Sensitivity and Uncertainty Information Incorporated Loss of Flow Accident Analyses for Research Reactors

2020 ◽  
Author(s):  
Tao Liu ◽  
Zeyun Wu
Keyword(s):  
Research Reactor ◽  
Safety Analysis ◽  
Primary Objective ◽  
Analysis Procedure ◽  
System Level ◽  
Normal Operation ◽  
Reactor Model ◽  
Research Reactors ◽  
Design Basis ◽  
Uncertainty Information

Abstract This paper outlines a system level safety analysis procedure for research reactors incorporating sensitivity and uncertainty components. The protected loss of flow (LOF) accident was selected as an exemplified design basis accident to demonstrate the analysis procedure. The conceptual NIST (National Institute of Standards and Technology) horizontally split-core based research reactor was adopted as a research reactor model in the study. Two system level dynamics codes, RELAP5-3D and PARET, were employed in this work in a comparison study manner. The primary objective of the present work is to demonstrate the analysis capability of integrating sensitivity and uncertainty information in addition to traditional predictions of the system code models for the study of the thermal-hydraulics (T/H) safety characteristics of research reactors under accidental transient scenarios. The canonical transient predictions on the LOF accident yielded from the two system codes mentioned above have demonstrated some noticeable yet acceptable discrepancies. To better understand the discrepancies observed in the simulations, sensitivity and uncertainty analyses were performed by coupling the RELAP5-3D model and the data analytic engines provided by the RAVEN framework developed by INL. The sensitivity information reveals the significances of key figure of merits such as the peak cladding temperature varies with different boundary and initial parameters in both normal operation and design basis transients. The uncertainty analysis informs the deviations of the responses contributed by the errors of various input components. Both the sensitivity and uncertainty information will be incorporated into a safety analysis framework as part of the safety characteristic predictions delivered by the framework.

Research reactor model for isotope production based on the design of VVER-440 reactors

Kerntechnik ◽  
2020 ◽  
Vol 85 (2) ◽  
pp. 105-108
Author(s):  
A. Terekhova ◽  
A. Mahdi ◽  
R. Zykova
Keyword(s):  
Research Reactor ◽  
Isotope Production ◽  
Reactor Model

Operator reliability study for Probabilistic Safety Analysis of an operating research reactor

2015 ◽  
Vol 80 ◽  
pp. 409-415 ◽  
Author(s):  
F. Mohamed ◽  
A. Hassan ◽  
R. Yahaya ◽  
I. Rahman ◽  
M. Maskin ◽  
...  
Keyword(s):  
Research Reactor ◽  
Safety Analysis ◽  
Reliability Study ◽  
Probabilistic Safety

scholarly journals A Safety Re-Evaluation of the AVR Pebble Bed Reactor Operation and Its Consequences for Future HTR Concepts

2008 ◽  
Author(s):  
Rainer Moormann
Keyword(s):  
Fission Product ◽  
End Of Life ◽  
Fission Products ◽  
Normal Operation ◽  
Primary Circuit ◽  
Pebble Bed ◽  
Pebble Bed Reactor ◽  
Hot Gas ◽  
Design Basis ◽  
Gas Tight

The AVR pebble bed reactor (46 MWth) was operated 1967–1988 at coolant outlet temperatures up to 990°C. Also because of a lack of other experience the AVR operation is a basis for future HTRs. This paper deals with insufficiently published unresolved safety problems of AVR and of pebble bed HTRs. The AVR primary circuit is heavily contaminated with dust bound and mobile metallic fission products (Sr-90, Cs-137) which create problems in current dismantling. The evaluation of fission product deposition experiments indicates that the end of life contamination reached several percent of a single core inventory. A re-evaluation of the AVR contamination is performed in order to quantify consequences for future HTRs: The AVR contamination was mainly caused by inadmissible high core temperatures, and not — as presumed in the past — by inadequate fuel quality only. The high AVR core temperatures were detected not earlier than one year before final AVR shut-down, because a pebble bed core cannot be equipped with instruments. The maximum core temperatures were more than 200 K higher than precalculated. Further, azimuthal temperature differences at the active core margin were observed, as unpredictable hot gas currents with temperatures > 1100°C. Despite of remarkable effort these problems are not yet understood. Having the black box character of the AVR core in mind it remains uncertain whether convincing explanations can be found without major experimental R&D. After detection of the inadmissible core temperatures, the AVR hot gas temperatures were strongly reduced for safety reasons. Metallic fission products diffuse in fuel kernel, coatings and graphite and their break through takes place in long term normal operation, if fission product specific temperature limits are exceeded. This is an unresolved weak point of HTRs in contrast to other reactors and is particularly problematic in pebble bed systems with their large dust content. Another disadvantage, responsible for the pronounced AVR contamination, lies in the fact that activity released from fuel elements is distributed in HTRs all over the coolant circuit surfaces and on graphitic dust and accumulates there. Consequences of AVR experience on future reactors are discussed. As long as pebble bed intrinsic reasons for the high AVR temperatures cannot be excluded they have to be conservatively considered in operation and design basis accidents. For an HTR of 400 MWth, 900°C hot gas temperature, modern fuel and 32 fpy the contaminations are expected to approach at least the same order as in AVR end of life. This creates major problems in design basis accidents, for maintenance and dismantling. Application of German dose criteria on advanced pebble bed reactors leads to the conclusion that a pebble bed HTR needs a gas tight containment even if inadmissible high temperatures as observed in AVR are not considered. However, a gas tight containment does not diminish the consequences of the primary circuit contamination on maintenance and dismantling. Thus complementary measures are discussed. A reduction of demands on future reactors (hot gas temperatures, fuel burn-up) is one option; another one is an elaborate R&D program for solution of unresolved problems related to operation and design basis accidents. These problems are listed in the paper.

scholarly journals Knowledge-Based Approach for System Level Electromagnetic Safety Analysis

2021 ◽  
Author(s):  
Lokesh Devaraj ◽  
Alastair R. Ruddle ◽  
Qazi Mashaal Khan ◽  
Alistair P. Duffy
Keyword(s):  
Safety Analysis ◽  
System Level ◽  
Knowledge Based

scholarly journals Innovative and safe supply of fuels for reactors

2020 ◽  
Vol 6 ◽  
pp. 40
Author(s):  
Stéphane Valance ◽  
Bruno Baumeister ◽  
Winfried Petry ◽  
Jan Höglund
Keyword(s):  
Power Plant ◽  
Nuclear Power ◽  
High Performance ◽  
Research Reactor ◽  
Fissile Material ◽  
Nuclear Fuels ◽  
Research Reactors ◽  
Enriched Uranium ◽  
Performance Research ◽  
And Training

Within the Euratom research and training program 2014–2018, three projects aiming at securing the fuel supply for European power and research reactors have been funded. Those three projects address the potential weaknesses – supplier diversity, provision of enriched fissile material – associated with the furbishing of nuclear fuels. First, the ESSANUF project, now terminated, resulted in the design and licensing of a fuel element for VVER-440 nuclear power plant manufactured by Westinghouse. The HERACLES-CP project aimed at preparing the conversion of high performance research reactor to low enriched uranium fuels by exploring fuels based on uranium-molybdenium. Finally, the LEU-FOREvER pursues the work initiated in HERACLES-CP, completing it by an exploration of the high-density silicide fuels, and including the diversification of fuel supplier for soviet designed European medium power research reactor. This paper describes the projects goals, structure and their achievements.

Steady-State Safety Analysis of Ghana Research Reactor-1 With Low-Enriched-Uranium Core

2020 ◽  
Vol 12 (5) ◽  
Author(s):  
Edward Shitsi ◽  
Prince Amoah ◽  
Emmanuel Ampomah-Amoako ◽  
Henry Cecil Odoi
Keyword(s):  
Steady State ◽  
Research Reactor ◽  
Flow Instability ◽  
Argonne National Laboratory ◽  
Safety Analysis ◽  
Nucleate Boiling ◽  
Pressurized Water ◽  
Safety Margins ◽  
Water Reactors ◽  
Enriched Uranium

Abstract Research reactors all over the world are expected to operate within certain safety margins just like pressurized water reactors and boiling water reactors. These safety margins mainly include onset of nucleate boiling ratio (ONBR), departure from nucleate boiling ratio (DNBR), and flow instability ratio (FIR) in addition to the maximum clad or fuel temperature and saturation temperature or boing point of the coolant inside the core of the reactor. This study carried out steady-state safety analysis of the Ghana Research Reactor-1 (GHARR-1) with low enriched uranium (LEU) core. Monte Carlo N-particle (MCNP) code was used to obtain radial and axial power peaking factors used as inputs in the preparation of the input file of plate temperature code of Argonne National Laboratory (PLTEMP/ANL code), which was then used to obtain the mentioned safety parameters of GHARR-1 with LEU core in this study. The data obtained on the ONBR were used to obtain the initiation of nucleate boiling boundary data with respect to the active length of the reactor core for various reactor powers. The obtained results for LEU core were also compared with that of the high enriched uranium (HEU) core. The results obtained show that the 34 kW GHARR-1 with LEU core is safe to operate just as the previous 30 kW HEU core was safe to operate.

scholarly journals Study on Typical Design Basis Conditions of HPR1000 With Nuclear Safety Analysis Code ATHLET

2020 ◽  
Vol 8 ◽  
Author(s):  
Xi Huang ◽  
Weixin Zong ◽  
Ting Wang ◽  
Zhikang Lin ◽  
Zhihao Ren ◽  
...  
Keyword(s):  
Safety Analysis ◽  
Nuclear Safety ◽  
Design Basis ◽  
Typical Design

scholarly journals Supporting infrastructures and research reactors: status, needs and international cooperation, IAEA ICERR (International CEntres based on Research Reactors) and IGORR (International Group on Research Reactors), FP7 and H2020 JHR access rights

2020 ◽  
Vol 6 ◽  
pp. 26
Author(s):  
Gilles Bignan ◽  
Jean-Yves Blanc
Keyword(s):  
Turning Point ◽  
Research Reactor ◽  
Project Proposal ◽  
Joint Project ◽  
Research Reactors ◽  
International Group ◽  
Access Rights ◽  
Evolving Context ◽  
Set Up ◽  
Under Construction

The panorama of research reactors in the world is at a turning point, with many old ones being shutdown, a very few new ones under construction and many newcomer countries interested to get access to one or to build one domestic research reactor or zero-power reactor. In this evolving context, several actions have been set up to answer this international collaboration need: the IAEA has launched the ICERR initiative, the OECD/NEA is proposing the P2M joint project proposal. In France, the Jules Horowitz Reactor (JHR), under construction at CEA Cadarache, within an International Consortium, will be one of the few tools available for the industry and research in the next decades. The paper presents some update of its construction, its experimental capacities and the European support through FP7 and H2020 tools. This paper provides also some insights of international tools (ICERR, P2M) and about the International Group on Research Reactors (IGORR) and how they complement or interact with the JHR.

Sign in / Sign up

Export Citation Format

Share Document