1.1. BACKGROUND
Research reactors are used primarily for research, training, radioisotope production, neutron radiography and materials testing, and thus they have unique design features and operational regimes that differ from those of power reactors. Being smaller in size and generally generating much less nuclear energy than power reactors, research reactors demonstrate a broader range of designs, nuclear fuel compositions, modes of operation and safety arrange-ments. The need for greater flexibility in their use requires a comprehensive approach to safety analysis.
Safety analysis is mainly used to enable the operator to understand the basis for safe operation of the reactor and to demonstrate to the regulatory body how the design of the facility and the related operational procedures contribute to the prevention and mitigation of accidents. Safety analysis must also consider experimental devices and programmes with respect to both their safety and their effect on reactor safety.
In meeting the requirements for research reactor safety, one of the initial steps is to determine the postulated initiating events (PIEs). The PIEs define the scope of the accidents to be used in the safety analysis. They establish the scenarios to be analysed in order to predict the consequences of accidents. The analytical and computational resources needed for the safety analysis of a reactor follow from an assessment of the PIEs. For each PIE, qualitative and quantitative information needs to be given on the source term derivation and the analysis of radiological consequences.
The source term is defined as the magnitude, composition, form (physical and chemical) and mode of release (puff, intermittent or continuous) of radioactive elements (fission and/or activation products) released during a reactor accident. The mechanism, time and location of the release must also be identified. To facilitate their assessment, the radiological consequences can be grouped into the following categories:
(a) Consequences inside the reactor building with doses to operating staff or personnel within the building;
(b) On-site consequences (outside the reactor building) from:
(i) Direct radiation doses from the containment;
(ii) Inhalation and ingestion doses from active material released from the containment;
(c) Off-site consequences (to members of the public) from:
(i) Direct radiation doses from the containment;
(ii) Inhalation and ingestion doses from active gaseous or liquid releases from the containment to the environment.
This report is part of the set of publications developed within the framework of the IAEA’s activities in the area of research reactor safety.
Reference [1] provides safety requirements for the design and operation of research reactors. It covers the large variety of designs, the wide range of power levels, the different modes of operation and purposes of utilization, the partic-ularities of siting and the differences among organizations operating research reactors, in particular concerning their resources. The wide variety of these characteristics demands flexibility in the setting and fulfilment of basic require-ments when dealing with certain specific topics. IAEA Safety Guides [2–4]
provide guidance on fulfilling these requirements.
This report offers guidance on and examples of methods of analysis that may be useful for meeting the general requirements related to the derivation of the source term and the evaluation of the radiological consequences of research reactor accidents. It should be used in conjunction with the IAEA Safety Requirements and Safety Guides for research reactors [1–4].
1.2. OBJECTIVE
The objective of this report is to provide a set of suggested methods and practices, based on current good practices around the world, for deriving the source term and analysing the radiological consequences of research reactor accidents. The report covers all steps involved in performing analyses, that is, the selection of initiating events and the analysis of the core damage mechanisms and progression, radionuclide inventory releases, and radiological consequences inside the reactor building, outside the reactor building and off the site. It also presents practical examples applied to existing research reactors.
The aim is to provide safety analysts and reviewers with methods and techniques for, and practical examples related to, the derivation of the source term and the evaluation of the radiological consequences of postulated research reactor accidents having the potential to lead to radioactive releases.
These practical examples will also assist in meeting the requirements and recommendations related to carrying out safety analyses and preparing the safety analysis report (SAR). In particular, the present publication provides guidance on the requirements established in paras 6.72–6.78 of Ref. [1] and
on the recommendations given in section A.16 (in particular, paras A.1626–
A.1645) of Ref. [2], including detailed discussions and examples of related topics.
1.3. SCOPE
The report will be particularly useful to safety analysts and reviewers in fulfilling the requirements and recommendations related to carrying out safety analyses and preparing SARs. In addition, it will help regulators to conduct safety reviews and assessments of the topics covered. It will be also useful to research reactor designers and operators.
The report may be applied, to varying degrees, to all research reactors.
Sophisticated calculations and methods such as those described herein may be used as deemed necessary for the preparation of safety analyses of newly designed or reconstructed research reactors. These methods may also be used for updating or reassessing previous safety analyses of operating research reactors when one or more of the following circumstances occur:
— The results of the application of conservative assumptions and simplified methods do not meet the applicable acceptance criteria.
— The review of the safety analysis of an operating research reactor reveals deficiencies in previous analyses.
— New accident scenarios are envisaged or postulated during the operation of the research reactor.
1.4. STRUCTURE
Sections 2–4 of this publication provide information and background on various aspects of source term derivation, and considerations to be taken into account when evaluating radiological consequences. Section 2 deals with the general aspects of the analytical approach to source term derivation (determin-istic or probabil(determin-istic); application of the graded approach to analysis of research reactors according to reactor type, size or power level, utilization and fuel characteristics; identification of PIEs and scenarios that may lead to core damage; and selection of the scenarios. Section 3 discusses the derivation of the source term. This includes: determination of inventories of fission products, transuranic elements and activation products; analysis of the core damage mechanisms and progression; and evaluation of radionuclide releases from different kinds of fuel and from experimental devices, taking into account
retention in systems and components. It also discusses releases from the reactor containment. Section 4 addresses the analysis of radiological consequences, grouped into three categories: those on the site inside the reactor building, those on the site outside the reactor building and those off the site. External and internal doses are analysed, taking into account the various pathways for gaseous and liquid releases. Section 5 provides an overview of an integrated approach to the derivation of the source term and related radiological conse-quences. Appendices I–VII complement the technical information provided in Sections 2–4, and Annexes I–III present practical examples from specific research reactors.