ML20321A107

From kanterella
Jump to navigation Jump to search
Chapter 12 - Conduct of Operations Evaluation
ML20321A107
Person / Time
Issue date: 04/30/2020
From:
Office of Nuclear Material Safety and Safeguards
To:
AWillis - 301.415.0479 ; SKumar
Shared Package
ML20321A086 List:
References
NUREG-2215
Download: ML20321A107 (46)
v • d • e


Text

12-1 CONDUCT OF OPERATIONS EVALUATION Review Objective The objective of the U.S. Nuclear Regulatory Commissions (NRCs) conduct of operations review is to ensure that the applicant has (1) described an appropriate infrastructure to manage, test, operate, and maintain the facility, including provisions for effective training, emergency planning, and physical security programs for a dry storage facility (DSF), and (2) developed appropriate acceptance tests and maintenance programs to ensure that its dry cask storage system (DSS) or DSF structures, systems, and components (SSCs) are fabricated and maintained in accordance with the design described in the safety analysis report (SAR).

Applicability This chapter applies to the review of an applicants SAR with respect to licenses for an independent spent fuel storage installation (ISFSI) or a monitored retrievable storage installation (MRS), categorized as a DSF. It also applies to the review of an applicants SAR for a certificate of compliance (CoC) of a DSS for use at a general license ISFSI. Sections and tables of this chapter that apply only to a specific license for a DSF application have (SL) in the heading. A section, paragraph, or table without an identifier applies to both types of applications.

Some of the review procedures in this chapter relate to the conduct of operations associated with spent nuclear fuel (SNF) pools. Carefully review all of the review procedures in this SRP for applicability to SNF pools.

Areas of Review This chapter addresses the following areas of review, which can have an impact on SSCs important to safety:

organizational structure (SL) acceptance tests preoperational testing and startup operations (SL) maintenance program normal operations (SL) personnel selection, training, and certification (SL) emergency planning (SL) physical security and safeguards contingency plans (SL)

Regulatory Requirements and Acceptance Criteria This section summarizes those parts of Title 10 of the Code of Federal Regulations (10 CFR) Part 72, Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater than Class C Waste, that are relevant to the review areas this chapter addresses. The NRC staff reviewer should refer to the exact language in the regulations. Tables 12-1a and 12-1b match the relevant regulatory requirements to the areas of review covered in this chapter.

12-2 12.4.1 Organizational Structure (SL)

The SAR should describe the organizational structure and administrative control system that will be used for the proposed ISFSI or MRS (i.e., through construction, preoperational testing and initial operations, normal operations, and decommissioning) as required in 10 CFR 72.24(h).

Chapter 10A, Radiation Protection Evaluation for Dry Storage Facilities, of this SRP, particularly Sections 10A.4.4, Health Physics Program, and 10A.4.4.1, Organization and Staffing, for regulatory requirements, and 10A.5.4, Health Physics Program, and 10A.5.4.1, Organization and Staffing, for review procedures, provides additional guidance on the information the SAR should include as related to the radiation protection and health physics aspects of the organizational structure and staffing.

Corporate Organization (SL)

The SAR must describe the corporate organization responsible for the ISFSI or MRS (10 CFR 72.24(h)), which should include organization charts and position descriptions. If the corporation is made up from two or more corporate identities, the SAR should describe the relationship and responsibilities between each entity.

The applicant must demonstrate the financial capabilities of the corporation to construct, operate, and decommission the installation, as required in 10 CFR 72.22(e). The scope of this SRP does not include specific guidance for reviewing the financial qualifications required in 10 CFR 72.22(e)(1) and 10 CFR 72.22(e)(2); this information is part of the application but separate from the SAR. However, Chapter 14, Decommissioning Evaluation, of this SRP does provide guidance for reviewing financial qualifications regarding decommissioning, as required in 10 CFR 72.22(e)(3) and 10 CFR 72.30, Financial assurance and recordkeeping for decommissioning, as part of the SAR. The Department of Energy (DOE) is exempt from financial assurance requirements per 10 CFR 72.22(e), 10 CFR 72.40(a)(6), and 10 CFR 72.40(a)(10).

Financial reviews should be coordinated with the NRC Office of Nuclear Reactor Regulation. The NRC project manager should ensure that the application contains financial data, in accordance with 10 CFR 72.22(e), that shows that the licensee can carry out the activities being sought for the requested duration. Information should state where the activity will be performed, the general plan for carrying out the activity, and the period of time for which the license is requested.

Table 12-1a Relationship of Regulations and Areas of Review for a DSF (SL)

Areas of Review 10 CFR Part 72 Regulations 72.22(e) 72.24 72.28 72.30 72.32 72.40(a) 72.82(d) 72.122 (a)(f) 72.124(b)

Organizational Structure (h)

(13)

Acceptance Tests (h)(i)

(a)

Preoperational Testing and Startup Operations (h)(i)(p)

(13)

Maintenance Program (13)

(f)

Normal Operations (h)

(5)(13)

Personnel Selection, Training, and Certification (h)(j)

(4)(9)

Emergency Planning (k)

(11)

Physical Security and Safeguards Contingency Plans (o)

(8)(14)

Areas of Review 10 CFR Part 72 Regulations (cont.)

72.156 72.162 72.174a 72.180 72.184 72.190 72.192 72.194 Organizational Structure (SL)

Acceptance Tests Preoperational Testing and Startup Operations (SL)

Maintenance Program Normal Operations (SL)

Personnel Selection, Training, and Certification (SL)

Physical Security and Safeguards Contingency Plans (SL) a This requirement specifies the retention of quality assurance records. Other requirements related to the retention of records include 10 CFR 72.70, Safety Analysis Report Updating, 72.72(a), 72.44(b)(4), 72.74, Reports of Accidental Criticality or Loss of Special Nuclear Material, 72.76, Material Status Report, 72.78, Nuclear Material Transaction Reports, and 73.21, Protection of Safeguards Information: Performance Requirements.

12-3

12-4 Table 12-1b Relationship of Regulations and Areas of Review for a DSS (CoC)

Areas of Review 10 CFR Part 72 Regulations 72.124(b) 72.162 72.234(a) 72.236 Acceptance Tests (j)(l)

Maintenance Program (g)

The SAR should describe the corporate functions, responsibilities, and authorities related to each aspect of the installation (e.g., design, engineering, construction, quality assurance (QA), testing).

The SAR must describe the inhouse organization and technical staff (e.g., numbers of personnel, qualifications, educational and experience backgrounds), as required in 10 CFR 72.28. The SAR should describe the relationship between the applicants inhouse organization and outside contractors and suppliers, including the extent of dependence on those sources for design, construction, QA, and other functions.

The applicant should also describe the relationship between the corporate and onsite organizations and explain the nature of interaction between corporate management and the site related to health and safety, including any role in policy/procedure development, audits, inspections, and investigations.

Onsite Organization (SL)

The SAR should describe the onsite organization, including organizational charts and position descriptions with emphasis on positions that perform functions important to safety. Such positions include, but are not limited to, those with responsibilities in health physics, nuclear criticality safety, training and certification, emergency planning and response, operations, maintenance, engineering, and QA.

The discussion of positions and responsibility should illustrate how these functions or aspects of these functions are performed, including the degree of separation between the facility operations organization and other parts of the onsite organization that perform functions important to safety.

The SAR should also identify alternates who are authorized to act in the absence of individuals assigned to key positions and identify which positions have shut-down or stop-work authority for health or safety reasons.

The SAR should identify minimum staffing levels for major entities within the onsite organization.

The SAR should identify whether the onsite organization includes a safety committee (or committees) and describe the membership, duties, responsibilities, operating characteristics, and reporting function of proposed safety committees.

Identification of Agents and Contractors (SL)

The SAR should identify the prime agents or contractors for the design, construction, and operation of the installation. All principal consultants and outside service organizations, including those providing QA services, should be identified. The SAR should clearly define the division and assignment of responsibilities among these parties.

12-5 Management and Administrative Controls (SL)

The SAR should describe the proposed management and administrative control system (10 CFR 72.24(h)), including provisions for the following:

administrative and general plant procedures including implementation of good radiation protection practices and objectives to ensure occupational exposures will remain as low as is reasonably achievable (ALARA) a program of surveillance, testing, and inspections of items and activities important to safety periodic independent audits change control employee training and certification programs records preparation and maintenance Administrative procedures address planning, administrative controls, and document issuance.

The procedures provide rules and instructions on personnel conduct, preparation and retention of plant documents, and interfaces among plant organizations. General facility procedures prescribe the actions required to achieve safe operation and provide necessary instruction for the operation and maintenance of facility systems and equipment, including implementation of good radiation practices and ALARA objectives. The SAR should describe the program for preparation, review, change, and approval of procedures. The applicant should also identify the onsite organizations that use procedures and the activities or operations that are covered by such procedures.

Sections 12.4.5.1 and 12.5.5.1 below provide guidance on evaluating procedures for normal plant operation.

The applicant should describe the program of surveillance, testing, and inspection to ensure satisfactory inservice performance of items and activities important to safety. The description should address the development and use of procedures that set forth the steps to be taken and identify the standards or criteria to be applied. The program should include provisions for the following:

preoperational testing (see Sections 12.4.3 and 12.5.3 below) to demonstrate facility operability and identify conditions adverse to safety operational testing and surveillance to verify and record characteristics of facility equipment and components surveillance, testing, and inspection after modification or when corrective actions have been completed The management control system description should also include requirements for planned and scheduled internal and external audits to evaluate the application and effectiveness of management controls, facility procedures, and other activities affecting safety. The audit program should describe audit frequency, methods for documenting and communicating audit findings, resolution of issues, and implementation of corrective actions.

12-6 The applicant should also describe the system for change control, including how change control is integrated into the management control system. The SAR should describe the coordination of change between and among potentially affected organizations (e.g., engineering, operations, maintenance, training). The SAR should describe how operations are shut down to effect changes and how all facility equipment and procedural changes are completed. The training of staff before resumption of operations should also be addressed.

The management system description should also include the system for maintaining records of facility operation (as addressed in Sections 12.4.5.2 and 12.5.5.2 below).

12.4.2 Acceptance Tests The acceptance tests demonstrate that the DSS or DSF SSCs and features have been fabricated in accordance with the design criteria and that the initial operation of the DSS or DSF SSCs and features complies with regulatory requirements. A comprehensive evaluation should encompass, but may not be limited to, the following acceptance tests:

structural/pressure tests leak tests visual and nondestructive examination (NDE) inspections shielding tests neutron absorber tests thermal tests storage container identification In general, the acceptance tests outlined in the SAR should cite appropriate authoritative codes and standards. Table 12-2 lists the standards and codes the NRC has previously accepted as the regulatory basis for the design, fabrication, inspection, and testing of SNF storage system and container components. The SAR should clearly identify any exceptions to the listed codes and standards (see SRP Chapter 17, Technical Specifications Evaluation).

12-7 Table 12-2 Acceptable Regulatory Basis for the Design, Fabrication, Inspection, and Testing of DSS or DSF Components System/Component Acceptable Regulatory Basis Confinement System American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel (B&PV) Code,Section III, Rules for Construction of Nuclear Facility Components, Division 1 American National Standards Institute (ANSI) N14.5, Radioactive MaterialsLeakage Tests on Packages for Shipment Confinement Internals (e.g., basket)

ASME B&PV Code,Section III, Subsection NG Metal Cask Overpack ASME B&PV Code,Section VIII, Rules for Construction of Pressure Vessels Concrete Cask Overpack American Concrete Institute (ACI) 318, Building Code Requirements for Structural Concrete and Commentary; ACI 349, Code Requirements for Nuclear Safety-Related Concrete, as appropriate Other Metal Structures ASME B&PV Code,Section III, Subsection NF American Institute of Steel Construction 360, Specification for Structural Steel Buildings.

12.4.3 Preoperational Testing and Startup Operations (SL)

The SAR must describe the plans for preoperational testing and initial facility (startup) operations (10 CFR 72.24(g)). Regulatory Guide (RG) 3.62, Standard Format and Content for the Safety Analysis Report for Onsite Storage of Spent Fuel Storage Casks, and RG 3.48, Standard Format and Content for the Safety Analysis Report for an Independent Spent Fuel Storage Installation or Monitored Retrievable Storage Installation (Dry Storage), provide guidance on the information to be included in the SAR related to preoperational testing and startup operations.

The preoperational testing and operating startup activities are determined by the type of radioactive material to be stored.

The SAR should describe the administrative procedures used for conducting the testing and startup activities. This description should include the system to be used for preparing, approving, and executing the test procedures and for evaluating, documenting, and approving test results.

Provisions should be made for incorporating changes to the system or individual procedures on the basis of inadequacies in test procedures or unexpected test results. The organizational responsibilities for administering the system should be identified, and the qualifications of involved personnel should be described.

12.4.3.1 Preoperational Testing Plan (SL)

The preoperational testing plan description should identify the testing objectives and the general methods to meet those objectives. The SAR should identify each item (facility, component, piece of equipment, operation) to be tested. For each physical or operational item, the applicant should provide the following information:

the type of test to be performed the expected response

12-8 the acceptable margin of difference from the expected response the method of validation (if applicable) appropriate corrective action for unexpected or unacceptable results If the proposed ISFSI or MRS contains any SSCs important to safety for which functional adequacy or reliability has not been demonstrated or otherwise validated, the preoperational test plan must include a description and schedule showing how these safety questions will be resolved before the initial receipt of the radioactive materials to be stored (10 CFR 72.24(i)).

12.4.3.2 Startup Plan (SL)

The operating startup plan should identify those specific operations involving the initial handling of radioactive material to be placed into storage. Although the operating startup plan does not necessarily include the facility procedures to be used for normal operations or during steady-state conditions, the plan should evaluate the effectiveness of those procedures. For considerations related to the ALARA principle, the applicant should perform as many of the operating startup actions as feasible during preoperational testing (i.e., before sources of exposure are present).

The operating startup plan should include the following elements:

tests and confirmation of procedures and exposure times involving actual radioactive sources (e.g., radiation monitoring) direct radiation monitoring of storage containers (and other SSCs used to handle or contain radioactive materials) and shielding for radiation dose rates, streaming, and surface hot spots and containment surveys verification of effectiveness of heat removal features documentation of results of tests and evaluations 12.4.4 Maintenance Program The maintenance program describes actions that the licensee needs to implement during the storage period to ensure that the DSS or DSF SSCs and features perform their intended functions. A comprehensive evaluation should identify and describe the necessary maintenance programs and address the following for each of the identified maintenance programs:

inspection tests repair, replacement, and maintenance In general, the maintenance programs outlined in the SAR should cite appropriate authoritative codes and standards. The NRC has previously accepted the codes and standards listed in Table 12-2 as the regulatory basis for the design, fabrication, inspection, and testing of SNF storage system components.

12-9 12.4.5 Normal Operations (SL) 12.4.5.1 Procedures (SL)

The SAR should describe or state that the licensee (i.e., applicant as the licensee) will conduct all facility operations that are important to safety according to detailed written procedures that are based on and consistent with the operations descriptions in the operating procedures chapter of the SAR and the acceptance tests and maintenance program descriptions in the conduct of operations chapter of the SAR. The SAR should also state that proposed procedures and revisions will be reviewed and approved by the health, safety, and QA organizations that are independent from the operating management function.

The identification of proposed written procedures should include all routine and projected contingency operations. The applicant should also describe the review, change, and approval practices for all operating, maintenance, and testing procedures. This description may refer to the appropriate management controls addressed in Section 12.4.1.4 above.

The listing of operations requiring written procedures should include the following, as applicable to the ISFSI or MRS:

all operations identified in the proposed technical specifications all operating, maintenance, testing, and surveillance functions important to safety The procedures listed should clearly indicate, by title or subject, their purpose and applicability.

The applicant should identify any standards used for the preparation of these procedures.

12.4.5.2 Records (SL)

The SAR should describe the management system for maintaining records. This description may refer to the appropriate management controls addressed in Section 12.4.1.4 above. Although all records need not be maintained centrally, the management system should ensure that cognizance is being maintained of all records, the responsible staff, and locations.

Records stored in electronic media will generally be acceptable if the capability is maintained to produce legible, accurate, and complete records over the required retention period. The record format should include all pertinent information, such as stamps, initials, and signatures. The SAR should specify the retention period for each type of record because it varies depending on applicable regulatory requirements. The management system should also provide for adequate safeguards against tampering and loss of records over the retention period.

The SAR should identify, by type, the records to be maintained. Records maintained should include the following:

construction records, as specified in applicable construction codes (e.g., ACI 349) and including as-built drawings and specifications, material certifications, and audit trail to the applicable SSCs, inspection records, test reports, and certifications (10 CFR 72.30(f)(2),

10 CFR 72.156, Identification and control of materials, parts, and components, and 10 CFR 72.174, Quality assurance records)

12-10 as required in 10 CFR 72.30(f)(3), a list of the following, contained in a single document and updated no less than every 2 years:

all areas designated and formerly designated as restricted areas as defined under 10 CFR 20.1003, Definitions all areas outside of restricted areas that require documentation under 10 CFR 72.30(f)(1) (see next entry) records of spills or other abnormal occurrences involving the spread of radiation in and around the facility, equipment, or site (10 CFR 72.30(f)(1))

records of the cost estimate performed for the decommissioning funding plan or of the amount certified for decommissioning and records of the funding method used for ensuring funds, if either funding plan or certifications are used (10 CFR 72.30(f)(4))

(i.e., record copy of proposed decommissioning plan filed with license application, attached decommissioning funding plan, any modifications to these plans, and final decommissioning plan when prepared) receipt, inventory, disposal, acquisition, and transfer of all SNF, high-level radioactive waste (HLW) and reactor-related greater-than-Class-C (GTCC) waste in storage, as required in 10 CFR 72.72(a) (including provisions for duplicate records storage at different locations, in accordance with 10 CFR 72.72(d))

records of physical inventories and current material control and accounting procedures (10 CFR 72.72(b) and 10 CFR 72.72(c))

operating records, including principal maintenance, alternations, or additions made (10 CFR 72.70(b)(1) and 10 CFR 72.70(c)(4)(ii))

records of off-normal occurrences and events associated with radioactive releases (10 CFR 72.44(d)(3))

records of employee certification (10 CFR 72.44(b)(4))

QA records (10 CFR 72.174) environmental survey records and environmental reports, including those related to the radiological environmental monitoring program (see SRP Sections 10A.4.2.5 and 10A.5.2.5 for a description of this program) radiation monitor readings or records (e.g., stripcharts or electronic results) radiation protection program records (per Subpart L, Records, of 10 CFR Part 20, Standards for Protection Against Radiation), including those related to the following:

program contents, audits, and reviews radiation surveys determination of prior occupational dose planned special exposures individual (worker) monitoring results

12-11 dose to individual members of the public radioactive waste disposal tests of entry control devices for very high radiation areas records of changes to the physical security plan (10 CFR 72.44(e) and 10 CFR 72.186, Change to Physical Security and Safeguards Contingency Plans), and other physical security records (10 CFR 73.21 and 10 CFR 73.70, Records) records of occurrence and severity of natural phenomena (10 CFR 72.92, Design Basis External Natural Events) record copies of the following:

SAR, SAR updates, final SAR (10 CFR 72.70) reports of accidental criticality or loss of special nuclear material (10 CFR 72.74 and 10 CFR 73.71, Reporting of safeguards events) material status reports (10 CFR 72.76) nuclear material transfer reports (10 CFR 72.78) physical security plan (10 CFR 72.180, Physical protection plan) other records and reports (10 CFR 72.82, Inspections and tests) report of preoperational test acceptance criteria and test results written procedures The radiation protection records required by 10 CFR Part 20, Subpart L should incorporate the units of curie, rad, and rem, as applicable, including multiples or subdivisions of those units (e.g., megacurie, millicurie, millirem). Where dose is part of a record, the dose quantity used on the record (e.g., total effective dose equivalent, committed effective dose equivalent, shallow dose equivalent) should be clearly indicated. Chapter 10A of this SRP, particularly Sections 10A.4.4 and 10A.5.4, includes additional guidance regarding records as related to the licensees radiation protection and health physics programs and operations.

12.4.6 Personnel Selection, Training, and Certification (SL)

The SAR should describe the organization responsible for personnel selection, training, and certification. The SAR should also describe the program that will be established and implemented to ensure that personnel whose responsibilities include functions that are important to safety will be appropriately qualified and trained. The process of selecting and training security guards should be described. Chapter 10A of this SRP, particularly Sections 10A.4.4 and 10A.5.4, includes additional guidance regarding personnel selection, training, and certification that relate to the radiation protection and health physics organization personnel and radiation safety training for all licensee personnel.

12.4.6.1 Personnel Organization (SL)

The SAR should include a discussion of the organization and management of the training component and should identify the personnel responsible for the development of training

12-12 programs, conducting training and retraining of employees (including new employee orientations),

and maintaining up-to-date records on the status of trained personnel.

12.4.6.2 Selection and Training of Operating Personnel (SL)

The applicant should identify the functions that are important to safety and describe the qualifications for personnel performing those functions. These personnel qualifications should include the following:

minimum qualification requirements for operating, technical, and maintenance supervisory personnel, including any qualification requirements identified in the evaluations throughout the SAR (e.g., certification requirements for individuals writing procedures for and performing leakage testing identified in the operating procedures and conduct of operations chapters of the SAR) qualifications, in resume form, of persons who will be assigned to managerial and technical positions.

The program description should identify the scope of operational and safety training. Operational training should include topics such as installation design and operations, instrumentation and control, methods of dealing with operating functions, decontamination procedures, and emergency procedures. Radiation safety training should include topics such as the nature and sources of radiation, methods of controlling exposure and contamination, radiation monitoring, shielding, dosimetry, biological effects, and criticality hazards control.

The SAR should list the type and level of training to be provided for each job description (personnel classification), including specific training provided to specific job descriptions.

Alternatively, the SAR may describe the basis used to identify the type and level of training by job description.

The SAR should clearly identify the requirements for the certification of personnel who will operate equipment and controls that are important to safety. The requirements must address the physical condition and general health of personnel to be certified in accordance with 10 CFR 72.194, Physical requirements.

The SAR should describe methods of testing to determine the effectiveness of the training program. The applicant should evaluate the effectiveness of the training program against established objectives and criteria, identifying any standards used for development and implementation of the training program.

The SAR should describe the frequency of retraining, and the nature and duration of the retention of training and testing records. Retraining should be periodic and not less than every 2 years.

Training records should be kept up to date and retained for a minimum of 3 years.

The SAR should describe implementation of the training program before conduct of operations involving radioactive material (i.e., preoperational training). The applicant should commit to a substantial completion of staff training and certification before the receipt of the radioactive material for storage.

The applicant should identify any standards used for the selection, training, and certification of personnel.

12-13 12.4.6.3 Selection and Training of Security Guards (SL)

The SAR must describe the process by which security guards (including watchmen, armed response persons) are selected and qualified (10 CFR 73.55(c)(4)). This information may be submitted as part of the applicants physical security plan.

The criteria used must conform to the general criteria for security personnel contained in Appendix B, General Criteria for Security Personnel, to 10 CFR Part 73, Physical Protection of Plants and Materials. RG 5.20, Training, Equipping, and Qualifying of Guards and Watchmen, provides guidance in this area.

12.4.7 Emergency Planning (SL)

The purpose of the review of the applicants emergency plan (EP) is to ensure that the plan (1) complies with regulatory requirements, (2) is based on the proposed ISFSI or MRS, and (3) provides acceptable hazards analysis. The emergency planning regulations in 10 CFR 72.32 have additional requirements for a MRS and an ISFSI that is licensed to process and/or repackage spent fuel. For the ease of clarification, the evaluation criteria for these additional requirements will be identified as MRS or MRS only.

12.4.7.1 Description of Facility and Site (SL)

The applicant should provide a concise description of all site features affecting emergency response, including communications and assessment centers, assembly and relocation areas, and emergency equipment storage areas. The EP should identify any additional site features related to the safety of site operations. Most of this information will be presented in the SARs discussion of site characterization. However, supplemental information may be presented with the information on emergency planning.

The applicant may provide a detailed map of the site. An enlarged duplicate of the drawing suitable for use as a wall map may also be provided. The detailed map may be drawn to scale and show the following:

ISFSI storage areas or storage structures, and any holding areas for loaded transportation packages MRS storage areas or storage structures, pool, dry transfer facilities, intermodal transfer stations, and any holding areas for loaded transportation packages onsite structures and adjacent structures with descriptive labels (and building numbers, if applicable) other major site features, such as administrative and public access areas bar scale in both meters and feet compass indicating north onsite roads and parking lots onsite routes for transferring material to and from storage

12-14 site, controlled area, and restricted area boundaries, including locations of gates liquid retention tanks and ponds (include note if tanks or ponds are potentially contaminated roads, railroads, and navigable water in close proximity to the site rivers, lakes, streams, wetlands, or other ground water sources on site and adjacent to the site 12.4.7.2 Description of the Area Near the Site (SL)

The EP should describe the principal characteristics of the area near the site (out to approximately 1.6 kilometers (1 mile)) that may impact emergency planning, such as impediments to emergency response to the site (e.g., drawbridges, rivers), or facilities that may pose a threat to the site (e.g., chemical plants, petroleum gas terminals). The applicant should provide a general map (an approximately 16-kilometer (10-mile) radius) and a U.S. Geological Survey topographical map.

Although most of this information will be presented in the SARs discussion of site characterization, supplemental information may be presented with the information on emergency planning.

The EP should include a map of the area surrounding the site (out to approximately 1.6 kilometers (1 mile)) that provides the following information:

locations of population concentrations (such as towns, cities, office buildings, factories, arenas, stadiums, hospitals, nursing homes, and recreational areas) locations of facilities (such as schools, arenas, stadiums, nursing homes, hospitals, prisons) identification of primary routes for access of emergency equipment or for evacuation, as well as potential impediments to traffic flow (such as rivers, drawbridges, railroad grade crossings) locations of fire and police stations, hospitals, and other offsite emergency support organizations (specify whether offsite emergency support organizations received training to handle exposure to radioactive materials) 12.4.7.3 Types of Accidents (SL)

The EP should identify and describe each type of accident for which actions may be needed to prevent or minimize exposure from radiation, radioactive materials, or both, to onsite personnel for the ISFSI and MRS. The accidents should be described in terms of the process and physical location where they could occur, how the accidents could occur (e.g., equipment malfunction, instrument failure, human error), possible contributing or complicating factors, and possible onsite consequences. The accident descriptions should include any radiological material releases that could impact emergency response efforts. Chapter 16, Accident Analysis Evaluation, of this SRP describes the evaluation of this information.

12-15 12.4.7.4 Classification of Accidents (SL)

NUREG-1140, A Regulatory Analysis on Emergency Preparedness for Fuel Cycle and Other Radioactive Material Licensees, issued January 1988, describes incidents involving radioactive material for an ISFSI. NUREG-1092, Environmental Assessment for 10 CFR Part 72 Licensing Requirements for the Independent Storage of Spent Fuel and High-Level Radioactive Waste, issued August 1984, describes postulated accidents for surface cask storage of canistered fuel for an MRS.

Regulations for ISFSIs located away from a reactor site require only one level of emergency classification: Alert.

Regulations for ISFSIs and MRSs authorized to process and repackage SNF have two classes of accidents: Alert or Site Area Emergency (SAE).

The EP should include the emergency action levels at which an Alert or SAE will be declared.

12.4.7.4.1 Alerts An Alert is defined as an incident that has led to or could lead to a release of radioactive or other hazardous material to the environment, but the release is not expected to require a response by an offsite response organization to protect offsite individuals. The EP should identify events that could lead to initiation of an alert, such as the following:

severe natural phenomena (e.g., beyond-design-basis earthquake, hurricanes) or other incidents (e.g., fire, release of flammable gas) that have the potential to affect the confinement barrier indications of severe loss of control (e.g., radiation or contamination levels at the facility that are a factor of 100 over normal levels) a security compromise lasting more than 15 minutes accidental release of radioactivity due to failure of the confinement barrier other conditions that warrant precautionary activation of the licensees emergency response organization The plan should include a description of the applicants emergency response organization mobilization, steps taken to mitigate consequences of the emergency, and steps to be taken to escalate the classification, if necessary.

12.4.7.4.2 Site Area Emergency An SAE is defined as an incident that has led to or could lead to a significant release of radioactive or hazardous material and that could require a response by an offsite organization to protect offsite personnel. The EP should identify the events that could initiate an SAE, such as the following:

a compromise to systems or SSCs important to safety or a compromise to the integrity of SNF, HLW or reactor-related GTCC because of severe natural phenomena

12-16 (e.g., earthquake, flood, tsunamis) or severe incidents (e.g., aircraft crash into the facility, explosion, fire) imminent or actual loss of physical control of the facility rupture of the storage container confinement barrier and release of radioactivity outside of outer confinement barrier (e.g., loading facility building, SNF building)

The EP should include a description of the applicants emergency response organization mobilization, steps taken to mitigate consequences of the emergency, and procedures to notify offsite response organizations (fire, medical, police).

12.4.7.5 Detection of Accidents (SL)

The EP should describe the means of detecting each type of accident identified in the plan (e.g., visual observation, monitors, detectors, process alarms).

The EP should also describe the means to notify the operating staff of any abnormal operating condition or of any other danger to safe operation (e.g., a severe weather warning).

12.4.7.6 Mitigation of Consequences (SL)

For the events identified in Section 12.4.7.3 above, the EP should briefly describe the means and equipment provided for mitigating the consequences of each type of accident. The plan should include the mitigation of consequences to workers on site. Mitigating actions could include steps to reduce or stop any releases and steps to protect personnel and environment (e.g., evacuation, shelter, decontamination).

12.4.7.6.1 Limiting Actions The EP must describe the means and equipment provided for limiting the consequences of each type of accident identified in the plan (e.g.; fire detection and suppression systems, automatic shutoff of process or ventilation flow). The plan should address the actions and systems in place to reduce the magnitude or the effect of a radioactive or hazardous material release that has occurred (e.g., filtration or holdup systems, use of water sprays on airborne releases). The plan should include actions to be taken to limit and mitigate the consequences to public and workers.

Based upon the type of emergency, the plan should describe the criteria for the shutdown of systems or the facility and the steps to be taken to ensure a safe, orderly shutdown and the approximate time required for a safe shutdown.

12.4.7.6.2 Onsite Protective Actions The EP should describe the nature of onsite protective actions, criteria for implementing those actions, the areas involved, and the procedures to notify potentially affected persons. The plan should allow for the timely relocation of onsite personnel, the effective use of protective equipment and supplies, and the use of appropriate contamination control measures.

The EP should describe the means for controlling and/or minimizing radiological exposures for emergency response workers. The onsite exposure guidelines should be consistent with the Environmental Protection Agencys (EPAs) PAG Manual: Protective Action Guides and Planning Guidance for Radiological Incidents, issued January 2017, Section 3.1, Controlling Occupational

12-17 Exposure and Doses to Emergency Workers, to be used in actions to control fires, stop releases, or protect the facilities.

The plan should include methods for onsite non-essential personnel evacuation and accountability, such as the following:

criteria for ordering a site evacuation at the site area emergency classification means and timely notification of onsite persons impacted search and rescue locations of onsite and offsite assembly areas evacuation routes and means for transporting onsite personnel (e.g., privately owned vehicles, buses, company vehicles) monitoring of evacuees for contamination and control measures if contamination is found criteria for command center and assembly area evacuation and re-establishment at an alternate location means for evacuating and treating onsite injured personnel, including potentially contaminated personnel provisions for determining and maintaining accountability of assembled and evacuated personnel, and for identifying and determining the locations of personnel that were not evacuated The EP should describe provisions for preventing further spread of radioactive materials and for minimizing personnel exposures from radioactive materials.

The EP should describe provisions for determining the doses and dose commitments from external radiation exposure and internally deposited radioactive material received by emergency response personnel, including personnel from offsite emergency response organizations (fire, medical, police).

The EP should describe arrangements made for hospital and medical services and their capabilities to evaluate and treat contaminated, injured persons and injuries involving radiation, and radioactive materials. The medical facility description should include capabilities to control any contamination that may be associated with the physical injuries. The EP should specify how injured personnel who are potentially contaminated will be transported to offsite medical facilities.

The commitment to provide ambulance and hospital personnel with health physics support should be included.

Emergency Response Equipment and Facilities The EP should describe the onsite equipment and facilities designated for use during emergencies. The plan should describe an emergency facility from which evaluation and coordination of all licensee activities related to an emergency is to be carried out, and from which the licensee shall provide information to Federal, State and local authorities responding to

12-18 emergencies. The notification and coordination must be planned so that unavailability of some personnel, parts of the facility, and some equipment will not prevent the notification and coordination.

The plan should identify locations from which licensee emergency workers would be dispatched to perform radiation surveys, damage assessment, emergency repair, or other mitigating tasks.

The EP should describe the protective equipment and supplies available to emergency-response personnel. Types of equipment and supplies may include the following:

individual respiratory equipment protective clothing firefighting equipment and gear supplemental lighting medical supplies contamination control and decontamination equipment communications equipment radiation detection equipment (e.g., radiation meters, air samplers, dosimeters)

The EP should describe locations of emergency equipment and supplies, and the means for distributing these items. The plan should also include inventory lists indicating the emergency equipment and supplies provided at specified locations.

The EP should describe the primary and alternate onsite and offsite communication systems that would be used to transmit and receive information throughout the emergency. The plan should state the planned frequency of operational tests. A backup means of offsite communication to a commercial telephone should be provided for the notification of emergencies and requests for assistance. The frequency of operability checks should be stated.

12.4.7.7 Assessment of Releases (SL)

The EP should discuss the actions to be taken to determine the extent of the problem and to decide what corrective actions may be required for each class of emergency. This should include the types and methods of onsite and offsite sampling and monitoring in case of a release of radioactive or other hazardous material. The EP should describe the provisions for projection of offsite radiation exposures.

12.4.7.8 Responsibilities (SL)

The EP should describe the emergency organization to be activated on site for possible events and offsite for augmentation and support. The plan should delineate the authorities and responsibilities of key positions and groups and identify the communication chain for notifying and mobilizing personnel during normal and nonworking hours. Personnel with the responsibility for promptly notifying offsite response organizations and the NRC should be identified.

The EP should identify by position those with responsibility to declare an emergency and to initiate the appropriate response. The EP should include provisions for an annual review and audit of the emergency preparedness program to ensure that the program remains adequate. Elements of the audit should include a review of the following:

12-19 EP and associated procedures Radiological emergency response training activities (MRS only) records of emergency facilities, equipment, and supplies records associated with offsite response agencies interface (such as radiological emergency response training (MRS only) and letters of agreement) exercises, drills, communications, and inventory checks activation of the EP since the last audit Onsite Emergency Response Organization The EP should identify the onsite emergency response organization for the facility, including during periods such as holidays, weekends, and extended periods when normal operations are not being conducted. If the organization is activated in phases, the plan should describe the basic organization and each additional component that may be activated to augment the organization.

The plan should clearly state the minimum level of staffing needed to effectively implement the plan for each period or phase described.

Direction and Coordination The EP should designate the position of the person, and alternate(s), with the principal responsibility for implementing and directing the emergency response. This persons duties and authorities would include the following:

control of the situation escalation or termination of the emergency condition coordination of the staff and offsite personnel who augment the staff communication with parties requesting information regarding the event request of support from offsite response organizations The plan should also describe this persons authority to delegate responsibilities and the individuals who may be delegated certain emergency responsibilities.

Onsite Staff Emergency Assignments The EP should specify the organizational group or groups assigned to the functional areas of emergency activity listed below. The plan should also describe strategies for staffing these positions if the emergency lasts longer than one working shift. The duties, authorities, and interface with other groups and offsite assistance should be described. The organizational groups should provide support in the following areas:

facility systems operations personnel evacuation and accountability search and rescue operations first aid communications radiological survey and assessment (both on site and off site)

12-20 personnel and facility decontamination facility repair and damage control post event assessment recordkeeping media contact Emergency Response Records The EP should describe the assignment of responsibility for reporting and recording incidents of abnormal operation, equipment failure, and accidents that led to a facility emergency. The EP records to be maintained should include the following information:

cause of the incident personnel and equipment involved extent of injury and damage (on site and off site) as a result of the incident locations of contamination with the final decontamination survey results corrective actions taken to terminate the emergency actions taken or planned to prevent a recurrence of the incident onsite and offsite assistance requested and received any program changes as a resulting from a critique of emergency response activities The records associated with emergency planning that will be kept should also be described.

These should include the following:

training and retraining (including lesson plans and test questions) drills, exercises, and related critiques inventory and locations of emergency equipment and supplies maintenance, surveillance, calibration, and testing of emergency equipment and supplies letters of agreement with offsite support organizations reviews and updates of the EP notification of onsite personnel and offsite response organizations affected by an update of the plan or its implementing procedures Responsibilities at Site of Government Agencies The EP should identify the principal State agency and other government (local, county, State, and Federal) agencies or organizations with authority for emergency preparedness and response.

The plan should list the location and specific response capabilities, in terms of personnel and resources, of these agencies and organizations.

12-21 12.4.7.9 Notification and Coordination (SL)

The EP should describe the means used to activate the emergency response organization for each class of emergency during both regular and nonregular hours. The plan should describe the means provided to detect and notify the licensees operating staff of any abnormal operating conditions or of any danger to safe operations (e.g., a severe weather warning). The means to promptly notify offsite response organizations and the NRC should be described.

The EP should describe the ability to request offsite assistance, including medical assistance for the treatment of contaminated injured onsite workers. The plan should include the commitment to notify the NRC response center immediately after notification of local authorities but no later than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after an emergency is declared.

12.4.7.10 Information to be Communicated (SL)

The EP should describe the type of information to be communicated to offsite response organizations and the NRC. The types of information to be communicated should include the status of the facility, if a release of radioactive material is occurring or could occur, and recommendations for protective actions that may be implemented by the offsite response organization responsible for implementing protective actions. The recommended approach is to have estimated a range of potential source terms for each accident type in the planning, and then decide in the planning what recommendations would be made to offsite response organizations for each accident type. The plan should include a standard reporting checklist to facilitate timely notification for each postulated accident.

12.4.7.11 Training (SL)

The EP should include a description of the training provided to licensee staff on how to respond to an emergency. The plan should also include special instructions and orientations provided to offsite emergency response organizations. For an MRS only, the plan should describe emergency radiological response training provided for offsite response organizations that may be called to assist in an emergency onsite. The plan should include a description of training requirements for each position in the emergency organization, frequency of retraining, and training of onsite personnel who are not members of the emergency response staff.

12.4.7.12 Safe Condition (SL)

The EP should generally describe procedures for restoring the facility to a safe status after an accident and recovery plans. The plan should describe the position/title, authority and responsibilities of individuals who will fill key positions in the facility recovery organization. This organization shall include technical personnel with responsibilities to develop, evaluate and direct recovery and reentry operations. The plan should describe requirements for returning emergency equipment and supplies used during an accident to a state of readiness.

12.4.7.13 Exercises (SL)

The EP should describe the provisions for periodic drills and exercises. Communications checks with offsite agencies and radiological/health physics, medical, and fire drills should be performed at the interval established in 10 CFR 72.32(a) or 10 CFR 72.32(b).

12-22 The biennial onsite exercise required by 10 CFR 72.32, Emergency Plan, should test the effectiveness of the personnel, plan, procedures, and readiness of facilities, equipment, supplies, and instrumentation.

The applicant should invite offsite response organizations to participate in the periodic drills and exercises although recommended, their participation is not required. The EP should describe who has authority to develop the exercises, requirements for nonparticipating observers to evaluate the effectiveness of the exercise, the need for a critique of the exercise, and, if deficiencies are found, how they will be corrected.

12.4.7.14 Hazardous Chemicals (SL)

The EP must certify compliance with the Emergency Planning and Community Right-to-Know Act of 1986, with respect to any hazardous materials processed at the facility (10 CFR 72.32(a)(13) and 10 CFR 72.32(b)(13)).

12.4.7.15 Comments on the Emergency Plan (SL)

The EP should contain requirements for obtaining comments from offsite response organizations on the initial plan before submittal to the NRC with the license application. The licensee should communicate changes to the EP to the affected offsite response organizations. Letters of agreement with offsite response organizations should be reviewed annually and renewed on a periodic basis. Letters of agreement may be included in the EP or maintained separately.

12.4.7.16 Offsite Assistance (SL)

The EP should describe provisions and arrangements for assistance from offsite response organizations during and after an emergency. The licensee should clearly communicate exposure guidelines to offsite emergency response personnel. The plan should identify the services to be performed, means of communication and notification, and types of agreements that are in place for the following:

medical treatment facilities first aid personnel and ambulance service, as needed fire fighters law enforcement assistance The EP should describe the measures that will be taken to ensure that offsite response organizations maintain an awareness of their respective roles in an emergency and have the necessary equipment, supplies, and periodic radiological emergency response training (MRS only) to carry out their emergency response functions. The plan should describe any provisions to suspend security or safeguards measures for site access during an emergency.

The licensee should offer to meet at least annually with each offsite response organization to review items of mutual interest, including relevant changes to the EP. The licensee should discuss the emergency action level scheme, notification procedures, and overall response coordination process during these meetings.

12-23 12.4.7.17 Public Information (SL)

The EP should describe how information to the news media and the public would be disseminated during an emergency.

12.4.8 Physical Security and Safeguards Contingency Plans (SL)

The SAR must contain a physical protection plan as required by 10 CFR 72.180 and a safeguards contingency plan as required by 10 CFR 72.184, Safeguards contingency plan. Security plans are Safeguards Information and must describe how the applicant will comply with the applicable requirements in 10 CFR Part 73 and the requirements imposed by NRC orders for additional security measures. The EP should provide for the physical security of materials during transport to and from the ISFSI or MRS, as well as during the storage period. The plan must establish a security organization and include the following:

physical protection design features safeguard contingency plan guard training plan tests, inspections, audits, and other means to demonstrate compliance If the application is from the DOE, the SAR must include (1) a description of the physical security plan for protection against radiological sabotage (as required by Subpart H, Physical Protection, of 10 CFR Part 72), and (2) a certification that the plan will provide safeguards at the ISFSI or MRS that meet the requirements for comparable surface DOE facilities (required by 10 CFR 72.24(o)).

The safeguards contingency plan must comply with the format and content requirements of Appendix C, Licensee Safeguards Contingency Plans, to 10 CFR Part 73. An acceptable plan must contain (1) a predetermined set of decisions and actions to satisfy stated objectives; (2) an identification of the data, criteria, procedures, and mechanisms necessary to efficiently implement the decisions; and (3) a stipulation of the individual, group, or organizational entity responsible for each decision and action.

RG 5.55, Standard Format and Content of Safeguards Contingency Plans for Fuel Cycle Facilities, provides guidance on safeguards contingency plans that are specifically applicable to DSF facilities.

Review Procedures To begin the conduct of operations review, determine whether the applicant has submitted the respective elements described in RG 3.61, Standard Format and Content for a Topical Safety Analysis Report for a Spent Fuel Dry Storage Cask, RG 3.62, and RG 3.48. The review guidance provided in the following sections is predicated on the reviewer having access to the required products for the review and is based on lessons learned that should be applied in evaluating the submitted documentation.

Figure 12-1 Overview of Conduct of Operations evaluation shows the interrelationship between the conduct of operations evaluation and the other areas of review described in this SRP.

12-24 An applicants conduct of operation is, in a significant way, implemented by the applicants procedures. Therefore, the reviewer of this chapter should coordinate with the reviewer of the operating procedures (SRP Chapter 11, Operation Procedures and Systems Evaluation) to ensure that there are no inconsistencies.

12-25 Figure 12-2 Overview of Conduct of Operations Evaluation Chapter 12 - Conduct of Operations Evaluation Organizational Structure (SL)

Preoperational Testing and Startup Personnel Selection, Training,

  • Corporate Organization Operations (SL) and Certification (SL)
  • Onsite Organization
  • Preoperational Testing Plan
  • Personnel Organization
  • Identification of Agents and Contractors
  • Startup Plan
  • Selection and Training of
  • Management and Administrative Controls Operating Personnel
  • Selection and Training of Security Guards Acceptance Tests Normal Operations (SL)
  • Structural and Pressure Tests
  • Procedures Physical Security and Safeguards
  • Leak Tests
  • Records Contingency Plans (SL)
  • Shielding Tests
  • Inspection
  • Neutron Absorber Tests
  • Tests
  • Thermal Tests
  • Repair, Replacement, and Maintenance Chapter 4 -

Structural Evaluation

  • Description of the SSCs
  • Design Criteria
  • Loads
  • Normal and Off-normal Conditions
  • Accident Conditions Chapter 5 -

Thermal Evaluation

  • Thermal Loads and Environmental Conditions Chapter 7 - Criticality Evaluation
  • Criticality Design Criteria and Features
  • Model Specification
  • Criticality Analysis
  • Reactor-Related GTCC Waste (SL)

Chapter 8 - Materials Evaluation

  • Material Properties
  • Environmental Degradation; Chemical and Other Reactions
  • Code Use and Quality Standards Chapter 9 -

Confinement Evaluation

  • Confinement Design Characteristics
  • Confinement Monitoring Capability
  • Confinement Analyses Chapter 6 -

Shielding Evaluation

  • Shielding Design Description
  • Shielding Model Specification
  • Shielding Analyses
  • Reactor-Related GTCC Waste (SL)

Chapter 15 - Quality Assurance Evaluation

  • Organization and Program
  • Design and Nonconformance
  • Procedures and Drawings
  • Document Control
  • Procurement and Test Control
  • Inspection and Audits Chapter 3 -

Principal Design Criteria Evaluation

  • Design Bases for SSCs Important to Safety
  • Design Criteria for Safety Protection Systems Chapter 10A (SL) -

Radiation Protection Evaluation

  • Radiation Protection Design Features
  • Dose Assessment
  • Health Physics Program (SL)

Chapter 11 - Operation Procedures and Systems Evaluation

  • Analytical Sampling (SL)
  • Repair and Maintenance (SL)
  • Storage Container Loading
  • Storage Container Handling and Storage
  • Storage Container Unloading
  • Operation Support Systems (SL)
  • Other Operating Systems (SL)

Chapter 15 - Quality Assurance Evaluation

  • Organization and Program
  • Design and Nonconformance
  • Procedures and Drawings
  • Document Control
  • Procurement and Test Control
  • Inspection and Audits Chapter 14 -

Decommissioning Evaluation (SL)

  • Proposed Decommissioning Plan
  • Design Features
  • Operational Features
  • DSF or DSS Description and Operational Features
  • Engineering Drawings
  • Contents
  • Quality Assurance Program (SL)

Chapter 17 - Technical Specifications Evaluation

  • Functional and Operating Limits, Monitoring Instruments, and Limiting Control Settings
  • Limiting Conditions
  • Surveillance Requirements
  • Design Features
  • Administrative Controls

12-26 12.5.1 Organizational Structure (SL)

In addition to the guidance here, see also Chapter 10A of this SRP, specifically Sections 10A.5.4 and 10A.5.4.1, for additional guidance regarding the radiation protection and health physics aspects of the organizational structure and staffing.

Corporate Organization (SL)

Ensure that the relationship between the corporate organizations and the site organizations is clearly defined. Review the submitted documentation to gain an understanding of the delineation of authority and responsibilities regarding site activities. Ensure that the SAR specifies the frequency and scope of any audits or inspections conducted by the corporate organizations.

Onsite Organization (SL)

Review the material to gain a clear understanding of the distribution of responsibility to specific parts of the site organization and ensure that the site organization and the distribution of responsibilities for functions important to safety are clearly evident. Verify that the functions of radiation protection, nuclear criticality safety, and other safety entities are organizationally separate from the entity responsible for facility operations.

Determine whether the onsite organization includes a safety committee (or equivalent function) with appropriate representation and responsibilities. In making this determination, consider whether membership includes representatives from operating and safety support organizations.

Ensure that the safety committee has appropriate review and approval authority and procedures for the systematic review of proposed operations and changes. Confirm that the committee reports directly to the facility manager or other senior management.

In reviewing the proposed staffing levels and descriptions, consider the extent of expected operations. For example, in cases where the full spectrum of radioactive materials (e.g., full range of fuel types, HLW, or reactor-related GTCC waste) to be stored and the potential storage configurations and the kinds of handling operations is limited in scope or the applicants evaluations significantly bound the contents spectrum and storage configurations and configurations of handling operations, the level of onsite technical support (e.g., in areas such as nuclear criticality safety or structural design analysis) can be lower than in cases where the spectrum of contents and the potential storage configurations are not limited in scope.

Identification of Agents and Contractors (SL)

Verify that the SAR identifies the prime agents or contractors for the design, construction, and operation of the installation. Verify that the SAR identifies all principal consultants and outside service organizations, including those providing QA services. Confirm that the SAR clearly defines the division and assignments of responsibilities among those parties.

Management and Administrative Controls (SL)

Ensure that the applicant paid adequate attention to a proposed system of management and administrative controls. Verify that the SAR addresses each of the system elements identified in the acceptance criteria for management and administrative controls (Section 12.4.1.4 above).

Pay particular attention to the proposed system for procedures, including provisions for initial preparation, review, change, and approval.

12-27 12.5.2 Acceptance Tests The review procedures in this SRP chapter are focused on the testing of the storage containers that are loaded with the proposed radioactive materials contents. Additional tests may be needed for specific license applications for DSF SSCs or features that perform important functions (e.g., shielding, subcriticality, and confinement of radioactive materials, including wastes generated from DSF operations) at the site to ensure that the DSF design and operations meet regulatory requirements.

The review procedures described in this section are presented in a format intended to facilitate a single, independent review. Although one or more individual(s) may be tasked with preparing the corresponding section of the safety evaluation report (SER) related to the proposed acceptance tests, all review team members should examine the related information presented in the SAR.

Information in the SAR related to the acceptance tests may be located in the chapters related to specific disciplines (e.g., those related to the thermal evaluation) or in the chapter of the SAR on conduct of operations evaluation, or elsewhere. Devote special attention to those tests (or the lack of tests) that affect the respective functional area of review. If the descriptions included in the SAR are not sufficiently detailed to allow a complete evaluation concerning fulfillment of the acceptance criteria, request additional information from the applicant.

In general, applicants state that they will design, construct, and test the DSS or DSF under review to the codes and standards identified in the chapter of the SAR on principal design criteria. The NRC does not generally review detailed test procedures as part of the licensing process; however, the applicant is expected to describe (in the SAR) the essential elements of the proposed test programs. The staff may inspect selected portions of test procedures as part of its onsite activities.

The following subsections provide representative examples of acceptance tests that should be described in the SAR. Review the description of each test to ensure that the applicant has identified the purpose of the test, explained the proposed test method (including any applicable standard to which the test will be performed), defined the acceptance criteria and bases for the test, and described the actions to be taken if the acceptance criteria are not satisfied.

The following guidance is presented on the basis of tests the NRC deems acceptable. The guidance is based on operational experience and the knowledge from past reviews. Alternative tests and criteria may be used if the SAR provides appropriate explanation and adequate justification. Additional tests and criteria may be needed, depending on the operational experience and uniqueness of the proposed DSS or DSF design.

Structural and Pressure Tests Lifting trunnions should be fabricated and tested in accordance with ANSI N14.6, Radioactive MaterialsSpecial Lifting Devices for Shipping Containers Weighing 10,000 pounds (4,500 kilograms) or More. For a DSS or DSF where operations involve movement of SSCs into or out of a pool (e.g., SNF pool), site-specific details of the pool and lifting procedures may enable the DSS or DSF storage container to be considered a noncritical load, as defined by this standard.

Generally, however, the DSS or DSF storage container is considered a critical load during its handling in the pool. Consequently, trunnion testing should be performed at a minimum of 150 percent of the maximum service load if redundant lifting is employed or at a minimum of 300 percent of the service load if nonredundant lifting applies. These load tests should be performed to ensure that the trunnions and DSS or DSF storage container are conservatively

12-28 constructed and provide an adequate margin of safety when filled with the proposed radioactive material contents (e.g., SNF). Trunnion load testing should also be performed annually for the transfer cask, for DSS or DSF designs that use them, and at least 1 year before use for the storage container. Load testing of integral trunnions is not required once the loaded storage container has been placed on the pad. Ensure that the SAR chapter on technical specifications and operating controls and limits includes any restrictions on storage container lifting resulting from these tests. Ensure that the SAR explicitly states the testing values. Periodic NDE, in lieu of annual load tests, is acceptable for the trunnions provided that other conditions, as specified in ANSI N14.6, are also met.

The entire storage container confinement boundary should be pressure tested hydrostatically or pneumatically to 125 or 110 percent of the design pressure, respectively. The pressure test should be performed in accordance with the governing code associated with the confinement boundary, which typically has been ASME B&PV Code,Section III, Division 1, Subsection NB or NC for DSSs. The test pressure should be maintained for a minimum of 10 minutes, after which a visual inspection should be performed to detect any leakage. Ensure that the sections in the SAR describing the acceptance tests and maintenance programs clearly specify the hydrostatic and pneumatic test pressures. The helium leakage test, per ANSI N14.5, is not considered as a substitute for the ASME B&PV Code-required pressure test, and, conversely, the ASME B&PV Code-required pressure test is not a substitute for the helium leakage test.

Some storage containers (or DSS or DSF SSCs) include a neutron shielding material that may off-gas at higher temperatures. Such material is usually contained inside a thin steel shell to prevent loss of mass and provide protection from minor accidents and natural phenomenon events. Rupture disks or relief valves are generally provided to prevent catastrophic failure of this shell. The shell should be tested to 125 percent of the rupture disk burst pressure, which is usually equivalent to 125 percent of the shell design pressure. Verify that the SAR clearly specifies the burst pressure for the rupture disk, along with its coincident burst temperature and tolerance on burst pressure.

Some storage container designs use ferritic steels that are subject to brittle fracture failures at low temperature. ASME B&PV Code,Section II, Materials, Part A, Ferrous Materials Specifications, contains procedures for testing ferritic steel used in low-temperature applications.

NUREG/CR-1815, Recommendations for Protecting Against Failure by Brittle Fracture in Ferritic Steel Shipping Containers Up to Four Inches Thick, issued June 1981, provides staff guidance concerning materials and thickness ranges subject to brittle fracture testing. On the basis of guidance in NUREG/CR-1815, Section 5.1.1, the NRC has established two methods for identifying suitable materials:

1.

The nil-ductility transition temperature should be determined by either direct measurement (American Society for Testing and Materials (ASTM) E208, Standard Test Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels) or indirect measurement (ASTM E604, Standard Test Method for Dynamic Tear Testing of Metallic Materials), and the minimum operating temperature of the steel should be specified as 28 degrees Celsius (50 degrees Fahrenheit) higher than the nil-ductility transition.

2.

The NRC staff accepts ASME Charpy testing procedures for verification of the materials minimum absorbed energy. Acceptable energy absorption values and test temperatures of Charpy V-notch impact tests are listed in the ASME B&PV Code,Section II, SA-20, Specifications for General Requirements for Steel Plates for Pressure Vessels,

12-29 Table A1.15. Coordinate with the thermal reviewer (SRP Chapter 5, Thermal Evaluation) to ensure that the applicant selected the correct temperatures for the tests and that the SAR specifies the method of testing. For storage containers (or DSS or DSF SSCs) with ferritic steel walls thicker than 102 millimeters (4 inches), follow the guidance provided in NUREG/CR-3826, Recommendations for Protecting Against Failure by Brittle Fracture in Ferritic Steel Shipping Containers Greater than Four Inches Thick, issued July 1984.

Leak Tests Confirm that the applicant has described the leak tests to be performed on all confinement boundaries except as excluded in Section 8.5.3.3 of this SRP, which only applies to the closure welds typically made in the field. Leak testing should show that the inner closure weld of the storage container lid and primary welds of the vent and drain port cover plates meet the leakage limit. For all-welded confinement boundaries, the NRC has, with adequate justification, considered it acceptable for licensees and CoC holders to omit leak testing of the second (i.e., redundant) welds associated with the lid and its corresponding vent and drain port cover plates (see Figures 8-2 and 8-3 of this SRP). As shown in the figures, the redundant welds are not pressurized (or potentially pressurized because of closure valves, as described in Section 8.5.3.3.2) at the time of welding. A fabrication leak test should be performed on every storage container in the shop to ensure that the tested leakage rate meets the appropriate design leakage rate criteria (and regulatory criteria). Leak tests of the confinement boundary should be performed during the fabrication process such that subsequent fabrication procedures do not adversely affect the integrity of the confinement boundary.

Leakage criteria in units of Pascal cubic meter per second or that reference cubic centimeters per second should be at least as restrictive as those specified in the principal design criteria provided in the SAR. The SAR should also indicate the general testing methods (e.g., pressure increase, mass spectrometer) and required sensitivities. If storage container closure depends on more than one seal (e.g., lid, vent port, drain port), the leakage criteria should ensure that the total leakage is within the design requirements. Leak testing should be conducted in accordance with ANSI N14.5.

Visual and Nondestructive Examination Inspections Verify that the applicant will fabricate and examine storage container components in accordance with an accepted design standard such as ASME B&PV Code,Section III or VIII. These sections define the examination requirements mentioned in Section II;Section V, Nondestructive Examination; and Section IX, Welding and Brazing Qualification. The following guidance assumes that the ASME B&PV Code is applicable to the storage container being reviewed.

Confirm that the NDE of weldments is well characterized on drawings, using standard NDE symbols and notations (see American Welding Society (AWS) A2.4, Standard Symbols for Welding, Brazing, and Nondestructive Examination). Verify that each fabricator is required to establish and document a detailed, written weld inspection plan in accordance with an approved QA program that complies with Subpart G, Quality Assurance, of 10 CFR Part 72. Verify that the inspection plan includes visual, liquid (dye) penetrant (PT), magnetic particle (MT), ultrasonic (UT), and radiographic (RT) testing, as applicable. Confirm that the inspection plan identifies welds to be examined, the examination sequence, type of examination, and the appropriate acceptance criteria as defined by either the ASME B&PV Code or an alternative approach proposed and justified by the applicant. Inspection personnel should be qualified, in accordance with the current revision of the American Society for Nondestructive Testing (ASNT)

12-30 Recommended Practice No. SNT-TC-1A, Personnel Qualification and Certification in Nondestructive Testing, as specified by the ASME B&PV Code. All weld-related NDE should be performed in accordance with written and approved procedures. Fabrication controls and specifications should be in place and field tested to prevent postwelding operations (such as grinding) from compromising the design requirements (such as wall thickness).

Verify that confinement boundary nonclosure welds meet the requirements of ASME B&PV Code,Section III, Division 1, Subsections NB or NC, Article NB/NC-5200, Required Examination of Welds for Fabrication and Preservice Baseline. This section requires volumetric examination and either PT or MT for all Category A and most Category B or Category C welded joints in vessels, and longitudinal or full-penetration welded joints in other components. The ASME-approved specifications for RT, UT, PT, and MT are detailed in ASME B&PV Code,Section V, Articles 2, 4, 6, and 7, respectively.

Confirm that the acceptance standards for nondestructive testing are in accordance with ASME B&PV Code,Section III, Division 1, Subsection NB or NC-5300, Acceptance Standards. Testers should reject unacceptable imperfections (such as a crack, a zone of incomplete fusion or penetration, elongated indications with lengths greater than specified limits, and rounded indications in excess of the limits in ASME B&PV Code,Section III, Division 1, Appendix VI).

Repaired welds should be reexamined in accordance with the original examination method and associated acceptance criteria.

For confinement welds that cannot be volumetrically examined using RT, the licensee may use 100 percent UT. The ASME-approved UT specifications are detailed in ASME B&PV Code,Section V, Article 4. Ensure that acceptance criteria are defined in accordance with ASME B&PV Code,Section III, Division 1, Subsection NB or NC-5330, Ultrasonic Acceptance Standards.

Cracks, lack of fusion, or incomplete penetration are unacceptable, regardless of length.

The NRC has accepted multiple surface examinations of welds, combined with helium leak tests for inspecting the final redundant seal welded closures.

For storage container internals, confirm that the licensee will perform all NDE testing in accordance with ASME B&PV Code,Section III, Division 1, Subsection NG.

Verify that nonconfinement welds will meet the requirements of ASME B&PV Code,Section III, Subsection NF, or Section VIII, Division 1, as applicable. Welds on internal components (e.g., baskets) should meet the requirements of ASME B&PV Code,Section III, Subsection NG.

The required volumetric examination of welds is either RT or UT, as discussed in ASME B&PV Code,Section III, NF-5200, Required Examination of Welds, and Section VIII, UW-11. The appropriate specifications from ASME B&PV Code,Section V, are invoked in Article 2 for RT and in Article 5 for UT. Acceptance standards for RT are detailed in ASME B&PV Code,Section III, Subsection NF, NF-5320, Radiographic Acceptance Standards, and for UT in NF-5330, Ultrasonic Acceptance Standards. For Section VIII weldments, ensure that the RT acceptance criteria are in accordance with ASME B&PV Code,Section VIII, Division 1, UW-51, and the repair of unacceptable defects is in accordance with UW-38. Repaired welds should be reexamined in accordance with the original acceptance criteria.

Nonconfinement welds that cannot be examined using RT should undergo UT in accordance with ASME B&PV Code,Section V, Article 4. Ensure that acceptance criteria are in accordance with ASME B&PV Code,Section VIII, Division 1, UW-53 and Appendix 12, and the repair of unacceptable defects is in accordance with UW-38. Repaired welds should be reexamined in

12-31 accordance with the original examination methods and associated acceptance criteria. If applicable, the SAR should also justify the rationale for not requiring RT examination of these welds.

Verify that nonconfinement welds for storage container components that are designed and fabricated in accordance with ASME B&PV Code,Section III, that cannot be examined using RT or UT undergo PT or MT examination in accordance with ASME B&PV Code,Section V, Articles 6 and 7, respectively. Ensure that acceptance criteria are in accordance with Articles NF-5350, Liquid Penetrant Acceptance Standards, and NF-5340, Magnetic Particle Acceptance Standards, respectively. Repaired welds should be reexamined in accordance with the original acceptance criteria. If applicable, the SAR should also justify the rationale for not requiring volumetric inspection techniques (RT or UT) for these welds.

Nonconfinement welds may also be welded, repaired, and examined in accordance with AWS D1.1, Structural Welding CodeSteel; D1.3, Structural Welding CodeSheet Steel; and D1.6, Structural Welding CodeStainless Steel. Confirm that the design drawings call out the use of these standards.

Finished surfaces of the storage container should be visually examined in accordance with the ASME B&PV Code Section V, Article 9. For welds examined using visual testing, ensure that the acceptance criteria are in accordance with ASME B&PV Code,Section VIII, Division 1, UW-35 and UW-36, or NF-5360, Acceptance Standards for Visual Examination of Welds. Note that O-ring seating, such as for a bolted lid cask design, may have surface finish acceptance criteria defined by the O-ring manufacturer.

Verify that the acceptance tests include the use of PT to detect discontinuities (such as cracks, seams, laps, laminations, and porosity) that open to the surface of nonporous metals. PT should be performed in accordance with ASME B&PV Code,Section V, Article 6. Ensure also that acceptance criteria for PT examination of confinement welds are in accordance with ASME B&PV Code,Section III, Subsection NB/NC, Article NB/NC-5350. Ensure that repair procedures are in accordance with ASME B&PV Code,Section III, Article NB/NC-4450, Repair of Weld Metal Defects. Ensure that acceptance criteria for PT examination of nonconfinement welds are in accordance with ASME B&PV Code,Section VIII, Division 1, Appendix 8, or NF-5350. Ensure that repair procedures are in accordance with ASME B&PV Code,Section III or NF-2500, Examination and Repair of Material, and NF-4450, Repair of Weld Material Defects.

Shielding Tests The materials that comprise the DSS or DSF SSCs should sufficiently maintain their physical and mechanical properties during all conditions of operations. DSS or DSF gamma shielding materials (e.g., lead, steel, and concrete) should not experience cracks, pinholes, uncontrollable voids, slumping, or loss of shielding effectiveness to an extent that compromises safety. The shield should perform its intended function throughout the licensed or certified period of storage operations.

DSS or DSF materials used for neutron shielding should be designed to perform their safety function without significant degradation, gas release, or physical alteration for the full term of the licensed or certified period of storage operations. Tests are required to ensure these conditions are met.

12-32 Tests of the effectiveness of both the gamma and neutron shielding may be required if, for example, the DSS or DSF design includes materials such as poured lead for gamma shielding or a special (polymer-based) neutron absorbing material. In such instances, verify that the SAR describes any scanning or probing with an auxiliary source for the purpose of characterizing the shielding effectiveness. This shield testing should be done for every DSS or DSF SSC that uses these kinds of shielding materials to demonstrate proper fabrication in accordance with the design drawings. Even in instances were these shields may be installed in the DSS or DSF SSCs in prefabricated pieces, verify that the SAR includes SSC fabrication descriptions and tests to ensure fit-up of the prefabricated shielding materials with the SSCs. Such descriptions and tests should ensure that the prefabricated materials perform as designed, have the necessary dimensional and material properties, and that fit-up precludes unanalyzed streaming paths in the SSCs. For materials such as polymer-based neutron shields, tests may need to include qualifications testing of the fabrication process to ensure proper material specifications and uniformity of these specifications and material composition throughout the material.

Verify that shielding effectiveness tests include dose rate scans over the extent of the SSC surfaces where the shielding materials are present. Ensure that the tests use appropriate acceptance criteria that are based on the design specifications of the SSCs and shielding materials, including any dimensional and material tolerances. The criteria may be dose rates that are calculated using a computer code or are measured using a mock-up of the SSC, with either method using the same radiation source (properties), source-SSC-detector geometry, and the design specifications of the SSC (including material and dimensional tolerances). Any SSC dose rates that exceed the dose rate criteria indicate the SSC shielding is not acceptable. Any areas of an SSC that are affected by efforts to fix any shielding problems should be re-tested to the same criteria.

Alternatively, the applicant may propose an alternate testing program(s) with appropriate justification. For example, the applicant may use dose rate measurements of loaded DSS or DSF storage containers, in lieu of an auxiliary source, to verify shielding effectiveness with appropriate scanning of the shield and an appropriate testing program that uses the actual source strength, configuration, and other appropriate characteristics of the loaded contents for determining the acceptance criteria of the test.

Neutron Absorber Tests Neutron absorber materials require both qualification and acceptance testing to provide assurance that the control of criticality by absorbing thermal neutrons will be met in systems designed for nuclear fuel storage, transportation, or both. Both qualification and acceptance testing are generally described in ASTM C1671, Standard Practice for Qualification and Acceptance of Boron Based Metallic Neutron Absorbers for Nuclear Criticality Control for Dry Cask Storage Systems and Transportation Packaging, with exceptions, additions, and clarifications provided in Chapter 8, Materials Evaluation, of this SRP. Section 8.5.7, Criticality Control Materials, of this SRP provides detailed guidance on qualification testing.

Acceptance tests are used to ensure that material properties for plates and other shapes produced in a given production run are in compliance with the materials requirements of the application. In one sense, acceptance tests verify that the material of a given production run has yielded products that have been shown to be like the products that were used in the qualification testing. Acceptance tests are used to ensure that the production process is operating in a satisfactory manner and use statistical data for selected measurable parameters. For all boron-containing absorber materials, acceptance tests should (1) verify boron-10 content and

12-33 uniformity, (2) require visual examinations to establish that only acceptable levels of defects are present from cracks, porosity, blisters, or foreign inclusions, and (3) make dimensional determinations (e.g., plate thickness which is important to the areal density).

Neutron attenuation tests are calibrated using appropriate standards such as those based on (coated with) zirconium diboride plates to ensure the accuracy of the measured values. As described in Appendix 8A, Clarifications, Guidance, and Exceptions to ASTM Standard Practice C1671-15, to Chapter 8 of this SRP, approved substitutes may be used for the attenuation tests for material for which 75-percent credit is taken for boron content. These include tests such as chemical analysis, provided that (1) both the neutron attenuation tests and the alternative tests have at least the sensitivity of tests specified in ASTM C1671 and (2) the alternate form of testing is regularly benchmarked against calibrated neutron attenuation tests.

Chemical analyses should also include spectrochemical analysis for material impurity levels and boron-10 content. Uniformity is assessed using statistical sampling techniques that ensure that the entire plate of material and all plates in a lot meet a 95/95 criterion. This means that a test result has a 95-percent likelihood of containing the minimum required amount of boron-10, and that this is known at the 95-percent confidence level.

Confirm that the calculation of minimum poison content (e.g., poison areal density) conservatively accounts for tolerance limits on material density, poison concentration, and component dimensions. Thickness tolerances on rolled plates, sheets, or other shapes are typically on the order of +/-10 percent. The acceptance testing should provide a representative sampling of coupons for plates or sheets from a given lot. Statistical sampling can be used to the extent practical, using test locations on a coupon that will account for local variations or anomalies within the coupon and hence within the plates represented by the coupon. Confirm that the applicant has taken the adequate numbers of samples to ensure the confidence level required for the application.

Acceptance Testing of Fabricated Materials for 75-Percent Boron Credit For multiphase absorber materials analyzed with 75-percent poison credit (or less), confirm that acceptance testing is consistent with the following:

The effective boron-10 content should be verified from plate coupons by either (1) neutron attenuation testing or (2) chemical assay to determine boron content with mass spectrometric analysis for isotopic composition (see conditions in Appendix 8A to Chapter 8 of this SRP).

Sufficient coupons should be taken for acceptance testing to justify the level of credit given. Rejection of a coupon should result in rejection of the plate from which it is taken.

Sampling may be reduced to lesser percentages of the coupons taken (e.g., to 50 percent of all coupons) after acceptance of all coupons in the first 25 percent of the lot. A rejection during reduced inspection should invoke a 100-percent inspection for coupons from that lot.

A visual examination of all plates for defects should be conducted.

12-34 Acceptance Testing for Greater Than 75-Percent Boron Credit For acceptance testing of borated absorbers at levels of poison credit beyond 75 percent, the extent of the acceptance testing and inspection is enhanced. Some of the data helpful in meeting the guidance in ASTM C1671, Section 5.3.4, are as follows:

The effective boron-10 content is verified by neutron attenuation testing of coupons. An adequate number of coupons should be acceptance tested for each lot of materials to statistically demonstrate that the 95/95 criterion is satisfied for the minimum required boron-10 content. The minimum areal density is specified in the SAR.

Sufficient coupons should be taken to satisfy the 95/95 criterion. For example, coupons are taken from at least every other plate unless justification for fewer is given.

Measurements are made on samples taken from 100 percent of all coupons. Rejection of a coupon should result in rejection of the plate. Sampling may be reduced to 50 percent of all coupons after acceptance of all coupons in the first 25 percent of the lot. A rejection during reduced inspection should invoke a return to 100-percent inspection for that lot.

The applicant should perform a statistical analysis of the neutron attenuation results for all plates in a lot. This analysis should show that the lot meets the 95/95 criterion. That is, using a one-sided tolerance limit factor for a normal distribution with at least 95-percent probability, the areal density is greater than or equal to the specified minimum value with 95-percent confidence level. Failure to meet this acceptance criterion of this statistical analysis should result in rejection of the entire lot for use at 100-percent (90-percent credit in keff calculations). Applicants may choose to convert all areal densities determined by neutron attenuation to a volume density by dividing by the thickness of the coupon. The one-side tolerance limit of volume density with 95-percent probability and 95-percent confidence may then be determined. The minimum specified value of the areal density may be divided by the 95/95 lower tolerance limit of boron-10 volume density to arrive at the minimum plate thickness. Hence, all plates that have any locations thinner than this minimum should be rejected, and those equal to or thicker may be accepted.

A visual examination of all plates for defects should be conducted.

Refer to Section 8.5.7.2, Computation of Percent Credit for Boron-Based Neutron Absorbers, of this SRP regarding how to compute the level of credit.

Thermal Tests Depending on the details of the design and operational aspects of the DSS or DSF SSCs, testing may be required to verify adequate thermal performance. Adequate thermal performance would be established based on the thermal analysis results and applicable technical specifications (limiting conditions for operation and surveillance requirements). Confirm that the applicant has established acceptance criteria on the basis of the conditions of the test (e.g., test heat loading, ambient conditions, temperatures, pressures).

12-35 12.5.3 Preoperational Testing and Startup Operations (SL)

Review the preoperational testing plan to determine that it includes all of the necessary tests and provides for proper evaluation, approval, and use of the test results. Determine that the testing descriptions, responses expected, and contingent corrective actions are appropriate for the item being tested. In performing these assessments, seek the assistance of NRC staff with expertise in the specific topical areas covered by the tests.

In determining whether the preoperational testing plan is comprehensive, consider the inclusion of the following types of testing and evaluation, as applicable:

tests associated with construction (or reference to submitted construction specifications) preoperational testing specified in technical specifications calibration and testing of all equipment and instruments, monitors, and systems with a safety or security function tests of supplier-owned equipment to be used in functional operations (e.g., storage container haul trailer and positioning equipment) and in testing load tests of rigging, spreaders, and lift points evaluations of the effectiveness of procedures and consideration of potentially improved alternatives tests of physical and programmed limits on travel of lifting and transfer equipment (e.g., travel over a pool, lift heights, positioning force) 12.5.4 Maintenance Program In general, applicants should design, construct, and periodically test the DSS or DSF under review to the codes and standards identified in the principal design criteria chapter of the SAR. The NRC does not generally review detailed periodic test and maintenance procedures as part of the certification or licensing process; however, the applicant is expected to describe important or essential elements of the maintenance programs in the SAR.

The following subsections describe (some of) the maintenance program elements that are subject to NRC review. Review each program element for each maintenance program included in the SAR to ensure that the applicant has identified the purpose of the periodic test, explained the proposed test method (including any applicable standard to which the test will be performed),

defined the acceptance criteria and bases for the test, and described the actions to be taken if the acceptance criteria are not satisfied. Confirm that the SAR describes the accessibility of SSCs important to safety for inspection, maintenance, and testing, in accordance with 10 CFR 72.122(f) for a specific license or 10 CFR 72.236(g) for a CoC.

DSSs or DSF storage containers are typically designed as passive units requiring minimal maintenance. Ensure that the SAR addresses the areas described in the subsections below, as applicable.

12-36 Inspection Usually, the DSS or DSF has at least one monitoring system (e.g., pressure, temperature, dosimetry). Confirm that the SAR discusses how such systems will be used to provide information regarding possible off-normal events and what surveillance actions may be necessary to ensure that these systems function properly. The licensee at the site will develop and implement detailed procedures.

Confirm that the SAR describes routine, periodic visual surface and weld inspections, which should be limited to the readily accessible surfaces (e.g., the exterior surface of the DSS or DSF storage container and all surfaces of empty transfer casks). In addition, the SAR should discuss inspection of lifting and rotating trunnion load-bearing surfaces. The SAR should discuss any other appropriate inspections for other DSF SSCs.

Tests Verify that the SAR describes any periodic tests of DSS or DSF SSCs and features or calibration of monitoring instrumentation, as well as periodic tests to verify shielding, thermal, and confinement capabilities. Confirm that the applicant has otherwise justified that aging and degradation of materials related to the shielding, confinement, and thermal designs are not credible during the certified storage or licensed period of the DSS or DSF. Verify that the SAR also describes procedures for any applicable periodic testing of neutron poison effectiveness. As an alternative to the periodic testing of neutron poison effectiveness, the applicant may show continued poison effectiveness in the manner described in Chapter 7, Criticality Evaluation, of this SRP. The qualification tests of the poison material, discussed in SRP Chapter 8 may also be useful in showing the materials continued effectiveness.

In addition, verify that the SAR discusses any routine testing of support systems (e.g., vacuum drying, helium backfill, and leak testing equipment). Ensure that the SAR discusses any other appropriate tests for other DSF SSCs.

Repair, Replacement, and Maintenance Verify that the SAR discusses the repair and replacement of DSS or DSF SSCs and features, as may be required during the lifetime of the DSS or DSF. This discussion should include methods of repair or replacement, testing procedures, and acceptance criteria. Confirm that the SAR also describes procedures for routine maintenance (such as lubrication and reapplication of corrosion inhibiting materials in the event of scratches) through the expiration of the service life of the equipment. Such information is also often included in the chapter of the SAR on accident analysis, which describes actions to be taken following an off-normal event or accident condition.

Ensure that the SAR describes any other appropriate repair, replacement, and maintenance activities for other DSF SSCs.

12.5.5 Normal Operations (SL)

Procedures (SL)

Ensure that the SAR states that the applicant, as the licensee, will conduct all operations that are important to safety according to detailed written procedures and that these procedures will be based on and consistent with the operations, acceptance tests, and maintenance programs descriptions in the SAR. Determine whether the identified subjects for written procedures include

12-37 all routine and projected contingency operations and correlate with the descriptions of operations at the ISFSI or MRS.

Records (SL)

Determine whether the records identified for retention include all those required by regulations (refer to the listing and guidance in the acceptance criteria in Section 12.4.5.2 above).

12.5.6 Personnel Selection, Training, and Certification (SL)

Review proposed training for inclusion of regulatory requirements relating to personnel selection, training, certification, exercises, and training records. Determine acceptability based on satisfaction of regulatory requirements, guidance in RG 3.62 and RG 3.48, and evidence of experience in planning and conducting training programs.

Review the minimum qualifications for operating, technical, maintenance, and supervisory personnel and compare proposed requirements with those of other approved license applications. If there are no standard minimum qualifications, this evaluation will rely on the reviewers judgment. However, the minimum qualifications for these personnel generally include a bachelors degree and several years of experience in a related technical area that is commensurate with the level of assigned responsibility. Higher level managers typically have the same experience requirements plus previous supervisory or management experience.

Discussion of leak testing qualifications can be found in Information Notice 16-04, ANSI N14.5-2014 Revision and Leakage Rate Testing Considerations, dated March 28, 2016.

Ensure that the SAR adequately addresses any qualification requirements identified in evaluations throughout the SAR (e.g., qualifications for writing leak test procedures and performing leak tests identified in the confinement evaluation). Chapter 10A of this SRP, particularly Sections 10A.4.4, 10A.4.4.1, 10A.5.4, and 10A.5.4.1, includes added guidance regarding personnel selection, training and certification for radiation protection and health physics personnel, and radiation safety training for all licensee personnel. Coordinate with the Chapter 10A reviewer to ensure that the SAR adequately addresses this guidance.

Ensure that the SAR adequately addresses the implementation of the training program before the initiation of operations with SNF, HLW, or reactor-related GTCC waste, including a statement that most of the staff training and certification will be completed before receipt of the radioactive material to be stored.

Review the selection and qualification process for security personnel. Determine whether the process will ensure that security personnel will meet the requirements in 10 CFR 73.55, Requirements for Physical Protection of Licensed Activities in Nuclear Power Reactors Against Radiological Sabotage, and will be qualified to perform each assigned security job duty in accordance with Appendix B to 10 CFR Part 73 or the requirements imposed by NRC orders for additional security measures.

12-38 The following references provide additional guidance on training criteria and training program content:

ANSI/ANS 8.20, Nuclear Criticality Safety Training ANSI/ANS 3.1, Selection, Qualification, and Training of Personnel for Nuclear Power Plants ASNT Recommended Practice No. SNT-TC-1A ANSI/ASNT CP-189, American National Standard ASNT Standard for Qualification and Certification of Nondestructive Testing Personnel.

ASTM E1168, Standard Guide for Radiological Protection Training for Nuclear Facility Workers RG 1.8, Qualification and Training of Personnel for Nuclear Power Plants RG 1.134, Medical Assessment of Licensed Operators or Applicants for Operator Licenses at Nuclear Power Plants RG 8.27, Radiation Protection Training for Personnel at Light-Water-Cooled Nuclear Power Plants RG 8.29, Instruction Concerning Risks from Occupational Radiation Exposure NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Section 13.2.2, Non-Licensed Plant Staff Training 12.5.7 Emergency Planning (SL)

The NRC staff should review the license application SAR and other applicable documents because they contain information that may be relevant to the EP.

Description of Facility and Site (SL)

Review the description of the facility and the site to ensure that the applicant adequately described both the site and the adjacent area. Review the maps submitted as part of the EP to determine whether the site, ISFSI or MRS cask storage areas, other onsite structures, and major site features are sufficiently detailed.

Description of Area Near Site (SL)

Review the EP to determine whether the applicant has adequately described the principal characteristics of the area near the site. Review the maps provided to ensure that locations with emergency planning significance have been identified.

12-39 Types of Accidents (SL)

Review the EP to determine whether the applicant has adequately identified and described the types of radioactive material accidents. Based on submittals from other licensees and other available information, determine whether the EP addresses all postulated accidents.

Classification of Accidents (SL)

Review the emergency action levels at which an Alert or SAE will be declared. Review the procedures available to the NRC staff for classifying accidents.

Alert Review the EP to determine whether the definition of an Alert is consistent with the NRC's definition and whether initiating events are realistic and comprehensive. Review the mobilization efforts at the Alert level to determine whether the workers will be adequately protected.

Site Area Emergency Review the EP to determine whether the definition of SAE is consistent with the NRC's definition and whether initiating events are realistic and comprehensive. Review the procedure for facility mobilization if an SAE is declared. Review the steps taken to notify offsite response organizations that an SAE has been declared.

Detection of Accidents (SL)

Review the means used at the facility for detecting accidents. Review the location of radiation monitors, smoke or heat detectors, process alarms, and criticality alarms.

Mitigation of Consequences (SL)

Limiting Actions Review the processes and equipment available to mitigate the consequences of accidents identified in the EP. Review whether sprinkler systems, other fire suppression systems, fire detection systems, and filtration or holdup systems are identified.

Onsite Protective Actions Review the EP to determine whether it describes onsite protective actions to be taken, criteria for implementing the actions, and notification procedures for potentially affected personnel. Review exposure guidelines to determine whether the guidelines are consistent with the EPAs PAG Manual: Protective Action Guides and Planning Guidance for Radiological Incidents. Review the evacuation and relocation procedures to determine whether they are adequate. Review arrangements with offsite medical facilities to determine whether provisions to transport injured site personnel are adequate.

Emergency Response Equipment and Facilities Review the EP to ensure that emergency response equipment and facilities are adequately described. Ensure that the EP specifies the types of equipment necessary and the locations of the equipment. Review the provisions to inventory emergency response equipment.

12-40 Offsite Protective Actions Review the conditions that would require offsite protective actions. The recommended approach is to estimate a range of potential source terms for each accident type in the planning and then decide in the planning what recommendations would be made to offsite response organizations for each accident type.

Assessment of Releases (SL)

Review the EP to determine how the licensee will assess radioactive releases to the environment.

Review the description of the types of sampling and monitoring equipment to determine adequacy. This does not mean real-time assessment. It means measurements made after the release has occurred to determine how much material was released. The recommended approach is to estimate a range of potential source terms for each accident type in the planning and then decide in the planning what recommendations would be made to offsite response organizations for each accident type.

Responsibilities (SL)

Review the description of the onsite emergency organization to determine its adequacy to properly assess the situation. Review authorities and responsibilities of key positions and groups.

Normal Facility Operation Review the description of the normal operating facility organization. Verify that it identifies the positions with responsibility to declare an emergency and to initiate the appropriate response, as well as the personnel with the responsibility for maintaining the EP and implementing procedures.

Onsite Emergency Response Organization Review the onsite emergency response organization to determine whether there is sufficient staff to manage the emergency situation. Review the method of activating the emergency response organization. Determine whether the EP includes the minimum level of staffing.

Direction and Coordination Review the EP to determine whether it designates the position of the person, and his or her alternates, who has the principal responsibility for implementing and directing the emergency response. Determine whether the EP contains authorization for delegating responsibilities.

Onsite Emergency Assignments Ensure that the EP specifies which personnel and organizational groups are to provide support in the event of an emergency. Review the strategies for staffing the facility if the emergency is of long duration.

Emergency Response Records Review the EP to determine if it describes how records will be retained and the length of retention, as required in 10 CFR 72.80(c).

12-41 Notification and Coordination (SL)

Review the means used to activate the emergency response organization for each class of accident. Ensure that the EP describes how the communication with licensee personnel during both regular and nonregular hours is performed. Review the method the licensee has in place to notify local, State, and Federal authorities if an accident occurs.

Information to be Communicated (SL)

Review the EP and implementing procedures to determine whether the licensee has developed a clear, concise statement to be communicated to offsite response organizations and the NRC.

Review the standard reporting checklist to determine whether the licensee has notified all responsible agencies during an emergency.

Training (SL)

Review the emergency response training program to determine whether licensee personnel will be adequately trained.

Safe Condition (SL)

Review the EP for the general plans of restoring the facility to safe operation after an accident.

Review the requirements for ensuring that emergency response equipment is restored to a state of readiness.

Exercises (SL)

Review the provisions for conducting periodic drills and exercises.

Hazardous Chemicals (SL)

Ensure that the licensee has certified compliance with the Emergency Planning and Community Right-to-Know Act of 1986, with respect to any hazardous materials processed at the facility.

Comments of the Plan (SL)

Review the EPs requirements for obtaining comments from offsite response organizations.

Review any comments received from the offsite organizations and the resolution of the comments.

Offsite Assistance (SL)

Review provisions for requesting assistance from offsite response agencies during and after an emergency. Review radiological emergency response training provided to offsite emergency responders to an MRS only.

12.5.8 Physical Security and Safeguards Contingency Plans (SL)

Review the physical security plan against the applicable requirements in 10 CFR Part 73 and the applicable NRC orders for additional security measures and ensure that the plan adequately provides for each of the required elements. If the application is from the DOE, verify that it includes a description of the physical security plan for protection against radiological sabotage (as

12-42 required in 10 CFR Part 72, Subpart H) and a certification that it will provide safeguards at the ISFSI or MRS that meet the requirements for comparable surface DOE facilities.

Ensure that the safeguards contingency plan complies with the format and content requirements of Appendix C to 10 CFR Part 73, including (1) a predetermined set of decisions and actions to satisfy stated objectives, (2) an identification of the data, criteria, procedures, and mechanisms necessary to efficiently implement the decisions, and (3) a stipulation of the individual, group, or organizational entity responsible for each decision and action. Consult RG 5.55 for guidance on acceptable contents and format of safeguards contingency plans applicable to ISFSI or MRS installations. Although the applicant currently is not required to submit the written procedures that will implement the safeguards contingency plan (although the procedures are subject to NRC inspection on a periodic basis), review these procedures as needed to verify that the safeguards contingency plan meets the requirements of Appendix C to 10 CFR Part 73.

Evaluation Findings The NRC reviewer should prepare evaluation findings upon satisfaction of the regulatory requirements in Section 12.4 of this SRP. If the documentation submitted with the application fully supports positive findings for each of the regulatory requirements, the statements of findings should be similar to the following:

F12.1 (SL)

The SAR includes an acceptable description of the applicants organization to demonstrate the financial capabilities to construct, operate, and decommission the installation, as required by 10 CFR 72.22(e).

F12.2 (SL)

The SAR includes an acceptable description of the program covering preoperational testing and initial operations, in compliance with 10 CFR 72.24(p).

F12.3 (SL)

The SAR includes an adequate, acceptable description of the applicants operating organization, delegations of responsibility and authority, and the minimum skills and experience qualifications relevant to the various levels of responsibility and authority, in compliance with 10 CFR 72.28(c).

F12.4 (SL)

The SAR provides acceptable assurance with regard to the management, organization, and planning for preoperational testing and initial operations that the activities authorized by the license can be conducted without endangering the health and safety of the public, in compliance with 10 CFR 72.40(a)(13).

F12.5 SSCs important to safety will be designed, fabricated, erected, tested, and maintained to quality standards commensurate with the importance to safety of the function(s) they are intended to perform. Chapter _____

of the SAR identifies the safety importance of SSCs, and Chapter(s) ______ present(s) the applicable standards for their design, fabrication, and testing in accordance with 10 CFR 72.82(d),

10 CFR 72.122(a), 10 CFR 72.122(f), 10 CFR 72.124(b), 10 CFR 72.162, 10 CFR 72.234(b) and 10 CFR 72.236(b), (g), (j) and (l).

12-43 F12.6 The applicant or licensee, as appropriate, will examine and test, as needed, the [DSS or DSF designation] SSCs and features to ensure they do not exhibit any defects that could significantly reduce their confinement effectiveness. Chapter(s) ______ of the SAR describe(s) this inspection and testing, in compliance with 10 CFR [72.162/72.236(l)] or 10 CFR 72.122(a).

F12.7 (SL)

The SAR includes an acceptable plan for the conduct of operations, in compliance with 10 CFR 72.24(h), that provides reasonable assurance that operations important to safety will be performed in accordance with detailed written procedures, that the operating procedures are adequate in accordance with 10 CFR 72.40(a)(5), and that describes a records management system that will provide retention for all those required by regulation.

F12.8 (SL)

The applicant has provided acceptable technical qualifications, including training and experience, for personnel who will be engaged in the proposed activities, in compliance with 10 CFR 72.24(j) and 10 CFR 72.28(a).

F12.9 (SL)

The SAR includes an acceptable description of a personnel training program to comply with 10 CFR 72.24(j), 10 CFR 72.28(b),

10 CFR 72.40(a)(9), and 10 CFR Part 72, Subpart I.

F12.10 (SL) The SAR includes information that ensures that the applicant will have and maintain an adequate complement of trained and certified installation personnel before receipt of SNF, HLW, or reactor-related GTCC waste for storage, in compliance with 10 CFR 72.24(j) and 10 CFR 72.28(d).

F12.11 (SL) The SAR provides acceptable assurance that the applicant is qualified by reason of training and experience to conduct the operations covered by the regulations in compliance with 10 CFR 72.40(a)(4).

F12.12 (SL) The SAR includes an acceptable description of the emergency planning program, in compliance with 10 CFR 72.24(k), 10 CFR 72.32, and 10 CFR 72.40(a)(11).

F12.13 (SL) The SAR provides an acceptable description of the physical security and safeguards contingency plans, in compliance with 10 CFR 72.24(o),

10 CFR 72.40(a)(8), 10 CFR 72.40(a)(14), 10 CFR 72.180 and 10 CFR 72.184.

F12.14 (SL)

[If appropriate] The design of the DSF includes ____________ [specify the SSCs], the functional adequacy or reliability of which has not been demonstrated by previous use for the same purpose. The SAR describes acceptable planned tests and demonstration of capability in the areas of uncertainty before use, in compliance with 10 CFR 72.24(i).

12-44 The reviewer should provide a summary statement similar to the following:

The staff concludes that the conduct of operations program is [or for a DSS, in place of conduct of operations program is, can read as: acceptance tests and maintenance programs are] in compliance with 10 CFR Part 72 and that the applicable acceptance criteria have been satisfied. The evaluation of the conduct of operations program provides reasonable assurance that the [DSS or DSF] will allow for the safe storage of SNF and, as applicable for a DSF, reactor-related GTCC waste and HLW throughout its licensed or certified period of storage. This finding is reached on the basis of a review that considered the regulation itself, appropriate regulatory guides, applicable codes and standards, and accepted practices.

References 10 CFR Part 20, Standards for Protection Against Radiation.

10 CFR Part 72, Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater than Class C Waste.

10 CFR Part 73, Physical Protection of Plants and Materials.

American Concrete Institute (ACI) 318, Building Code Requirements for Structural Concrete and Commentary.

ACI 349, Code Requirements for Nuclear Safety-Related Concrete Structures and Commentary.

American Institute of Steel Construction 303-10, Code of Standard Practice for Steel Buildings and Bridges, included in the Steel Construction Manual.

American National Standards Institute (ANSI) N14.5, Radioactive MaterialsLeakage Tests on Packages for Shipment, 2014 ANSI N14.6, Radioactive MaterialsSpecial Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4500 Kilograms) or More.

ANSI/American Nuclear Society (ANS) 3.1-1993, Selection, Qualification, and Training of Personnel for Nuclear Power Plants, 1993.

ANSI/ANS 8.20, Nuclear Criticality Safety Training, 1991, reaffirmed 2015.

American Society of Mechanical Engineers, Boiler and Pressure Vessel (B&PV) Code, 2007 Addenda 2008.

Section II, Materials, Part A, Ferrous Materials Specifications, SA-20 Section III, Rules for Construction of Nuclear Facility Components Division 1, Metallic Components, Subsections NB, NC, NF, and NG Section V, Nondestructive Examination, Articles 2, 4, 5, 6, 7, 9 Section VIII, Rules for Construction of Pressure Vessels Division 1 Section IX, Welding and Brazing Qualifications American Society for Nondestructive Testing (ASNT) Recommended Practice No.SNT-TC-1A, Personnel Qualification and Certification in Nondestructive Testing.

12-45 ANSI/ASNT CP-189, American National Standard ASNT Standard for Qualification and Certification of Nondestructive Testing Personnel.

American Society for Testing and Materials (ASTM) C1671, Standard Practice for Qualification and Acceptance of Boron Based Metallic Neutron Absorbers for Nuclear Criticality Control for Dry Cask Storage Systems and Transportation Packaging.

ASTM E208, Standard Test Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels.

ASTM E604, Standard Test Method for Dynamic Tear Testing of Metallic Materials.

ASTM E1168, Standard Guide for Radiological Protection Training for Nuclear Facility Workers.

American Welding Society (AWS) A2.4, Standard Symbols for Welding, Brazing, and Nondestructive Examination.

AWS D1.1, Structural Welding CodeSteel.

AWS D1.3, Structural Welding CodeSheet Steel.

AWS D1.6, Structural Welding CodeStainless Steel.

American Society for Nondestructive Testing Recommended Practice No. SNT-TC-1A, Personnel Qualification and Certification in Nondestructive Testing.

Emergency Planning and Community Right-to-Know Act of 1986, 100 Stat. 1613, Public Law 99-499.

NRC, Information Notice 16-04, ANSI N14.5-2014 Revision and Leakage Rate Testing Considerations, March 28, 2016.

NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition.

NUREG-1140, A Regulatory Analysis on Emergency Preparedness for Fuel Cycle and Other Radioactive Material Licensees, January 1988.

NUREG/CR-1815, Recommendations for Protecting Against Failure by Brittle Fracture in Ferritic Steel Shipping Containers Up to Four Inches Thick, Lawrence Livermore National Laboratory, June 1981.

NUREG/CR-3826, Recommendations for Protecting Against Failure by Brittle Fracture in Ferritic Steel Shipping Containers Greater than Four Inches Thick, Lawrence Livermore National Laboratory, July 1984.

Regulatory Guide 1.8, Qualification and Training of Personnel for Nuclear Power Plants.

Regulatory Guide 1.134, Medical Assessment of Licensed Operators or Applicants for Operator Licenses at Nuclear Power Plants.

Regulatory Guide 3.48, "Standard Format and Content for the Safety Analysis Report for an Independent Spent Fuel Storage Installation or Monitored Retrievable Storage Installation (Dry Storage)."

12-46 Regulatory Guide 3.61, Standard Format and Content for a Topical Safety Analysis Report for a Spent Fuel Dry Storage Cask.

Regulatory Guide 3.62, Standard Format and Content for the Safety Analysis Report for Onsite Storage of Spent Fuel Storage Casks.

Regulatory Guide 5.20, Training, Equipping, and Qualifying of Guards and Watchmen.

Regulatory Guide 5.55, Standard Format and Content of Safeguards Contingency Plans for Fuel Cycle Facilities.

Regulatory Guide 8.27, Radiation Protection Training for Personnel at Light-Water-Cooled Nuclear Power Plants.

Regulatory Guide 8.29, Instruction Concerning Risks from Occupational Radiation Exposure.

U.S. Environmental Protection Agency (EPA) EPA-400/R-17/001, PAG Manual: Protective Action.

Retrieved from "https://kanterella.com/w/index.php?title=ML20321A107&oldid=3829674"