IAEA International Atomic Energy Agency Presenter Name School of Drafting Regulations for Borehole Disposal of DSRS 2016 Vienna, Austria Overview of radiation protection considerations during the operational and post-closure period

IAEA 2 Content Radiation protection requirements in the operational phase. General overview of the operational safety assessment and its key components as applied to the borehole disposal of disused sealed radioactive sources. Overview of the work that has been undertaken to date on the development of operational safety assessments for the borehole disposal of disused sealed radioactive sources. Radiation protection requirements in the the post-closure phase. General overview of the post-closure safety assessment and its key components as applied to the borehole disposal of disused sealed radioactive sources. Overview of the work that has been undertaken to date on the development of post-closure safety assessments for the borehole disposal of disused sealed radioactive sources

IAEA What is a safety assessment? Iterative procedure for evaluating performance of a system and its potential impact on human health and environment Aim is to provide reasonable assurance that system will provide a sufficient level of safety and meet relevant requirements for protection of human health and environment

IAEA Safety assessment (cont.) Includes quantification of: disposal system performance analysis of associated uncertainties comparison with relevant design and safety standards Can be used to focus site investigations on key issues Refined and more site-specific as site investigations and design studies progress

IAEA Safety assessment (cont.)

IAEA Safety assessment (cont.) IAEA’s General Safety Requirements Part 4 (GSR Part 4) on the Safety Assessment for Facilities and Activities place 24 requirements on a safety assessment Includes concept of “graded approach” in which the level of detail in the safety assessment is consistent with the magnitude of the possible radiation risks arising from the facility

IAEA Safety assessment (cont.) IAEA’s Specific Safety Requirements (SSR-5) on the Disposal of Radioactive Waste place the same requirements on the safety assessment as the safety case (i.e. Requirements 3, 12, 13 and 14) Safety assessment forms part of the safety case

IAEA Radiation protection requirements in the operational phase I The radiation safety requirements and the related safety criteria for the operational period of a disposal facility are the same as those for any nuclear facility or activity involving radioactive material and are established in the International Basic Safety Standards. Disposal facilities for small quantities of waste (e.g. borehole facilities) may not be regarded as nuclear facilities in some States but have to be subject to an appropriate regulatory process and have to be licensed accordingly. In radiation safety terms, the disposal facility is considered to be a source of radiation that is under regulatory control in a planned exposure situation. 8

IAEA Radiation protection requirements in the operational phase II The primary goal is to ensure that radiation doses are as low as reasonably achievable and within the applicable system of dose limitation. The optimization of protection is considered in the design of the disposal facility and in the planning of all operations. For a disposal facility, as for any other operational nuclear facility or facility where radioactive material is handled, used, stored or processed, an operational radiation protection programme, commensurate with the radiological hazards, is required to be put in place to ensure that doses to workers during normal operations are controlled and that the requirements for the limitation of radiation doses are met. 9

IAEA Radiation protection requirements in the operational phase III Relevant considerations in the optimization of measures for protection and safety include: the separation of mining and construction activities from waste emplacement activities; the use of remote handling equipment and shielded equipment for waste emplacement, where necessary; the control of the working environment so as to reduce the potential for accidents and their potential consequences; and the minimization of the need for maintenance in supervised areas and controlled areas. Contamination is required to be controlled and prevented to the extent possible. 10

IAEA Radiation protection requirements in the operational phase IV Emergency plans are required to be put in place for dealing with accidents and other incidents, and for ensuring that any consequent radiation doses are controlled to the extent possible, with due regard for the relevant emergency action levels. The doses and risks associated with the transport of radioactive waste through public areas to a disposal facility are required to be managed in the same way as the doses and risks associated with the transport of other radioactive material. 11

IAEA Operational safety assessment Assess exposures to workers and public during: transport of conditioned sources to borehole disposal of sources in borehole and the subsequent backfilling and closure Generally need to consider: any expected exposures non-routine/accidental exposures/releases

IAEA Operational safety assessment A generic (non-site specific) operational safety assessment has been undertaken by the Nuclear Energy Corporation of South Africa (NECSA) No site-specific operational safety assessment has yet been fully developed

IAEA NECSA Operational SA NECSA’s generic work comprised: a Hazard and Operability (HAZOP) study a hazard assessment to calculate doses for routine and non-routine conditions Considered four categories of activity: Placement of sources in capsules Placement of capsule in a cement lined disposal container Transport of disposal container to borehole Emplacement of disposal container into borehole

IAEA NECSA HAZOP Study For each activity, HAZOP study considered: Potential deviations from planned operation Causes of each deviation Consequences of each deviation Impact of each deviation Safeguards against each deviation Actions to prevent/mitigate each deviation

IAEA NECSA HAZOP: transport to borehole DeviationCausesConsequences Impact Area and Ranking SafeguardsActions RPO action omitted. RPO fails to measure the dose rate on the disposal container or to perform any of his functions. Personnel may be exposed to higher doses than anticipated. Personnel. Low. Personal dosimeters, with cumulative and dose rate alarms will be worn by operators. Suitably qualified RPO must be used. Operator action omitted. The operator fails to perform a prescribed action. The personnel may be exposed to higher doses than anticipated. Personnel. Low. Qualified RPO will supervise the operation. Work instructions must be written. Train operators in these work instructions. More of pressure Regulator on gas bottle fails. The hose to the release mechanism of the hoist will burst or come loose. Operations. Polymer should burst before any damage to equipment. None The transfer may be interrupted. The transport vehicle may fail for any number of reasons. Personnel may be exposed for longer than anticipated. Personnel. Low. Personal dosimeters, with cumulative and dose rate alarms will be worn by operators. RPO will handle the situation. None

IAEA NECSA HAZOP: disposal in borehole DeviationCausesConsequences Impact Area and Ranking SafeguardsActions Operator action omitted. The operator fails to perform a prescribed action. The activity may be released earlier than anticipated into the disposal container or the personnel may be exposed to higher doses than anticipated. Public. Low. Personnel. Low. A leak test will be performed on every container. The disposal container will be sealed and leak tight. Qualified RPO will supervise the operation. The post closure Safety Assessment will indicate if this is a significant event. Work instructions must be written. Train operators in these work instructions. No compressed air available to release the container at the bottom of the borehole. Air supply failure. Operations are delayed. Container cannot be released at the bottom of the borehole. Operations.None Consider taking a spare hose and cylinder on the mission. Unable to lower the disposal container. Power failure or failure of hoisting equipment. Operations delayed. The disposal container may drop down the borehole, resulting in loss of containment. Operations. Public. Low. None Make sure the hoisting equipment does not allow free fall when the power fails. Too much rain.Act of nature Water in the borehole may make operations such as capping difficult. OperationsNone Consider not disposing during rain, and keeping a hood on the borehole until just before disposal. The container may be damaged when it falls down the borehole. The cable failure or failure of cable link. The disposal container may drop down the borehole. Activity may be released sooner than anticipated. Equipment. Public. Low. The air supply hose may be strong enough to carry the weight of the container. Procedure to check hoisting equipment before every mission. Scenario to be addressed in the Safety Assessment.

IAEA NECSA Hazard Assessment For each activity, considered doses to workers and public for normal operations and accident situations Doses to workers can be controlled to acceptable levels through an appropriate radiological protection programme Even without appropriate controls, a worker will receive a dose of <1 mSv – well below the dose limit of 50 mSv in any single year Doses to the public are considered to be insignificant

IAEA Routine operational doses for transport and disposal of 40 GBq Co-60 source NECSA Hazard Assessment Activity Duration (min) Distance from source (m) Factor for shielding Dose (µSv) Transport Transporting the shielded transfer container to the borehole on a transport vehicle Manoeuvring the transport vehicle over the borehole Subtotal9 Disposal Opening the bottom of the shielded transfer container Lowering the disposal container from the shielded transfer container down the borehole Capping the disposal container with backfill material Subtotal63

IAEA NECSA Operational SA: SHARS SHARS (spent high activity radioactive sources) are sources of greater than 1 Ci Co-60 equivalent NECSA have designed a mobile hot cell (MHC) for remote handling and conditioning of SHARS of up to an equivalent of a 1000 Ci Co-60 source MHC could be used for direct disposal to borehole

IAEA NECSA Operational SA: SHARS (cont) NECSA HAZOP study and hazard assessment for conditioning using MHC Not yet considered transport to or disposal in borehole of conditioned SHARS Could extend operational safety assessment for conditioning to include transport and disposal

IAEA Radiation protection requirements in the post operational phase I The primary goal of the disposal of radioactive waste is the protection of people and the environment in the long term, after the disposal facility has been closed. The dose limit for members of the public for doses from all planned exposure situations is an effective dose of 1 mSv in a year. This and its risk equivalent are considered criteria that are not to be exceeded in the future. To comply with this dose limit, a disposal facility (considered as a single source) is so designed that the calculated dose or risk to the representative person who might be exposed in the future as a result of possible natural processes affecting the disposal facility does not exceed a dose constraint of 0.3 mSv in a year or a risk constraint of the order of 10–5 per year. 22

IAEA Radiation protection requirements in the post operational phase II In relation to the effects of inadvertent human intrusion after closure, if such intrusion is expected to lead to an annual dose of less than 1 mSv to those living around the site, then efforts to reduce the probability of intrusion or to limit its consequences are not warranted. If human intrusion were expected to lead to a possible annual dose of more than 20 mSv to those living around the site, then alternative options for waste disposal are to be considered, for example, disposal of the waste below the surface, or separation of the radionuclide content giving rise to the higher dose. 23

IAEA Radiation protection requirements in the post operational phase III If annual doses in the range 1–20 mSv are indicated, then reasonable efforts are warranted at the stage of development of the facility to reduce the probability of intrusion or to limit its consequences by means of optimization of the facility’s design. It is recognized that radiation doses to people in the future can only be estimated and that uncertainties associated with these estimates will increase for periods farther into the future. Optimization under constraints is the central approach adopted to ensure the safety of a waste disposal facility. In this context, the optimization of protection is a judgemental process, social and economic factors being taken into account. 24

IAEA Radiation protection requirements in the post operational phase IV The optimization is conducted in a structured but essentially qualitative manner, supported by quantitative analysis. The impact of non-radioactive material present in a disposal facility has to be assessed in accordance with national or other specific regulations and this may be significant in some cases, for example, for some mining wastes and mixtures of radioactive and toxic wastes. If non-radioactive material may affect the release and migration of radioactive contaminants from the radioactive waste, then such interactions have to be considered in the safety assessment. 25

IAEA Post-closure safety assessment Assess exposures to public following closure of the disposal facility Might need to consider timescales of thousands of years Assessment results are estimates rather than predictions of future impacts Need to consider both the expected and unexpected, low probability, events affecting the disposal system

IAEA Post-closure SA for the BDC Generic post-closure safety assessment was produced for NECSA in 2003 and 2004 This was further developed for the IAEA between 2004 and 2008 and the resulting report (the “GSA”) is currently in the final phase of editing ready for publication. A lecture on the generic safety assessment will be presented in this training event

IAEA GSA Regulatory Framework Generic safety assessment – not specific to any particular country or legislative framework Framework: based on recommendations in IAEA Specific Safety Guide on Borehole Disposal Facilities for Radioactive Waste (SSG-1) consistent with other IAEA and ICRP recommendations Dose constraint of 0.3 mSv y -1 for natural processes

IAEA Environmental and non-radiological concerns In the past it has been assumed that, subject to appropriate definition of exposed groups, the protection of humans against the radiological hazards associated with a borehole disposal facility would also satisfy the need to protect the environment. The need to consider the protection of the environment against ionizing radiation, and possible protection standards, are currently under discussion internationally 29

IAEA Environmental and non-radiological concerns The impact of non-radioactive materials present in a borehole disposal facility should also be assessed. Factors that should be considered may include: the content of chemically or biologically toxic materials in the waste or the engineered barrier materials, the protection of groundwater resources and the ecological sensitivity of the environment into which contaminants may be released. For example, if disused sealed sources were to be disposed of together with their lead shielding, safety assessments would need to examine the potential migration of the lead. 30

IAEA GSA Findings For most radionuclides, including longer-lived radionuclides such as Ra-226, the BDC provides adequate post-closure safety for the systems evaluated. Even for radionuclides with half-lives in excess of 100,000 years (e.g. Pu-238, Pu-239 and Np-237 (ingrown from Am- 241)), ~1TBq of the parent radionuclide can be disposed in a single borehole. 31

IAEA Use of the GSA Starting point for the post-closure safety assessment of a specific site (e.g. the Ghana-specific assessment). Worked example that can be used to guide/inform a site- specific assessment. Identify key waste and site attributes that need to be characterised as part of a waste and site characterisation programme. 32

IAEA GSA “Health Warnings” Need to be aware of the various assumptions adopted in the GSA, especially those affect the range of conditions to which the results of the GSA can be applied, e.g.: Narrow diameter sources (<15 mm) Depth of cover (> 30 m) Activities up to Bq Absence of natural resources liable to be abstracted Limited geomorphological activity 33

IAEA Summary Operational safety assessments consider the doses to workers and public for normal operations and accident situations A generic operational safety assessment has been undertaken by NECSA for the BDC Doses can be controlled by a radiological protection programme Even without appropriate controls, doses are acceptable

IAEA Summary Post-closure safety assessments consider the doses to the public following borehole closure Generic post-closure safety assessments have been undertaken by NECSA and the IAEA for the BDC For most radionuclides, the BDC provides adequate post-closure safety for a wide range of systems Even for long-lived radionuclides, ~1TBq can be safely disposed in a single borehole

IAEA 36 Thank you!