Chapter 7: Radiological Protection under ICRP 103 and 122

Radiological protection in radioactive waste management covers three concerns together: occupational exposure during operations, exposure of members of the public around a facility, and long-term impacts after a repository has been closed. ICRP Publication 103 provides the basic framework of protection principles and exposure situations, and Publication 122 addresses how that framework applies to the geological disposal of long-lived solid radioactive waste.

WIA-ENE-026 sets out a structure for recording and exchanging information on waste packages, storage facilities, radiation measurements and their history. This chapter links the scientific concepts needed for protection decisions to that data and API. It distinguishes, in turn, the items and values defined by WIA-ENE-026 v1.0, the values that should be updated to current international recommendations before use, and the matters to be settled by facility-specific assessment and implementation. Conformance with the data format demonstrates only that records are consistent; the safety and legal compliance of a facility are determined through separate assessment and regulatory processes.

7.1 Protection Principles and Exposure Situations

7.1.1 Justification, Optimization and Dose Limits

WIA-ENE-026 v1.0 §5.2 presents justification, optimization and dose limits as the principles of protection. Each principle answers a different question: whether an activity is warranted, how protection options compare, and how individual doses are controlled. Each of these is decided through its own process.

Justification

Any decision that introduces exposure or alters an exposure situation is justified by weighing the benefit it brings against the harm it causes. When waste treatment methods are compared, account is taken not only of exposures during treatment but also of the consequences that differ in subsequent storage, transport and disposal.

Optimization

Exposures are kept as low as reasonably achievable, taking economic and societal factors into account. This is known as ALARA (As Low As Reasonably Achievable). Opportunities for improvement are examined even below the limits, but optimization does not mean unconditional minimization that disregards cost and other safety consequences.

Application of Dose Limits

In planned exposure situations, the dose to each individual to whom the limits apply is controlled so that it does not exceed the prescribed limits. Compliance with a limit does not end the optimization review, and a low average dose across a workforce does not by itself demonstrate that every individual complies with the limits.

7.1.2 Planned, Emergency and Existing Exposure Situations

A planned exposure situation is one in which the introduction or operation of a source is planned in advance and protection can be designed beforehand. The operation of waste treatment, storage and disposal facilities is managed within this framework. An emergency exposure situation is one requiring urgent protective action, for example after an accident, whereas an existing exposure situation is one that already exists when a decision on control has to be taken. Long-term management of residual contamination after an accident, among other cases, may be addressed within the existing exposure framework, depending on the circumstances.

Potential exposure is exposure that is not certain to occur but may result from particular events or conditions. ICRP Publication 103 addresses potential exposure mainly in the context of planned exposure situations, although the concept can also be applied to other exposure situations. For planned facilities, potential exposure is assessed with the likelihood of occurrence considered separately from the consequences should it occur.

Because geological disposal is a planned activity, the long-term post-closure safety assessment that addresses exposures of future generations also applies the protection system for planned exposure and the assessment of potential exposure. Expected evolution of the facility, events whose occurrence is uncertain, and contamination that has actually occurred are each judged according to their own conditions.

7.2 The Meaning of Dose and the Application of Limits

7.2.1 Activity, Dose and Dose Rate

Activity is the number of decays per unit time, expressed in becquerels (Bq). Absorbed dose is the energy imparted per unit mass of matter, expressed in grays (Gy). Equivalent dose is the absorbed dose in a tissue weighted for the type of radiation, and effective dose is the sum of the tissue equivalent doses each weighted by a tissue weighting factor. Equivalent dose and effective dose are both expressed in sieverts (Sv), but they apply to different targets and have different meanings.

Converting the activity of waste into a dose to an individual requires information on the radionuclides, radiation energies, the positions of the source and the person, shielding, occupancy time and whether any intake occurs. The dose rate at the surface of a waste package describes the radiation field at that point; it is a quantity distinct from the annual effective dose to a worker or the annual dose to a member of the public. Dose-related values obtained from instruments are likewise interpreted after confirming the definition of the measured quantity and the calibration conditions.

7.2.2 Dose Criteria in the Standard and the Lens of the Eye Limit

The 20 mSv/year for worker effective dose is a five-year averaging condition. A value above 20 mSv in a single year therefore cannot, on its own, be judged a breach of this condition. Conversely, a value of 50 mSv or less in a given year does not, on its own, show that the five-year averaging condition has also been met. The records for the entire assessment period and the single-year maximum are checked together.

The equivalent dose limit for the lens of the eye of workers is the value that ICRP reduced in its April 2011 statement on tissue reactions. The current criterion is an average of 20 mSv per year over defined periods of 5 years (100 mSv in total over 5 years), with no more than 50 mSv in any single year. The statement was published in ICRP Publication 118 (2012), and IAEA GSR Part 3 (2014) adopts the same value as the occupational limit for the lens of the eye while noting that the previous limit was 150 mSv per year. Since the 150 mSv in the table of WIA-ENE-026 v1.0 §5.1.1 reflects this former value, implementations should preferably apply the current conditions of 20 mSv/year averaged over 5 years and 50 mSv in a single year.

As with effective dose, the lens of the eye limit is judged on both the averaging condition and the single-year condition. For an actual compliance determination, the national legislation and licence conditions applicable to the facility are confirmed, and the criteria and editions used in the determination rules are recorded.

7.2.3 Limits, Constraints, Reference Levels and Alarm Values

A dose limit is a criterion that restricts the dose to an individual to whom it applies. A dose constraint is a restriction used in the prospective optimization of protection for a particular source or activity. Because doses received from several sources have to be controlled, the constraint for a particular facility is not treated as equivalent to the overall dose limit applying to an individual.

A reference level is used as the benchmark for optimizing protection strategies in emergency or existing exposure situations. Reasonable opportunities for further reduction are examined below the reference level as well. Not every comparison value used in a geological disposal assessment should be labelled a reference level, nor should values from different situations be applied interchangeably.

A sensor alarm value is a set point for initiating a field response or a review of measurements. An alarm on instantaneous dose rate and an annual cumulative dose limit differ in units and assessment period. The alarm values in the standard's monitoring examples are used as triggers for initiating a field response.

7.3 Worker Protection and ALARA Optimization

7.3.1 Improvements in Design, Operation and Management

In design, shielding, remote handling, equipment layout and access for maintenance are reviewed together. Even for the same task, exposure can be reduced by shortening the time spent close to the source and eliminating unnecessary rework. The effect of additional shielding on equipment access and handling safety is also included in the comparison of options.

In operation, pre-job confirmation of the radiation field, preparation of the work sequence and movement routes, checks of individual dose, and procedures for stopping and reassessing when conditions differ from those expected are linked together. In management, differences between expected and actual doses are analyzed to identify the tasks on which exposure is concentrated and the effect of improvements. The division into design, operation and management is an explanatory perspective intended to ensure that no opportunity for improvement is overlooked.

7.3.2 Conditions for Judging Shielding Performance

Shielding design takes into account, in addition to material thickness, the emission characteristics of each radionuclide, source strength and distribution, energy, the geometry of the source and the assessment point, the density and composition of the material, openings and penetrations, and scattered and secondary radiation. Where gamma rays and neutrons are both present, the assessment must reflect the behaviour and interactions of each type of radiation.

A simple exponential attenuation formula is useful for explaining the basic principle of shielding under fixed conditions. It cannot, however, adequately represent extended sources, scattered radiation, composite materials or leakage through gaps in a real facility. The adequacy of an actual composite shielding structure is therefore judged by a detailed assessment that accounts for scattered radiation and leakage.

The concrete thickness of 1.5–2 m in WIA-ENE-026 v1.0 §6.2 appears in the description of dry storage concrete modules and is therefore to be interpreted only within the scope of that description. The adequacy of shielding is demonstrated by a facility-specific shielding assessment, irrespective of whether the thickness falls within this range.

7.3.3 Monitoring of Individual Dose and Contamination

Thermoluminescent dosimeters (TLDs) and optically stimulated luminescence (OSL) dosimeters are used to assess cumulative exposure over the wearing period. Electronic personal dosimeters (EPDs) can be used to check dose and dose rate during work and to provide alarms. The role of each device depends on the type of radiation, energy response, measurement range and approved purpose of use.

The combination of dosimeters, wearing positions and readout intervals are determined according to working conditions and the criteria applicable to the facility. For tasks in which exposure of the lens of the eye or the extremities is significant, it is also examined whether whole-body dosimeter results alone are sufficient to assess the dose to those parts of the body.

In waste handling, the possibility of intake of radioactive material is considered along with external exposure. Surface contamination, airborne radioactive material and individual dose each provide different information. Workplace measurements and individual dose records are linked to one another but managed separately, each as its own measured quantity. Calibration status, wearing periods, readout anomalies, missing data and reassessment history are retained as the basis for interpreting the results.

7.4 Public Protection and Dose Assessment for the Representative Person

Public dose assessment links the movement of radioactive material from a facility into the environment with the pathways by which people are exposed. Releases, concentrations in environmental media such as air, water and soil, external exposure and inhalation and ingestion pathways, and habits and occupancy times are examined. The annual effective dose to members of the public is determined by combining this pathway information with concentrations in environmental samples or dose rates at the site boundary.

The representative person is a hypothetical individual for assessment purposes, defined to represent the doses to the more highly exposed people. The representative person differs from a simple average of the entire local population, and is not a hypothetical individual who combines extreme values of every parameter without qualification. ICRP Publication 101a (2006) addresses methods for assessing the dose to the representative person for the purpose of protecting the public.

The assessment states the population group and age, spatial extent, assessment period, principal exposure pathways and the basis of the habit data. The contribution of the facility is distinguished from background, and measured values, model estimates and conservative assumptions are presented separately. A result below the detection limit does not mean that radioactive material is absent, so the detection limit and the data treatment method are explained together.

When results are communicated to the public, it is stated what was measured, where and for how long, and what can be inferred from the results. Even when comparisons are made with natural background radiation or medical exposure, values that differ in population, period and purpose of exposure are not treated as having the same meaning. Medical use, which varies between individuals, is not generalized into a single average additional dose without information on the population concerned.

Public dose assessment and assessment of impacts on the environment itself differ in purpose and subject. A low dose to residents is not taken, on its own, as accounting for all environmental impacts, and the scope of any necessary environmental assessment is determined separately.

7.5 Post-Closure Safety Assessment of a Repository

7.5.1 The Boundary Between Operational Protection and Post-Closure Assessment

During operation, exposures arising from the receipt, emplacement and handling of waste and from facility maintenance are controlled. Exposure during closure work itself is also a matter of worker protection. Post-closure assessment addresses how the closed facility and its surrounding environment evolve over time and how that evolution affects the isolation and migration of radioactive material. Exposure during the operational period is addressed in the assessment of operational protection.

The assessment period and its time frames are determined on the basis of radionuclide characteristics, barrier functions, the geological and hydrogeological conditions of the site, the range of applicability of the models and regulatory requirements. ICRP Publication 122 (2013) describes time frames in terms of the form of oversight: a period of direct oversight, when the disposal facility is being constructed and operated and is actively overseen; a period of indirect oversight, when the facility has been sealed and is overseen by regulators or society for additional assurance; and a period of no oversight, when memory of the facility has been lost. These time frames are defined by the way oversight is exercised, not by a fixed number of years. The same publication gives a dose constraint of 0.3 mSv per year for members of the public (the same value as the SSR-5 criterion in Section 3.2.1).

7.5.2 Scenarios and Uncertainty

The assessment distinguishes the expected evolution of the facility and environment from events whose occurrence is uncertain. For each scenario, the assumed events, the barriers affected, the radionuclide migration pathways, the assessment endpoint and the scope within which the results can be interpreted are stated. Potential exposure is examined with the likelihood of an event considered separately from its consequences should it occur.

For issues such as inadvertent human intrusion, where it is difficult to predict human behaviour in the distant future in any specific way, calculations based on simplified assumptions may be used. Such calculations are tools for examining the protective characteristics of the facility under the assumed conditions; they are not predictions that fix when an actual future event will occur or what dose it will cause.

Long-term assessment examines the processes that govern the results, such as radionuclide inventory and decay, changes in barrier function, groundwater movement, and radionuclide retardation and dispersion. Where the uncertainty of data and models grows, it becomes important to consider evidence on safety functions, sensitivity analyses and the results of alternative assumptions together, rather than relying on numerical precision.

7.5.3 Elements of the Safety Case

The safety case brings together not only calculation results but also site investigation, the design basis, quality management, and arguments concerning safety functions and uncertainty. IAEA Specific Safety Guide SSG-23, The Safety Case and Safety Assessment for the Disposal of Radioactive Waste (2012), addresses the safety case and safety assessment for radioactive waste disposal facilities. The safety case and assessment for treatment and storage facilities in the stage before disposal (predisposal management) are addressed by General Safety Guide GSG-3, The Safety Case and Safety Assessment for the Predisposal Management of Radioactive Waste (2013), so the two documents are cited separately.

WIA radionuclide information and the geological and barrier information for storage facilities can be used to trace assessment inputs. Long-term safety is demonstrated through the arguments of the safety case, independently of field inputs. The reference dates of the data used, model versions, the rationale for scenario selection and the results of reviews are linked and retained.

7.6 Emergency Exposure and Response Planning

IAEA GSR Part 7, Preparedness and Response for a Nuclear or Radiological Emergency (2015), sets out the safety requirements for preparedness and response for a nuclear or radiological emergency. Response plans for waste facilities should be based on an assessment of the types and scale of emergencies that could occur at the facility and should be linked to the actual command and notification arrangements. The criteria below are used to describe the stages of decision-making in an emergency response.

Emergency Action Level (EAL)

A decision criterion for classifying an emergency and initiating the response, based on facility conditions, observable conditions and similar indicators.

Operational Intervention Level (OIL)

A criterion set so that the need for protective actions can be judged using measurable quantities such as dose rates or concentrations of radioactive material.

Reference levels during recovery

Criteria used when reviewing protection strategies to reduce residual exposure. They are distinguished in purpose and conditions of application from the decision criteria used in the early phase of an emergency.

When a sensor alarm is raised, the reliability of the measurement and the conditions in the field are checked, and the established procedures are followed through to notification and decisions on protective actions. Recording the times of the alarm, acknowledgement of receipt, decision, implementation of actions and follow-up measurements separately makes it possible to review delays and omissions in the response. A sensor alarm is the starting point of the process for determining whether a dose limit has been exceeded and whether protective actions are needed.

The fields safety.emergencyContacts, safety.emergencyProcedure, safety.lastDrill and safety.nextDrill in the WIA-ENE-026 v1.0 waste package structure (§4.1) link contacts, procedures and drill schedules. Whether the contact network actually functions and how effective drills are is confirmed through drill results. Drills examine omissions in notification, the transfer of measurement data, hand-overs between roles and the implementation of follow-up improvements. Mandatory matters such as statutory drill intervals and evacuation criteria are confirmed in the legislation and emergency plans applicable to the facility.

7.7 Scope of Use of WIA Data and the API

7.7.1 Data Linked to Protection Decisions

WIA-ENE-026 v1.0 defines information structures for radioactive waste packages (§4.1) and storage facilities (§4.2), but defines no dedicated fields for protection principles, exposure situations or post-closure time frames. Where these concepts are recorded in internal systems, they are marked as implementation-specific extensions and kept distinct from the standard fields.

As a non-normative implementation design, an assessment record can capture the applicable criteria, exposure situation, calculation assumptions, data sources, reviewers and approval history, and be linked to the identifiers of the waste packages concerned. This is a design example intended to improve the reproducibility and explainability of assessments; the request and response formats of the standard are used unchanged.

7.7.2 APIs Defined by the Standard

The full list of APIs is covered in Section 2.8; the following table sets out those endpoints used for processing protection data and the points to check when using them.

The SDK examples in §11 of the standard give distance and shielding thickness in metres and time in seconds. The REST request example for dose calculation in §10 of the standard shows the same items as bare numbers, so an actual implementation should document the units and calculation conditions used in requests and confirm that they match the units in §11. The dose quantity and unit that the response value represents should also be recorded.

The following is the example request body for POST /dose/decay given in §10.1.3 of the standard. The input values are illustrative and are neither measurements of actual waste nor assessment results.

{
  "isotope": "Cs-137",
  "initialActivity": 1.0e14,
  "decayTime": 30,
  "unit": "years"
}

isotope is the radionuclide to be calculated, initialActivity the initial activity, and decayTime and unit the elapsed time and its unit. A decay calculation deals with the change in activity over time; it does not calculate migration into the surrounding environment or doses to people. Where the ingrowth of progeny in a decay chain is significant, it is also checked whether the calculation model includes it.

Dose assessment for the representative person and post-closure safety assessment are carried out with separate assessment models outside the scope of the §10 API. The model inputs, calculation scope, basis of validation and version are recorded in the assessment record and linked to WIA data through the relevant packageId and facility identifiers.

7.7.3 Security and External Information Exchange

Security requirements such as transport security, encryption and retention of audit logs are set out in Section 4.8.1. In systems that are combined with individual dose records, role-based access rights, the extent to which identifying information is disclosed and the history of record corrections are designed separately. The retention period for audit logs and the obligation to retain individual dose records are each set on their own basis, and requirements under personal data protection law are confirmed against the processing as a whole, independently of whether encryption is applied.

Exchanging data with external systems requires separate agreement on the meaning of items, units, identifiers, permissions and transfer procedures. NEWMDB (Net Enabled Waste Management Database) was a database operated by the IAEA to collect information on radioactive waste inventories and management status; it has since been replaced by the Spent Fuel and Radioactive Waste Information System (SRIS). RANET (Response and Assistance Network) is the IAEA system for connecting international assistance in nuclear or radiological emergencies. Submitting data to, or requesting assistance through, such systems follows the procedures of each system and of the national authorities concerned.

7.8 Aggregation and Interpretation of Dose Data

7.8.1 Individual Dose and Collective Dose

Individual dose is the value for one person over a defined period. Collective effective dose is the sum of individual effective doses for a defined group over a defined period, expressed in person-Sv. The average per worker, the annual total for a facility and the total for a single task differ in the scope of aggregation and cannot be used interchangeably.

Example of a collective dose calculation: Suppose that 10 workers each receive 1 mSv during a single waste packaging task. The collective effective dose for that task is then 10 person-mSv, or 0.010 person-Sv. This value is limited to that task. The annual collective dose is obtained by summing the doses received by the same group from all of its work during that year.

In the same example, if 20 workers each receive 0.5 mSv, individual doses are lower but the collective effective dose remains 0.010 person-Sv. Worker rotation can contribute to the control of individual doses, while any change in collective dose is calculated and confirmed separately. The effectiveness of ALARA is judged by looking at both the distribution of individual doses and the dose for the task as a whole.

7.8.2 Averaging Periods and Completeness of Data

Example of applying the effective dose limit: Suppose that an individual's effective doses over the five years defined for assessment are, in order, 12, 18, 24, 16 and 20 mSv. The total is 90 mSv, the annual average is 18 mSv and the single-year maximum is 24 mSv. On the assumption that the records are complete, this example meets both the five-year averaging condition and the single-year maximum condition of the standard.

This example illustrates only the two conditions for effective dose. It is not a determination of doses to the lens of the eye, skin or extremities, nor of the compliance of the facility as a whole. If necessary data are missing, such as occupational exposure received at other workplaces or internal exposure assessments, compliance with the limits must not be confirmed by treating the missing values as zero.

When trends are compared between periods or facilities, it is confirmed that the workers included, types of work, workload, measurement methods and treatment of missing data are the same. Multiplying an annual average individual dose by the number of people taking part in a particular inspection task requires evidence that the average represents the doses for that inspection period and the participating group. A low or declining national average does not on its own establish the effectiveness of a particular project or the relative merit of ALARA implementation.

7.9 Reference Documents and Implementation Review

7.9.1 Role of the Reference Documents

The following documents are reference material for understanding protection concepts and the scope of assessment. The fields and API of WIA-ENE-026 v1.0 are defined by the text of the standard; the documents below provide background for interpreting and assessing that data. When applying values, the edition, scope and conditions of the relevant document are confirmed.

• ICRP Publication 103 (2007), The 2007 Recommendations of the International Commission on Radiological Protection: the basic recommendations describing the protection principles and exposure situations.
• ICRP Publication 122 (2013), Radiological Protection in Geological Disposal of Long-lived Solid Radioactive Waste: addresses how the 2007 Recommendations apply to the geological disposal of long-lived solid radioactive waste and the distinction between periods of oversight.
• ICRP Publication 101a (2006), Assessing Dose of the Representative Person for the Purpose of Radiation Protection of the Public: addresses dose assessment for the representative person for the protection of the public.
• ICRP Publication 118 (2012): addresses early and late effects of radiation in normal tissues and organs and threshold doses for tissue reactions in a radiation protection context, and contains the 2011 statement that lowered the dose limit for the lens of the eye.
• IAEA GSR Part 3 (2014), Radiation Protection and Safety of Radiation Sources: International Basic Safety Standards: the basic safety requirements for radiation protection and the safety of radiation sources, which set the occupational limit for the lens of the eye at 20 mSv/year averaged over 5 years (50 mSv in a single year).
• IAEA GSR Part 7 (2015), Preparedness and Response for a Nuclear or Radiological Emergency: addresses preparedness and response for a nuclear or radiological emergency.
• IAEA SSG-23 (2012), The Safety Case and Safety Assessment for the Disposal of Radioactive Waste: addresses the safety case and safety assessment for the disposal of radioactive waste.
• IAEA GSG-3 (2013), The Safety Case and Safety Assessment for the Predisposal Management of Radioactive Waste: addresses the safety case and assessment for treatment and storage facilities in the stage before disposal.

7.9.2 Checklist for Protection Assessment

1. Divide the subject of assessment into occupational exposure, public exposure and post-closure impacts, and define the exposed persons, period, dose quantity and principal exposure pathways for each assessment.
2. Record the documents and editions of the criteria applied. Check the averaging period and the single-year condition for worker effective dose together. For the lens of the eye limit, follow the national legislation applicable to the facility, but review against the current international recommendation (20 mSv/year averaged over 5 years, 50 mSv in a single year) rather than the 150 mSv in the standard's table. Where past records were judged against the 150 mSv in the standard's table, mark the criterion used so that it can be distinguished.
3. Confirm the relevant packageId and facility information, and link the reference date of the radionuclide inventory, the measurement position of dose rates, and units and uncertainty to the assessment inputs.
4. State the models and assumptions required for the purpose of the assessment, such as shielding, working time, environmental migration and long-term barrier performance. Confirm that the actual conditions fall within the range of applicability of the calculation tools.
5. When using the §10 API of the standard, check the implementation's request and response formats and units against it. Do not use calculated values with unresolved missing data or unit interpretation directly in compliance determinations.
6. Review individual cumulative dose and collective dose separately. Even when the dose.exceeded event of §10.3 of the standard is used, state the subject of the determination, the assessment period and the criteria applied, and do not treat the event name itself as a complete determination rule.
7. Link and retain calculation inputs, results, and the basis for review and approval. Check emergency contacts, procedures and drill schedules against actual operating documents, and track the implementation of improvements identified in assessments and drills.

Chapter Summary

• Justification, optimization and dose limits each perform a different judgement. ALARA includes a reasonableness that takes economic and societal factors into account.
• For worker effective dose, the five-year average of 20 mSv/year and the single-year maximum of 50 mSv are checked together. For the lens of the eye limit, the national legislation applicable to the facility is followed, but review is against the current international recommendation (20 mSv/year averaged over 5 years, no more than 50 mSv in a single year) rather than the former value of 150 mSv in the standard's table.
• Post-closure exposure of future generations is addressed through the protection system for planned exposure and the assessment of potential exposure.
• Individual dose and facility safety are assessed by adding source, geometry, pathways, period and uncertainty to shielding thickness, waste package dose rate and environmental concentrations.
• WIA data and the API are used within the scope defined by the standard. The functions of separate assessment records or tools are not turned into mandatory WIA fields or conformance requirements.
• Individual and collective doses are aggregated with the subject and period stated. Worker rotation or a low average dose alone cannot demonstrate a reduction in collective dose or compliance with individual limits.
• Public protection, long-term safety assessment and emergency response must allow the basis of measurements, calculations and decisions to be traced, and the existence of records is distinguished from actual protection performance.

References

WIA Standards — WIA-ENE-026 — Chapter 7. Hongik Ingan (弘益人間) — Benefit All Humanity.