scholarly journals HTGR spent fuel element decay heat and source term analysis

1977 ◽  
Author(s):  
R.E. Sund ◽  
D.E. Strong ◽  
B.A. Engholm
Keyword(s):  
Fuel Element ◽  
Source Term ◽  
Spent Fuel ◽  
Decay Heat ◽  
Term Analysis

Source Term Analysis Considering B4C/Steel Interaction and Oxidation During Severe Accidents

2017 ◽  
Author(s):  
Jun Ishikawa ◽  
Tomoyuki Sugiyama ◽  
Yu Maruyama
Keyword(s):  
Source Term ◽  
Severe Accident ◽  
Water Pool ◽  
Energy Agency ◽  
Severe Accidents ◽  
Station Blackout ◽  
Carbon Dioxide Co2 ◽  
Accident Sequences ◽  
Term Analysis

The Japan Atomic Energy Agency (JAEA) is pursuing the development and application of the methodologies on fission product (FP) chemistry for source term analysis by using the integrated severe accident analysis code THALES2. In the present study, models for the eutectic interaction of boron carbide (B4C) with steel and the B4C oxidation were incorporated into THALES2 code and applied to the source term analyses for a boiling water reactor (BWR) with Mark-I containment vessel (CV). Two severe accident sequences with drywell (D/W) failure by overpressure initiated by loss of core coolant injection (TQUV sequence) and long-term station blackout (TB sequence) were selected as representative sequences. The analyses indicated that a much larger amount of species from the B4C oxidation was produced in TB sequence than TQUV sequence. More than a half of carbon dioxide (CO2) produced by the B4C oxidation was predicted to dissolve into the water pool of the suppression chamber (S/C), which could largely influence pH of the water pool and consequent formation and release of volatile iodine species.

scholarly journals Cooling Period Calculation of Evolutionary Power Reactor Spent Fuel for Dry Management Safety

2016 ◽  
pp. 17-21
Author(s):  
M. I. Youssef ◽  
G. F. Sultan ◽  
F. Morsi Hassan
Keyword(s):  
Heat Load ◽  
Power Reactor ◽  
Computer Code ◽  
Normal Operation ◽  
Spent Fuel ◽  
Cooling Period ◽  
Decay Heat ◽  
Safety Margins ◽  
Epr Parameters

The calculation of the evolutionary power reactor (EPR) spent fuel (SF) cooling period (CP) was performed. The CP was determined by comparing the heat load of a cask with the calculated value of EPR decay heat (DH). The EPR DH was calculated by the ORIGEN computer code based on the EPR parameters. For conservatively study, the EPR and ORIGEN parameters that lead to higher DH values were selected and safety margins were considered. The fitting tool was utilized in the calculation of CP to overcome the ORIGEN limitation. The resultant values of CP will maintain the peak cladding temperature (PCT) of SF lower than 400°C during storage, transport, and disposal. The results show that -for normal operation- the SF of EPR should stay in the pool at least 4.75 years before it is loaded to the passively cooled dry casks.

Evaluation of source term parameters for spent fuel disposal in foreign countries

2016 ◽  
Vol 23 (1) ◽  
pp. 31-54
Author(s):  
Masanobu NAGATA ◽  
Takahiro CHIKAZAWA ◽  
Kuniaki AKAHORI ◽  
Akira KITAMURA ◽  
Yukio TACHI
Keyword(s):  
Source Term ◽  
Spent Fuel ◽  
Foreign Countries

scholarly journals Decay Heat of Fast Reactor Spent Fuel.

1999 ◽  
Vol 41 (9) ◽  
pp. 946-953 ◽  
Author(s):  
Takafumi AOYAMA ◽  
Tadahiko TORIMARU ◽  
Akihiro YOSHIDA ◽  
Yoshio ARII ◽  
Soju SUZUKI
Keyword(s):  
Fast Reactor ◽  
Spent Fuel ◽  
Decay Heat

Calculation of decay heat rate in VVER-1000 spent fuel pool including uncertainties

2020 ◽  
Vol 366 ◽  
pp. 110754
Author(s):  
Danail Hristov ◽  
Plamen V. Petkov ◽  
Ivo Naev
Keyword(s):  
Heat Rate ◽  
Spent Fuel ◽  
Decay Heat ◽  
Spent Fuel Pool

Source Term Analysis of Tritium in HTR-10

2017 ◽  
Vol 71 (4) ◽  
pp. 671-678 ◽  
Author(s):  
Yi Xu ◽  
Hong Li ◽  
Feng Xie ◽  
Jianzhu Cao ◽  
Jiejuan Tong
Keyword(s):  
Source Term ◽  
Term Analysis

Verification of Source Term Analysis System for Decommissioning Wastes From a CANDU Reactor

2010 ◽  
Author(s):  
Dong-Keun Cho ◽  
GwangMin Sun ◽  
JongWon Choi ◽  
Donghyeun Hwang ◽  
Hak-Soo Kim ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Source Term ◽  
Core Model ◽  
Code System ◽  
Structural Components ◽  
Evaluation Tool ◽  
Chemical Assay ◽  
Term Evaluation ◽  
Analysis System ◽  
Term Analysis

There are now twenty commercial nuclear power reactors operating as of May 2010 in South Korea. As nuclear capacity becomes higher and installations age, the Korean government and industry have launched R&D to estimate appropriate decommissioning costs of power reactors. In this paper, MCNP/ORIGEN2 code system which is being developed as a source term evaluation tool was verified by comparing the estimated nuclide inventory from MCNP/ORIGEN2 simulation with the measured nuclide inventory from chemical assay in an irradiated pressure tube discharged from Wolsong Unit 1 in 1994. Equilibrium core model of Wolsoung unit 1 was used as a neutron source to activate in-core and ex-core structural components. As a result, the estimated values from the analysis system agreed with measured data within 20% difference. Therefore, it can be concluded that MCNP/ORIGEN system could be a reliable tool to estimate source terms of decommissioning wastes from CANDU reactor, although this system assumes constant flux irradiation and snapshot equilibrium core model as a reference core.

Air Cooling Temperature Analysis of Spent Fuel

2005 ◽  
Author(s):  
A. L. Laursen ◽  
F. J. Moody ◽  
J. C. Law
Keyword(s):  
Nuclear Reactor ◽  
Spent Nuclear Fuel ◽  
Natural Circulation ◽  
Maximum Temperature ◽  
Air Cooling ◽  
Spent Fuel ◽  
Decay Heat ◽  
Spent Fuel Pool ◽  
Large Water ◽  
The Relationship

Spent nuclear fuel is currently being stored at nuclear reactor sites. The spent fuel removed from the reactor is first placed in a large water pool to remove the initial decay heat. After several years, when the decay heat has dropped below a set level, the fuel is moved into concrete storage casks where natural circulation continues the cooling process. The purpose of this report is to predict, using a simplified analysis, how hot the fuel rods get when cooled by air in the cask. The increase in temperature and the decrease in density cause a chimney effect in the cask. This paper presents an analytical method of obtaining maximum fuel clad temperature in the cask. A non-dimensional model is derived, which is used to calculate the entrance and exit air velocities of the cask. The relationship between these velocities and the temperature used to obtain the maximum fuel clad temperature. A numerical scheme used to predict the maximum temperature is presented here and the results are compared to the analytical model. Both methods yielded corroborating results for fuel placed in the casks after spending similar amounts of time in a spent fuel pool.

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