Chapter 2: Spent Nuclear Fuel Management

Spent nuclear fuel continues to release heat and emit radiation after it has been removed from a reactor. Its management requires cooling, shielding, confinement, criticality control, material accountancy and the maintenance of reliable records, all working together. Wet storage, dry storage, transport and eventual disposition each require an appropriate safety assessment. The waste package record links the evidence needed for decisions at each of these stages.

WIA-ENE-026 v1.0 defines data structures for waste packages, storage facilities, monitoring and traceability in the management of radioactive waste. PHASE-1, Foundation & Classification, establishes the classification scheme and its scientific basis, and lays the groundwork for regulatory review and the development of tracking systems. This chapter relates that foundation to spent nuclear fuel management, engineering assessment and jurisdiction-specific licensing. Consistent data formats support the exchange of information, while facility safety and transport authorization are judged on the basis of the applicable assessments and approvals.

2.1 Classification and Inventory of Spent Nuclear Fuel

Spent nuclear fuel contains residual uranium, fission products and transuranic elements formed during irradiation. These radionuclides differ in half-life and radiation characteristics. During storage the radionuclide composition changes; for example, the radioactive decay of parent radionuclides produces daughter radionuclides. Assessments of activity and heat generation must therefore take account of a radionuclide inventory that changes with time.

The classification table in WIA-ENE-026 v1.0 uses HLW-1 for spent nuclear fuel, whereas the extended classification in PHASE-1 uses HLW-SF. The relationship between the two codes follows the Standard Reading Notes, item 3, in Section 1.3. The waste package examples in this chapter use HLW-1, in line with the waste package data structure. The legal classification and the authorized disposition route in each jurisdiction must be established separately. Systems that additionally exchange jurisdiction-specific classifications should document the meaning and regulatory basis of each extension.

2.1.1 Counts, Mass and Fuel History

Inventory management must distinguish between reactors, sites, storage facilities, containers, storage positions, and fuel assemblies or fuel bundles. A single site may host several reactors and storage facilities, and a single container may hold several fuel assemblies. Converting a number of fuel bundles into a number of canisters requires a documented loading configuration, and uranium mass must be confirmed independently of both counts. Comparisons between reactor types also require a consistent mass basis. Typical pressurized water reactors (PWRs) generally use low-enriched uranium as their initial fuel, whereas many pressurized heavy water reactor (PHWR) designs use natural uranium fuel. Natural uranium and low-enriched uranium are distinct terms and should be used as such.

Uranium quantity expressed in tonnes, heavy-metal mass, total fuel assembly mass and the weight of a loaded container are different quantities. In particular, containerInfo.weight.loaded records the weight of the container with its contents loaded. Determining the uranium inventory requires the mass of uranium contained in those contents.

Burnup expresses the energy extracted per unit mass of fuel material. WIA-ENE-026 v1.0 expresses it in MWd/tU, that is, megawatt-days per tonne of uranium. Burnup, discharge date and cooling time help to characterize spent nuclear fuel, but the irradiation history, initial composition, mass and condition of the fuel must also be considered. This additional information can affect heat generation, shielding calculations and acceptance for storage or transport.

2.1.2 Inventory and Available Capacity

WIA-ENE-026 v1.0 defines interfaces for facility inventory and capacity. The facility capacity structure records design, current and available values in m3 for each of HLW, ILW and LLW. All of these values are volumes. Available positions in pool racks, dry storage positions, numbers of fuel assemblies and allowable heat load each require their own stated basis for capacity estimation.

Any utilization calculation should state its numerator, denominator and reference date. For example, the number of occupied rack positions divided by the number of positions available for the fuel under assessment, and the volume occupied by waste divided by the design volume, are different indicators. A worked example of volume-based utilization divides capacity.current by capacity.design for the same waste class, after the units have been aligned. Even where physical space remains, the capacity actually usable may be reduced by thermal limits, fuel compatibility, handling space and space that must be held in reserve for operations.

A cumulative inventory represents the quantity present at a reference time, while annual arisings represent the quantity added during a stated period. When comparing reactor types, the quantity used as the denominator should be stated. The number of waste packages, the number of fuel assemblies in those packages and the total mass of material should each be counted separately on their own basis.

Receiving systems should check the uniqueness of identifiers, the referenced facilities, units of measurement, reference dates, location information and the consistency of chain-of-custody events. When waste package records are converted into an organization's reports, the scope of aggregation and the classification rules should also be clearly defined.

2.2 Decay Heat and Its Calculation Basis

Radioactive decay continues to release energy after a reactor has shut down and the sustained power generation from the operating chain reaction has stopped. The part of that energy deposited in the fuel and surrounding materials is converted to heat. This decay heat is why fuel requires cooling even after discharge. The magnitude of decay heat depends on operating history, fuel composition, burnup and the time elapsed since shutdown. Because the contributions of individual radionuclides change with time, describing the total heat generation generally requires the half-lives and contributions of several radionuclides to be considered together.

2.2.1 An Illustrative Power-Law Approximation

The Way–Wigner approximation represents the decline of decay heat after operation at constant power with a single power law.

P_d(t) / P_0 ≈ 0.066 * [t^(-0.2) - (t + T_irr)^(-0.2)]

Here P_d(t) is the estimated decay heat, P₀ is the thermal output assumed to have been held constant during operation, T_irr is the irradiation period and t is the time elapsed since shutdown. The subscript irr is attached to the irradiation period to distinguish it from T_½, which denotes half-life (Section 1.4.2). P₀ is thermal output, as distinct from electrical output. Both T_irr and t are numerical values in seconds; changing only the time unit while leaving the formula unchanged gives a different result. The formula applies for positive t; at the moment of shutdown, t = 0, it diverges and yields no finite value.

The leading coefficient varies between sources, at about 0.065–0.066, depending on assumptions such as the energy recovered per fission. Sources citing Todreas and Kazimi use 0.066, and a review of the Way–Wigner approach by a working party of the OECD Nuclear Energy Agency (NEA) gives 0.065. The examples in this chapter use the coefficient 0.066. Whichever coefficient is used, the formula is an approximation for estimating the magnitude of decay heat.

This worked example assumes five years of operation at constant power and defines one year as 365.25 days, so T_irr is 157,788,000 seconds. Substituting directly into the formula and multiplying P_d(t)/P₀ by 100 gives the results below. The operating conditions are a hypothetical example chosen to illustrate the calculation, and the results are approximate values obtained with the coefficient 0.066.

The table shows the mathematical result of an approximation under the stated operating assumptions. Applying estimated heat values to storage design requires an assessment of the actual fuel composition, irradiation history and the applicability of the calculation method. The applicability of the method needs particularly careful review at early cooling times and at long cooling times. Among published sources treating power-law approximations of the same form, one gives the range of applicability as roughly 10 seconds to 100 days after shutdown. On that range, the 1-second, 1-year and 10-year values in the table lie outside it and should be read as extrapolations. The heat loads of fuel assemblies and casks likewise each need a calculation basis appropriate to them.

2.2.2 Calculation Standards and Qualified Assessments

The American Nuclear Society (ANS) standard ANSI/ANS-5.1, Decay Heat Power in Light Water Reactors, is a relevant calculation standard for light water reactors. As of October 2026, the edition in force is ANSI/ANS-5.1-2014 (reaffirmed 2019). Demonstrating the validity of a calculation requires a statement of the method and the edition of the standard used, the assumptions about the fuel and irradiation conditions to which it is applied, the contributions of the relevant radionuclides, the treatment of uncertainty and the basis for verification. The correctness of the software, its applicability to the fuel under assessment and its acceptance in licensing assessments must each be supported by evidence.

Activity and heat generation are related but different physical quantities. Activity is the number of radioactive transformations per second, measured in becquerels. Heat generation depends on the energy deposited by those transformations and their products. Assessments of an irradiated fuel assembly or a loaded cask must therefore consider the radionuclide inventory, the ingrowth of daughter radionuclides and energy deposition, as well as the activity of individual isotopes.

Calculation records should link results to the calculation method and input conditions and should distinguish calculated values from measured values. Fuel with the same burnup and discharge date can still differ in decay heat if its composition and irradiation history differ.

2.2.3 Assembly and Package Normalization

The heat load of a waste package is the sum of the heat loads of its individual contents at a common reference time. Where an assessment gives a mass-specific heat generation rate for each fuel item, the total heat load can be expressed as follows.

Q_package(t) = sum over fuel items i of [m_i * q_i(t)]

In this expression m_i is the mass of fuel material on a stated basis and q_i(t) is the mass-specific heat generation rate on the same basis. For example, an initial uranium mass in kilograms combined with a heat generation rate in watts per kilogram of initial uranium is a pairing with a consistent mass basis. The calculation also requires the relevant irradiation and cooling history of each fuel item.

Estimating the total heat load of a cask requires the number of fuel assemblies, the average burnup and the operating period, together with the mass basis and the irradiation and cooling history of each fuel item. A reported value should state whether it refers to one fuel assembly, one fuel bundle, one loaded canister or an entire storage module. Thermal acceptance depends not only on the total heat load but also on how heat is distributed within the approved loading configuration.

2.3 Safety of Wet and Dry Storage

2.3.1 Wet Storage

A spent fuel pool uses water to remove heat and to provide radiation shielding. Maintaining these safety functions requires that the water inventory, heat removal capacity, fuel arrangement and water quality conditions defined in the assessment be maintained. Criticality control depends on keeping geometry, storage racks and neutron absorbers in an approved combination and, where applicable, on also controlling the concentration of neutron absorber dissolved in the water. Pool depth, temperature, water chemistry and boron concentration should be interpreted against the design of the specific reactor and storage facility.

The record for each fuel item should link its location, discharge history, condition and supporting assessment. Pool temperature and water level observations describe the storage environment. Assessing fuel rod temperature, fuel integrity, shielding and rack integrity each requires its own evidence. Changes in fuel arrangement or storage conditions should be assessed against the facility's approved operating basis.

2.3.2 Dry Storage

Dry storage places sufficiently cooled fuel in a storage system that provides heat removal, shielding, confinement and criticality control. Depending on the design, a sealed canister or cask functions together with a basket, shielding structures and heat transfer paths. Configurations include metal casks and canisters placed inside concrete shielding modules. Canisters, external shielding structures and storage modules differ both in function and in the unit by which they are counted.

Many designs rely on passive cooling, removing heat by conduction, radiation and natural convection. The performance of passive cooling also depends on the configuration and environmental conditions defined in the assessment. The condition of air passages, ambient temperature and container arrangement affect heat removal, and container spacing and storage area layout also influence thermal interaction.

Suitability for dry storage is determined by the approved system and the contents to be loaded. Relevant factors include fuel geometry, enrichment, burnup, cooling history, physical condition and calculated heat load. Demonstrating suitability requires cooling time to be considered together with these other relevant factors. A stated container capacity should likewise be interpreted on the basis of the fuel types and loading configurations included in the relevant assessment.

WIA-ENE-026 v1.0 gives a wet storage period of 40–60 years in the HLW-1 classification table. PHASE-1 describes a decay heat cooling period of 30–50 years for HLW-SF. Each period is a descriptive value for its own classification (Standard Reading Notes, item 4, in Section 1.3). The actual date on which fuel is transferred is determined by the assessed heat load, condition and loading configuration of the particular load and by the applicable approvals.

2.3.3 Design Requirements and Acceptance Limits

US 10 CFR Part 72 sets out licensing requirements for the independent storage of spent nuclear fuel, high-level radioactive waste and reactor-related GTCC (Greater Than Class C) waste. Section 72.122 (Overall requirements) belongs to the General Design Criteria in Subpart F. Canister surface dose rate, criticality margin and cladding temperature limits should be managed together with a statement of the basis and scope that apply to the particular storage system.

Acceptance criteria should be preserved together with their source and scope of application. Radiation criteria can differ depending on whether they address workers, members of the public, package surfaces or transport vehicles. Criticality assessments address calculational bias, uncertainty and the conditions assessed. Temperature criteria for fuel and cladding depend on the fuel, the operational stage and the approved assessment. Normal operation, off-normal conditions and accident conditions are separate categories, and this distinction should be preserved in records and in translation.

Storage assessments therefore require the relevant facility licence, system approval and supporting safety analysis. Measured values and alarm setpoints represent, respectively, observations and the thresholds for response. Judgements on these values follow the applicable regulatory criteria and design criteria.

2.4 Transfer and Transport

Moving fuel within a licensed facility and transporting a package under the regulations for the transport of radioactive material are different activities. Moving fuel from a pool to a storage installation within a facility requires a safety basis for that facility and its equipment; whether a Type B(U)F package is needed is then determined by the requirements applicable to that movement. Shipments subject to the transport regulations require a package type, approvals and conditions appropriate to the contents and the journey. Deciding whether the regulations apply should take into account not only distance but also the route and the facility operations and transport arrangements, including handling equipment.

IAEA SSR-6, Regulations for the Safe Transport of Radioactive Material, distinguishes requirements for normal conditions of transport from those for accident conditions of transport. In the designation Type B(U)F, B(U) denotes a Type B(U) package design that, as a rule, requires unilateral approval by the competent authority of the country of origin of the design, and the appended F indicates a design for fissile material (paras 205, 808 and 832 of Rev. 1). A package design for fissile material may, however, require multilateral approval under additional requirements (para. 808). The designation is distinct from generic names for spent fuel storage containers. The editions of SSR-6 and the package types are discussed in detail in Chapter 4. The test paragraphs cited in this section (paras 719, 727 and 728) have the same content in the 2018 Edition (Rev. 1) and in the 2025 Edition (Rev. 2) that superseded it.

The fire temperature is an external test condition and is applied separately from the allowable operating temperature of fuel cladding. Compliance is assessed against the applicable acceptance criteria for release of radioactive material, shielding and criticality. Containers used for both storage and transport must have a basis for each use.

Records should distinguish a storage container identifier from the associated transport approval. WIA-ENE-026 v1.0 defines containerInfo and regulatory.transportApproval; the approval object records approvalNumber, validUntil and authority. Assessing an actual shipment requires the scope of the approval to be checked against the contents, container configuration and conditions of transport.

Movement plans and records of completed movements serve different purposes. Once an actual transfer is complete, the system should reconcile the storage location, the chain of custody (hand-over and acceptance records) and the receiving organization's confirmation. The record should also state whether each event represents a planned action, a physical movement or an acceptance of responsibility.

2.5 Standards, Regulations and Responsibilities

The reference documents used in the management and disposal of spent nuclear fuel play different roles. A normative reference forms part of the requirements of the document that cites it. Safety guides support assessment practice, regulations establish obligations within their jurisdiction and treaties set obligations for their Contracting Parties. Each reference should be used within its stated scope and the applicable regulatory framework.

Responsibilities should be linked to the actual decisions recorded. The fuel custodian maintains inventory and condition records; the receiving organization confirms the suitability of acceptance against the scope of its own authorization; the qualified assessor provides the analysis supporting the decision; and the competent authority determines the approvals that apply. The fields of WIA-ENE-026 v1.0 support the exchange of this evidence. Legal authority, jurisdiction-specific obligations and national implementation schedules are governed by the respective legal systems.

2.6 A Spent Nuclear Fuel Package Record

WIA-ENE-026 v1.0 organizes information around packageId. Classification, container information, radiologicalData, storage, monitoring and traceability are linked to this identifier. The following is a hypothetical example record showing selected fields, identifiers and events. Each value serves only to illustrate the example. An actual registration request must contain every field required by the receiving system's documented requirements.

{
  "packageId": "RW-2025-KR-00001234",
  "timestamp": "2025-12-25T10:30:00Z",
  "wasteClass": "HLW-1",
  "containerInfo": {
    "containerId": "CONT-HLW-2025-001",
    "containerType": "spent_fuel_cask"
  },
  "radiologicalData": {
    "totalActivity": {
      "value": 1.5e15,
      "unit": "Bq",
      "referenceDate": "2025-12-25"
    },
    "thermalOutput": {
      "value": 2.5,
      "unit": "kW"
    }
  },
  "storage": {
    "storageLocation": {
      "facilityId": "FAC-STORE-KR-001",
      "building": "Building-A",
      "row": "A",
      "column": "15",
      "level": "2"
    },
    "storageType": "dry_storage"
  },
  "traceability": {
    "fuelAssemblyId": "FA-2015-KR-003456",
    "burnup": {
      "value": 45000,
      "unit": "MWd/tU"
    },
    "dischargeDate": "2020-06-15",
    "chainOfCustody": [
      {
        "timestamp": "2020-06-15",
        "event": "discharge_from_reactor",
        "location": "Example reactor pool",
        "responsible": "Pool operations team"
      },
      {
        "timestamp": "2025-11-15",
        "event": "transferred_to_dry_storage",
        "location": "Building-A",
        "responsible": "Storage facility operations team"
      }
    ]
  }
}

Each identifier refers to a different object. packageId identifies the waste package record, containerId identifies the container and fuelAssemblyId refers to a fuel assembly. The waste package structure uses a single fuel assembly identifier. Systems that manage loaded casks should retain the defined field formats while recording, in a contents manifest, the links between multiple fuel assemblies and the container and waste package.

The example thermal output of 2.5 kW is a record value set independently of the approximation in Section 2.2. Using this value in an assessment requires a statement of the normalization basis, the reference time and the supporting calculation or measurement. thermalOutput records a value and unit, and the associated assessment record should preserve the basis of the calculation. totalActivity.referenceDate gives the reference date for activity, while the waste package timestamp gives the time of the package record.

The following separate hypothetical example record shows facility information, a monitoring configuration and a cooling time of five years. The dates in this record were set independently of the previous example and apply to this example only.

{
  "timestamp": "2025-12-25T10:30:00Z",
  "facility": {
    "facilityId": "FAC-NPP-KR-001",
    "facilityName": "Example Nuclear Power Plant",
    "facilityType": "nuclear_power_plant"
  },
  "storage": {
    "storageLocation": {
      "facilityId": "FAC-STORE-KR-001",
      "facilityName": "Example Storage Facility",
      "building": "Building-A",
      "row": "A",
      "column": "15",
      "level": "2"
    },
    "storageType": "dry_storage"
  },
  "monitoring": {
    "monitoringFrequency": "continuous",
    "sensors": [
      {
        "sensorId": "SENS-TEMP-001",
        "type": "temperature",
        "model": "PT-100",
        "location": "container_center",
        "readingInterval": 300,
        "unit": "seconds"
      }
    ],
    "lastInspection": "2025-12-20",
    "nextInspection": "2026-01-20"
  },
  "traceability": {
    "originReactor": "EXAMPLE-REACTOR-01",
    "dischargeDate": "2020-12-25",
    "coolingTime": {
      "value": 5,
      "unit": "years"
    },
    "chainOfCustody": [
      {
        "timestamp": "2020-12-25",
        "event": "discharge_from_reactor",
        "location": "Example reactor pool",
        "responsible": "Example generator facility custodian"
      },
      {
        "timestamp": "2025-12-25",
        "event": "transferred_to_dry_storage",
        "location": "Example Storage Facility Building-A",
        "responsible": "Example storage facility custodian"
      }
    ]
  }
}

The reference date for this cooling-time example is 2025-12-25, five calendar years after the discharge date of 2020-12-25. Implementations should document the rules for expressing reference dates and elapsed time in years. This calendar-based example and the approximation in seconds in Section 2.2 each use their own stated time basis.

The monitoring configuration describes sensors, inspection dates and measurement intervals. continuous describes the overall monitoring mode, while readingInterval gives the measurement interval of an individual sensor; in this example the temperature sensor records a reading every 300 seconds. Actual observations are handled as monitoring readings.

The waste package structure in WIA-ENE-026 v1.0 uses storage.storageLocation, whereas the TypeScript SDK example uses storage.location. Waste package exchange in this chapter follows storage.storageLocation. Where conversion between the two paths is required, the mapping should be defined as an explicit implementation rule (Standard Reading Notes, item 8, in Section 1.3).

Field names and enumerated values remain the same whatever the display language. A user interface may describe HLW-1 as spent nuclear fuel while the exchanged data keeps the code unchanged. Preserving the full identifier strings keeps references between records intact.

2.7 Lifecycle Records and Chain of Custody

A fuel history links discharge, storage, transfer, inspection and disposition. Storage retains the possibility of retrieval, whereas disposal aims at long-term isolation without planned retrieval. Each transition between stages requires evidence appropriate to the receiving system. Planned disposal dates and disposal sites represent future plans, while records of completed transfers and disposal represent actions actually taken and the approvals supporting them.

WIA-ENE-026 v1.0 defines traceability.chainOfCustody entries containing timestamp, event, location and responsible. Together these fields describe what was recorded, when and where it occurred and who was responsible. In the examples, a date-only timestamp gives the date on which the event occurred. Ordering several events on the same day requires additional time information or evidence establishing their sequence.

Implementations should document their controls for authorization, amendment and record retention. Where append-only processing, signature algorithms for hand-over and acceptance, or canonical encoding have been adopted, these should be included as well. An amended record should remain linked to the earlier record it supersedes. These controls are implementation policies applied to the chain-of-custody fields of this standard.

WIA-ENE-026 v1.0 also contains traceability.blockchain and provisions for electronic signatures on IAEA reporting. Blockchain records and the verified value should be interpreted together with the results of checks on the input data, the custodian's authority, the physical inventory and the controls on record changes. Signatures on IAEA reports and controls on chain-of-custody events serve different purposes.

On transfer, the recorded identification and contents should be reconciled against the sending inventory, the receiving inventory and the location at which acceptance was confirmed. A waste package may change location within the same facility while keeping the same custodian, or responsibility may pass along with the movement. Supporting operational records should distinguish between these two kinds of event. Discrepancies must be resolved before the waste package can be treated as reconciled at its destination.

Repackaging and the grouping of fuel assemblies require ongoing management of the relationships between fuel, containers and waste packages. The system's contents manifest should preserve the relationships arising from configuration changes and from the splitting or merging of waste packages. Maintaining these relationships allows the history of individual fuel items to be traced even when the packaging changes.

Cooling, storage and transport periods depend on the fuel under assessment, the authorized system and actual operating practice. Security requirements, including the retention period for audit logs, are covered in Section 4.8.1. Because audit logs record information system activity, the fuel pedigree, safety assessments and the records needed from storage through to disposition require retention policies suited to their continuing use.

2.8 Documented Interfaces for Spent Nuclear Fuel Records

WIA-ENE-026 v1.0 defines waste package, facility, monitoring and calculation interfaces that support spent nuclear fuel management. These interfaces link waste package records with facility information, observations and assessment inputs. The following table collects the complete set of interfaces defined in §10.1 of the standard; other chapters interpret only the paths they use.

Of these, POST /facilities can be used to register the storage facility that will receive spent nuclear fuel and its capacity, and POST /dose/calculate to estimate the dose from handling operations. How a calculated dose is used as a basis for shielding or work planning is decided in the relevant safety assessment.

The waste package list interface supports facility and isotope filters. The relationship between the top-level values of the wasteClass filter and subclass codes such as HLW-1 is summarized in the code layer table in Section 1.3. Interoperability profiles should state whether subclass codes are accepted as filters and which subclasses a top-level class filter includes.

The location endpoint updates location information. Transfer workflows define the consistency rules that link storage type, chain of custody and the receiving inventory to this update. System design should also include methods for detecting and recovering from cases in which only some of the linked updates succeed.

An activity decay request takes an isotope, an initial activity, an elapsed time and a time unit. Assessing the heat generation of spent nuclear fuel additionally requires the fuel inventory, irradiation history and energy deposition basis described in Section 2.2. For a radionuclide with no further production from a parent, the basic relationship is the same as the formula in Section 1.4.2; written in terms of the decay constant, it is as follows.

A(t) = A_0 * exp(-lambda * t)
lambda = ln(2) / T_half

Here A₀ is the activity at the reference time and T_half is the half-life. The decay constant and elapsed time must use compatible units, and the elapsed time and half-life must use the same time basis. Parent–daughter chains may require decay and ingrowth to be calculated together. Differences in the half-life value of Cs-137 follow the Standard Reading Notes, item 1, in Section 1.3. Reproducing a calculation requires the adopted nuclear data values to be stated, and the same basis should be kept when comparing results.

The implementation profile for POST /dose/calculate should define the units of distance, duration and shielding.thickness before numerical request values are exchanged. WIA-ENE-026 v1.0 names the request object for submitting monitoring data MonitoringReading. Interoperating systems should document the permitted fields and the rules for mapping sensor observations to that request, and should distinguish a waste package's sensor configuration from its measurement data.

Security requirements such as API key authentication, OAuth 2.0 authorization, mutual TLS, electronic signatures and end-to-end encryption are summarized in Section 4.8.1. Because spent nuclear fuel records contain the location and composition of nuclear material, implementations should link these measures to the permissions of authenticated principals, key management and access logging.

2.9 Monitoring, Safety Evidence and Long-Term Review

Monitoring becomes more useful when each observation is linked to the safety function it is meant to inform. Temperature readings, radiation measurements, inspection results and nuclear material accountancy records each provide information needed for different judgements. Even when presented together in a waste package management system, these records should remain distinguishable.

2.9.1 Monitoring Configuration and Readings

The waste package monitoring object in this standard includes monitoringFrequency, sensors, lastInspection, nextInspection and alarmThresholds. The monitoring frequency values are continuous, daily, weekly and monthly. Each sensor entry describes identification, type, model, location and measurement interval.

WIA-ENE-026 v1.0 also gives an example of a monitoring system structure in which timestamped observations are held in measurements and calibration information in calibration. The waste package monitoring configuration, actual readings and alarm handling records complement one another. Sensor measurement intervals, communication transmission intervals and external reporting frequencies are each managed as separate schedules.

Interpreting a reading requires clear sensor identification, location, timestamp and unit. Calibration information and measurement uncertainty determine how far the value can be relied upon as a basis for decisions. Comparisons over time also require consistent measurement conditions. Missing readings and readings whose update has been delayed should be distinguishable. In particular, distinguishing a communication loss from a valid zero reading prevents missing data from being mistaken for evidence of normal conditions.

Sensor location determines what a reading means. Cask surface temperature is not a direct measurement of peak fuel cladding temperature. Area dose rate measurements and contact dose rate measurements on a waste package should likewise be applied separately, according to their respective locations and purposes. Each reading should be linked to the physical quantity and location it represents.

alarmThresholds includes doseRateAlarm, temperatureAlarm and contaminationAlarm. These values are used to decide whether to raise an alarm according to the measured quantity, sensor location, assessed facility conditions and the approved response arrangements. Regulatory limits and fuel damage criteria are subject to their own assessment bases. The alert interface distinguishes states such as active, acknowledged and resolved; these states should be interpreted together with the evidence of the response to the physical cause.

2.9.2 Dose Rates and Personal Dose

radiologicalData.doseRate distinguishes the contact dose rate from the dose rate at 1 m. A value at 1 m applies at that position; assessing a position at 2 m requires appropriate measurement or calculation. Personal dose assessment takes account of occupancy time, work location, shielding, the radiation field and the applicable assessment method.

WIA-ENE-026 v1.0 distinguishes an occupational effective dose limit of 20 mSv per year averaged over five years from a limit of 50 mSv in any single year. It gives the annual effective dose limit for members of the public as 1 mSv, excluding natural background radiation and medical exposure. These dose criteria play a different role from facility alarm setpoints and task-specific dose constraints. Current values of the equivalent dose limit for the lens of the eye are discussed in Section 7.2.2.

The ALARA (As Low As Reasonably Achievable) principle, which holds that exposures should be kept as low as reasonably achievable, guides the optimization of exposure. Applying it to storage operations requires consideration of access conditions, working time, distance and shielding, together with the economic and social factors relevant to optimization.

2.9.3 Long-Term Thermal and Condition Review

Long-term review links monitoring trends with inspection results, materials ageing and updated assessments. As storage continues, the radionuclides contributing to heat generation and radiation also change. Because the ingrowth of daughter radionuclides can increase the activity of particular radionuclides, activity and location-specific dose rates may follow different trends.

Even for waste packages with the same total thermal output, the peak temperature can differ depending on the distribution of high-heat fuel, internal heat transfer and external cooling conditions. Package-specific assessments therefore review the inventory and loading configuration of the actual contents in conjunction with the storage environment.

radiologicalData.thermalOutput records a single thermal output value and its unit. For long-term prediction, the system should preserve dated predictions together with the relevant fuel composition, discharge history, calculation method and uncertainty. Such assessment records allow the basis of each prediction to be reviewed later.

storage.storageDuration.planned and storage.storageDuration.maximum record storage durations. Continued use of a storage system also requires the condition of containers, seals, shielding and monitoring systems to be assessed against the applicable design and licence. Changes to the fuel inventory, container condition or storage configuration can affect assessment assumptions.

Preserving the calculation basis applied and the record history allows them to be used in later assessments for retrieval, transport and disposition. Each assessment should confirm that the recorded values remain valid for the current state of the fuel and system.

2.9.4 Access Control and Record Preservation

Spent nuclear fuel records may contain sensitive information about the location and composition of nuclear material and the responsibility for its management. Radiation safety, safeguards and security serve different purposes: protecting people and the environment, supporting the verification of nuclear material, and controlling unauthorized access or interference, respectively. regulatory.safeguardsApplicable records whether safeguards apply, while physical protection categorization is determined by the relevant security regime.

The chain of custody records management events for fuel or waste packages. System audit logs record information system activity, such as who accessed or changed the data. Because the two kinds of record serve different purposes, each needs its own appropriate access controls and retention policy. Apart from the audit log retention requirements (Section 4.8.1), the need to retain fuel and safety records for longer periods should also be considered.

Long-term records should maintain fuel composition, reference dates, units, calculation bases and chain-of-custody relationships together with the original data and amendment history. When file formats or storage media change, preservation procedures should confirm the consistency and recoverability of the content.

Traceability identifiers link related records. Continued use of the records also requires backup, recovery, format conversion and the succession of access rights. So that future custodians can reconstruct the meaning of the records, the edition of the standard applied and the definitions of organization-specific extensions should be preserved with them.

Chapter Summary

• The example spent nuclear fuel waste packages under WIA-ENE-026 v1.0 use HLW-1, while HLW-SF in PHASE-1 keeps its meaning within that classification scheme.
• Assessments of inventory and available space distinguish the numbers of reactors and sites, fuel counts, material mass, storage volume and thermal capacity.
• The Way–Wigner approximation takes time in seconds. The coefficient varies between sources, at about 0.065–0.066, depending on the assumptions; the examples in this chapter use 0.066. Assessing the heat generation of a waste package also requires its contents, normalization basis, reference time and irradiation history.
• Wet storage, dry storage and transport each require their applicable safety assessments, acceptance criteria and approvals.
• SSR-6 normal-condition tests and accident-condition tests are separate categories. Requirements for Type B(U)F packages depend on the transport activity and package concerned.
• SSG-23 addresses the safety case and safety assessment for disposal; GSG-3 addresses the safety case and safety assessment for predisposal management.
• Waste package, facility and monitoring interfaces are linked with chain-of-custody entries containing timestamp, event, location and responsible. Activity decay calculations and dose calculations are each used for their defined purposes.
• Monitoring results are interpreted according to their safety purpose, dose rates are distinguished from personal dose, and the supporting assessments are preserved. Long-term fuel records require retention policies suited to their continuing use.

References

WIA Standards — WIA-ENE-026 — Chapter 2. Hongik Ingan (弘益人間) — Benefit All Humanity.