheanandan ◽

X. Cao ◽

J.-H. Choi ◽

D. Dupleac ◽

D.-H. Kim ◽

...

Keyword(s):

Boundary Conditions ◽

International Atomic Energy Agency ◽

Heavy Water ◽

Severe Accident ◽

Energy Agency ◽

Heavy Water Reactor ◽

Computer Codes ◽

Water Reactors ◽

Definition Of ◽

Selection Of

The International Atomic Energy Agency (IAEA) organized a coordinated research project (CRP) on “Benchmarking Severe Accident Computer Codes for Heavy Water Reactors (HWR) Applications,” (IAEA TECDOC Series No. 1727), and the activity was completed in 2012. This paper summarizes the results from the CRP: the selection of a severe accident sequence, definition of appropriate geometrical and boundary conditions, benchmarking code analyses, comparison of the code results, evaluation of the capabilities of existing computer codes to predict important severe accident phenomena, and suggestions for code improvements and/or new experiments to reduce uncertainties.

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FISSION-PRODUCT RELEASES FROM A PHWR TERMINAL DEBRIS BED

CNL Nuclear Review ◽

10.12943/cnr.2015.00056 ◽

2016 ◽

Vol 5 (1) ◽

pp. 95-105 ◽

Cited By ~ 4

Author(s):

M.J. Brown ◽

D.G. Bailey

Keyword(s):

Fission Product ◽

Heavy Water ◽

Fission Products ◽

Severe Accident ◽

Water Reactor ◽

Decay Heat ◽

The Core ◽

Heavy Water Reactor ◽

Release Model ◽

Tank Water

During an unmitigated severe accident in a pressurized heavy water reactor (PHWR) with horizontal fuel channels, the core may disassemble and relocate to the bottom of the calandria vessel. The resulting heterogeneous in-vessel terminal debris bed (TDB) would likely be quenched by any remaining moderator, and some of the decay heat would be conducted through the calandria vessel shell to the surrounding reactor vault or shield tank water. As the moderator boiled off, the solid debris bed would transform into a more homogeneous molten corium pool located between top and bottom crusts. Until recently, the severe accident code MAAP-CANDU assumed that unreleased volatile and semi-volatile fission products remained in the TDB until after calandria vessel failure, due to low diffusivity through the top crust and the lack of gases or steam to flush released fission products from the debris. However, national and international experimental results indicate this assumption is unlikely; instead, high- and medium-volatility fission products would be released from a molten debris pool, and their volatility and transport should be taken into account in TDB modelling. The resulting change in the distribution of fission products within the reactor and containment, and the associated decay heat, can have significant effects upon the progression of the accident and fission-product releases to the environment. This article describes a postulated PHWR severe accident progression to generate a TDB and the effects of fission-product releases from the terminal debris, using the simple release model in the MAAP-CANDU severe accident code. It also provides insights from various experimental programs related to fission-product releases from core debris, and their applicability to the MAAP-CANDU TDB model.

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Analysis of Possibility of Pressure Tube Cold Pressurization During ECCS Injection for a Small Break LOCA

Journal of Pressure Vessel Technology ◽

10.1115/1.1465438 ◽

2002 ◽

Vol 124 (4) ◽

pp. 483-486 ◽

Cited By ~ 1

Author(s):

D. Mukhopadhyay ◽

S. K. Gupta ◽

V. Venkat Raj

Keyword(s):

Control System ◽

System Design ◽

Heavy Water ◽

Pressure Control ◽

Severe Accident ◽

Loss Of Coolant Accident ◽

Loss Of Coolant ◽

Pressure Control System ◽

Safe Limits ◽

Water Reactors

ECCS is designed to keep the reactor fuel temperatures within safe limits. The paper describes an additional criterion for Indian pressurized heavy water reactors (IPHWRs) evolving from the need to avoid a small break loss of cooling accident (LOCA) developing into a more severe accident. During a small break loss of coolant accident (LOCA) in PHWRs, the hydro-accumulators ride on the system and inject emergency coolant. The atmospheric steam discharge valves (ASDVs) open and cool the system due to energy discharge. In addition, the pressure control system tends to maintain the pressure. Depending on the system design, this could lead to cold pressurization of the system. This paper examines this issue.

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