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 Fission Product Transport and Source Terms in HTRs: Experience from AVR Pebble Bed Reactor

2008 ◽  
Vol 2008 ◽  
pp. 1-14 ◽  
Author(s):  
Rainer Moormann
Keyword(s):  
Fission Product ◽  
Dominant Role ◽  
Fission Products ◽  
Normal Operation ◽  
Pebble Bed ◽  
Pebble Bed Reactor ◽  
Pebble Bed Reactors ◽  
Maintenance Work ◽  
Air Ingress ◽  
Gas Tight

Fission products deposited in the coolant circuit outside of the active core play a dominant role in source term estimations for advanced small pebble bed HTRs, particularly in design basis accidents (DBA). The deposited fission products may be released in depressurization accidents because present pebble bed HTR concepts abstain from a gas tight containment. Contamination of the circuit also hinders maintenance work. Experiments, performed from 1972 to 88 on the AVR, an experimental pebble bed HTR, allow for a deeper insight into fission product transport behavior. The activity deposition per coolant pass was lower than expected and was influenced by fission product chemistry and by presence of carbonaceous dust. The latter lead also to inconsistencies between Cs plate out experiments in laboratory and in AVR. The deposition behavior of Ag was in line with present models. Dust as activity carrier is of safety relevance because of its mobility and of its sorption capability for fission products. All metal surfaces in pebble bed reactors were covered by a carbonaceous dust layer. Dust in AVR was produced by abrasion in amounts of about 5 kg/y. Additional dust sources in AVR were ours oil ingress and peeling of fuel element surfaces due to an air ingress. Dust has a size of about 1  m, consists mainly of graphite, is partly remobilized by flow perturbations, and deposits with time constants of 1 to 2 hours. In future reactors, an efficient filtering via a gas tight containment is required because accidents with fast depressurizations induce dust mobilization. Enhanced core temperatures in normal operation as in AVR and broken fuel pebbles have to be considered, as inflammable dust concentrations in the gas phase.

scholarly journals Chemical Forms of Important Fission Products in Primary Circuit of HTR-PM under Conditions of Normal Operation and Overpressure and Water Ingress Accidents: A Study with a Chemical Thermodynamics Approach

2019 ◽  
Vol 2019 ◽  
pp. 1-12
Author(s):  
Chuan Li ◽  
Wenqian Li ◽  
Lifeng Sun ◽  
Haoyu Xing ◽  
Chao Fang
Keyword(s):  
High Temperature ◽  
Fission Products ◽  
Point Of View ◽  
Normal Operation ◽  
Primary Circuit ◽  
Chemical Thermodynamics ◽  
Chemical Forms ◽  
Operational Conditions ◽  
Water Ingress ◽  
High Temperature Gas

The chemical forms of important fission products (FPs) in the primary circuit are essential to the source term analysis of high-temperature gas-cooled reactors because the volatility, transfer, and diffusion of these radionuclides are significantly influenced by their chemical forms. Through chemical reactions with gaseous impurities in the primary circuit, these FPs exist in diverse chemical forms, which vary under different operational conditions. In this paper, the chemical forms of cesium (Cs), strontium (Sr), silver (Ag), iodine (I), and tritium in the primary circuit of the Chinese pebble-bed modular high-temperature gas-cooled reactor (HTR-PM) under normal conditions and accident conditions (overpressure and water ingress accident) are studied with chemical thermodynamics. The results under normal conditions show that Cs exists mainly in the form of Cs2CO3 at 250°C and gaseous form at 750°C, and for I and Ag, Ag3I3 and Ag convert to gaseous CsI and AgO, respectively, with increasing temperature, while SrCO3 is the only main kind of compound for Sr. It is also observed that new compounds are generated under accidents: I exists in HI form when a water ingress accident occurs. Regarding tritium, the chemical forms of FPs change little, but compounds need higher temperature to convert. Furthermore, hazard of some FPs in different chemical forms is also discussed comprehensively in this paper. This study is significant for understanding the chemical reaction mechanisms of FPs in an HTR-PM, and furthermore it may provide a new point of view to analyze the interaction between FPs and structural materials in reactor as well as their hazards.

Water-Ingress Accident of the 250MW Pebble-Bed Modular High Temperature Gas-Cooled Reactor

2009 ◽  
Author(s):  
Zheng Yanhua ◽  
Shi Lei
Keyword(s):  
High Temperature ◽  
Safety Valve ◽  
Primary Circuit ◽  
Fuel Temperature ◽  
Pebble Bed ◽  
Water Ingress ◽  
Design Basis ◽  
Design Basis Accident ◽  
Water Gas ◽  
High Temperature Gas

Water-ingress accident, caused by the steam generator heating tube rupture of a high temperature gas-cooled reactor, will introduce a positive reactivity to lead the nuclear power increase rapidly, as well as the chemical reaction of graphite fuel elements and reflector structure material with steam. Increase of the primary circuit pressure may result in the opening of the safety valve, which will cause the release of radioactive isotopes and flammable water gas. The analysis of such an important and particular accident is significant for verifying the inherent safety characteristics of the pebble-bed modular high temperature gas-cooled reactor. Based on the preliminary design of the 250MW Pebble-bed Modular High Temperature Gas-cooled Reactor (HTR-PM), the design basis accident of double-ended guillotine break of a heating tube has been analyzed by using TINTE, which is a special transient analysis program for high temperature gas-cooled reactors. Some safety relevant concerns, such as the fuel temperature and primary loop pressure, the graphite corrosion inventory, the water gas releasing amount, as well as the natural convection influence under the condition of the failure of the blower flaps shut down, have been studied in detail. The calculation result of the design basis accident indicates that, the maximal possible water ingress amount is less than 600 kg and the maximal fuel temperature keeps far below the design limitation of 1620°C. The result also shows that the slight amount of graphite corrosion will not damage the reactor structure and the fuel element, and there is no potential explosive risk caused by the opening of the safety valve.

Analysis of Radioactive Source Term for Modular HTGR During Normal Operation

2010 ◽  
Author(s):  
Jianzhu Cao ◽  
Tao Liu ◽  
Yuanyu Wu ◽  
Hong Li ◽  
Yuanzhong Liu
Keyword(s):  
Source Term ◽  
Noble Gas ◽  
Fission Products ◽  
Radioactive Source ◽  
Normal Operation ◽  
Design Parameters ◽  
Primary Circuit ◽  
Secondary Circuit ◽  
Activation Products ◽  
Basic Prediction

The methods of radioactive source term analysis are introduced in detail for the modular high temperature gas cooled reactor in China. Radioactive fission products and activation products produced in the reactor are described. For fission products, the emphasis is on the process from production through release to the environment for noble gas, iodine and long-lived metallic nuclides. For activation products, it mainly introduces the behaviors of H-3 and C-14. Especially the permeation process from primary circuit to secondary circuit is described for H-3. Using the preliminary design parameters of demonstration HTGR in China, basic prediction of radioactive source term is done and the results are given.

Design and Analysis of Testing Platform for Study of Graphite Dust Emission Behavior During Accident Conditions

2010 ◽  
Author(s):  
Zhipeng Chen ◽  
Suyuan Yu
Keyword(s):  
Dust Emission ◽  
Fission Products ◽  
Primary Circuit ◽  
Primary System ◽  
Potential Hazard ◽  
Dust Deposition ◽  
Pebble Bed ◽  
Testing Platform ◽  
Emission Behavior ◽  
Accident Conditions

Very High temperature gas-cooled reactor (VHTR), especially the pebble-bed core type reactor, is inevitable to take place the wear of graphite components and generate the graphite dust in the core. The graphite dust was taken away by helium coolant and deposited on the surface of the primary circuit, and the fission products may be absorbed on the dust. Since it is possible that the fission products are released with dust under the accident conditions such as depressurization events, they have a potential hazard of radiation exposure to the environment. In this paper, VHTR as the research object, a testing platform is to be built with the purpose of investigating the behavior of graphite dust emission during the accident conditions. Circuit loop design is used to simulate the primary system and nitrogen is used as the working substance. Experiments of graphite dust deposition and resuspension study, as well as the study of graphite dust emission behavior during the accident conditions in pebble-bed type design VHTR will be conducted on the testing platform. The experimental data will be used for the development of VHTR source term analysis modeling.

Development of Modeling Transport of Graphite Dust in HTGR During Normal Operation Condition

2012 ◽  
Author(s):  
Zhipeng Chen ◽  
Fei Xie ◽  
Yanhua Zheng ◽  
Lei Shi ◽  
Fu Li
Keyword(s):  
Fission Products ◽  
Life Time ◽  
Normal Operation ◽  
Primary Circuit ◽  
Potential Hazard ◽  
Operation Condition ◽  
Modeling Transport ◽  
Full Power ◽  
Accident Conditions ◽  
Helium Coolant

High temperature gas-cooled reactor (HTGR), especially the pebble-bed core type reactor, will inevitably cause the wear the graphite components and generate graphite dust in the core. The graphite dust is taken away by helium coolant and deposited on the surface of the primary circuit, and the fission products may be absorbed on the dust. Since it is possible that the fission products are released with dust under the accident conditions such as depressurization events, they have a potential hazard of radiation exposure to the environment. The objective of this paper is to develop a code for calculating the behaviour of graphite dust in the primary circuit of HTGR. The paper is focused on development of models for predicting the deposition rates of the dust. The purpose of the work is to estimate the amount and distribution of deposited dust during plant life time, which was assumed to be 40 full-power years. The result will lay the foundation for further studies of fission products releasing and interaction with dust under accident conditions.

South African Licensing Framework for the Pebble Bed Modular Reactor

2008 ◽  
Author(s):  
P. Bester ◽  
T. Hill
Keyword(s):  
South Africa ◽  
South African ◽  
Safety Management ◽  
Regulatory Framework ◽  
Normal Operation ◽  
Engineering Practice ◽  
Pebble Bed ◽  
Regulatory Guidance ◽  
Design Basis ◽  
Safety Case

The electrical utility in South Africa (Eskom) plan to construct a first of a kind Pebble Bed Modular Reactor (PBMR). It has been recognized that there is a need to adapt the licensing process for the PBMR to ensure that credible and effective licensing process be developed and implemented for this technology. This paper will briefly outline the regulatory framework within South Africa, explain the licensing process adopted and present the challenges that the South African National Nuclear Regulator (NNR) was facing in developing and implementing the licensing process and how these are being addressed. The paper will discuss the update of the regulatory framework and the gaps identified in terms of regulatory requirements needed for such a project. The scope of the regulatory assessment for the licensing of the PBMR is based on the licensing requirements and criteria defined by the NNR in regulatory documents that expand on the current legislative requirements. In addition guidance is provided on selected issues in regulatory guidance documents and position papers. The requirements comprise, besides the general requirements to respect good engineering practice and the ALARA and defense-in-depth principle, specific risk criteria and radiation dose limits. These are categorized for normal operation and operational occurrences as well as for design basis events and beyond design basis events for workers and the public. Additional requirements and recommendations are stipulated by the NNR on safety important areas like quality and safety management, qualification of the nuclear fuel and the core structures, core design, verification and validation of computer codes, source term analysis and others. Selected NNR Position Papers have been developed to elaborate and provide further clarification on NNR requirements. For preparation of the PBMR safety case so-called Key Licensing Issues have been defined and agreed with the applicant. Discussions relating to these Key Licensing Issues allow important nuclear safety aspects identified for the PBMR demonstration plant to be clarified in advance of the safety case submittal.

scholarly journals Initial prediction of dust production in pebble bed reactors

2011 ◽  
Vol 2 (2) ◽  
pp. 189-195 ◽  
Author(s):  
M. Rostamian ◽  
S. Arifeen ◽  
G. P. Potirniche ◽  
A. Tokuhiro
Keyword(s):  
Computational Simulation ◽  
Fission Products ◽  
Initial Attempt ◽  
Pebble Bed ◽  
Pebble Bed Reactor ◽  
Pebble Bed Reactors ◽  
Safety Issues ◽  
Dust Generation ◽  
Very High Temperature Reactor ◽  
Very High

Abstract. This paper describes the computational simulation of contact zones between pebbles in a pebble bed reactor. In this type of reactor, the potential for graphite dust generation from frictional contact of graphite pebbles and the subsequent transport of dust and fission products can cause significant safety issues at very high temperatures around 900 °C in HTRs. The present simulation is an initial attempt to quantify the amount of nuclear grade graphite dust produced within a very high temperature reactor.

Preliminary Design Analysis of Hot Gas Ducts for a Nuclear Hydrogen System of 200 MWt

2008 ◽  
Author(s):  
Kee-Nam Song ◽  
Yong-Wan Kim
Keyword(s):  
Chemical Process ◽  
High Energy ◽  
Design Analysis ◽  
Normal Operation ◽  
Primary Circuit ◽  
Preliminary Design ◽  
Constant Flow ◽  
Strength Evaluation ◽  
Hot Gas ◽  
Hydrogen System

Very High Temperature Gas Cooled Reactor (VHTR) has been selected as a high energy heat source for a nuclear hydrogen generation. The VHTR can produce hydrogen from heat and water by using a thermo-chemical process or from heat, water, and natural gas by steam reformer technology. Korea Atomic Energy Research Institute (KAERI) is in the process of carrying out a nuclear hydrogen system by considering the indirect cycle gas cooled reactors that produce heat at temperatures in the order of 950°. The nuclear hydrogen system is planning to produce hydrogen by using nuclear energy and a thermo-chemical process. Helium gas is the choice for the coolant of the nuclear hydrogen system since it is an inert gas, with no affinity to a chemical or nuclear activity; therefore a radioactivity transport in the primary circuit of the nuclear hydrogen system is minimal under a normal operation. Moreover, its gaseous nature avoids problems related to a phase change and water-metal reactions and therefore improves its safety. A coaxial double-tube hot gas duct (HGD) is a key component connecting the reactor pressure vessel and the intermediate heat exchanger (IHX) for the nuclear hydrogen system. In this study, a preliminary design analysis for the primary and secondary HGDs of the nuclear hydrogen system was carried out. These preliminary design activities include a preliminary decision on the geometric dimensions, a preliminary strength evaluation and an appropriate material selection. A preliminary decision on the geometric dimensions of the HGDs was undertaken based on three engineering concepts, such as a constant flow velocity model (CFV model), a constant flow rate model (CFR model), a constant hydraulic head model (CHH model), and also based on a heat balanced model (HB model). We compared the geometric dimensions and their preliminary strength evaluation results from the various models.

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