**NRC INSPECTION MANUAL** RNRP

INSPECTION PROCEDURE 69013

RESEARCH AND TEST REACTOR DECOMMISSIONING

PROGRAM APPLICABILITY: 2545

69013-01 INSPECTION OBJECTIVE

To determine if dismantlement and decontamination activities at a
research reactor are conducted safely and in accordance with regulatory
requirements, licensee commitments, and the NRC approved Decommissioning
Plan (DP).

69013‑02 INSPECTION REQUIREMENTS

02.01 [Organization and Staffing]{.underline}

a.  Determine if the licensee\'s organization satisfies the
    Decommissioning Plan requirements, license and technical
    specification requirements, and licensee commitments.

b.  Determine whether temporary staffing satisfies Decommissioning Plan
    requirements, licensee commitments, and provides the expertise and
    skills needed to ensure safe completion of decommissioning.

c.  Evaluate whether the licensee\'s decommissioning organization and
    staffing is adjusted to account for changes as decommissioning
    progresses.

02.02 [Work Controls]{.underline}

a.  Determine if the controls, policies, procedures, and work
    requirements are implemented as specified in the DP and the
    technical specifications (TS).

b.  Determine whether oversight and control of contracted work are
    adequate.

c.  Determine if engineering safety evaluations are completed prior to
    dismantlement or demolition of load-bearing structures.

02.03 [Health Physics]{.underline}

a.  Determine whether the licensee\'s radiation protection and As Low As
    Reasonably Achievable (ALARA) programs are documented as required by
    10 CFR 20.1101 and the DP.

b.  Determine if the required radiation surveys, sampling, and
    monitoring are performed and recorded in accordance with regulatory
    requirements and licensee procedures.

c.  Determine whether an annual review of the radiation protection
    program is performed as required by 10 CFR 20.1101(c). Verify that
    the Radiation Safety Officer (or equivalent) reviews and approves
    radiation protection program changes, radiological event
    investigations, and corrective actions in accordance with the DP,
    the TS, and licensee procedural requirements.

d.  Determine whether required calibrations of radiological survey,
    sampling, or monitoring instruments are commensurate with the DP and
    the licensee\'s procedures.

e.  Determine whether the personnel dosimetry program is conducted in
    accordance with 10 CFR 20 Subpart F and occupational doses are
    limited as specified in 10 CFR 20 Subpart C.

f.  Determine whether restricted areas and radioactive materials are
    labeled in accordance with 10 CFR 20 Subpart J.

g.  Determine whether access to radioactive materials and contaminated
    areas is controlled in accordance with 10 CFR 20 Subpart I.

h.  Determine whether use of protective clothing and exit frisking
    practices are in accordance with the licensee\'s procedural
    requirements.

i.  Determine whether personnel working around radioactive materials are
    instructed in radiation safety as required by the licensee\'s
    procedures and by 10 CFR 19.12.

j.  Determine whether the principles of ALARA are implemented by the
    licensee.

k.  Determine whether the licensee\'s control of internal exposures
    satisfies the requirements of 10 CFR 20 Subpart H.

l.  Where applicable, determine whether the dose to the embryo or fetus
    of declared pregnant women meets the requirements of 10 CFR 20.1208

02.04 [Effluent Control and Environmental Monitoring]{.underline}

a.  Determine whether releases of radioactive materials are controlled,
    monitored, and recorded as required by the DP and TS.

b.  Determine whether radioactive liquid, gaseous, and particulate
    effluent releases are controlled based on the environmental ALARA
    Program and 10 CFR 20 Subpart D.

c.  Determine whether environmental samples are collected, analyzed, and
    recorded as required.

02.05 [Change Control]{.underline}

a.  Determine if changes to the DP are made in accordance with
    10 CFR 50.59, and the policies and procedures described in the DP.

02.06 [Audits and Reviews]{.underline}

a.  Determine whether the decommissioning safety committee meets the
    administrative requirements (membership, qualifications, quorums,
    meeting frequencies, etc.) of the DP and TS.

b.  Determine whether the decommissioning safety committee provides the
    audits, reviews, and oversight as required in the DP or TS.

c.  Review management evaluations of events, abnormal conditions,
    reportable occurrences, review and audit findings, employee safety
    concerns, and other issues and problems that arise. Determine
    whether management response is timely and appropriate.

02.07 [Emergency Planning]{.underline}

a.  Determine whether the policies and procedures for responding to
    emergencies as required by the DP and Emergency Plan are current.

b.  Determine whether on-site and off-site emergency response personnel
    can acceptably respond to emergency conditions.

02.08 [Solid Radioactive Waste Management and
Transportation]{.underline}

a.  Determine if arrangements with burial sites, waste brokers,
    shippers, and processors will allow disposal of wastes identified
    during site characterization with particular attention to disposal
    of mixed waste.

b.  Determine if staging and temporary storage areas for waste are
    controlled as required by 10 CFR 20.1801. Determine whether the
    packages and the storage area are properly labeled as required by 10
    CFR 20 Subpart J.

c.  Determine if individuals who package and ship waste are trained as
    required by 49 CFR 172 Subpart H. Determine if additional training
    required for users of Type B packages and spent fuel casks is
    conducted in accordance with 10 CFR 71.105(d). Verify that
    replacement personnel are properly trained.

d.  Determine if waste is packaged for shipment in accordance with 49
    CFR 173 Subpart I. Determine if the licensee=s quality assurance
    program for the use of Type B packages and spent fuel shipping casks
    has been approved by the NRC in accordance with 10 CFR 71 Subpart H
    and is properly implemented.

e.  Determine the adequacy of the licensee=s program for ascertaining
    that the waste class and form are suitable for burial as required by
    10 CFR 61.55 and 61.56. Verify that compliance with these
    restrictions is assured in accordance with 10 CFR 20 Appendix G
    III.A.3.

f.  Determine if shipping manifests and records are prepared and
    maintained in accordance with 10 CFR 20.2006 and 10 CFR 20 Appendix
    G

g.  Determine whether written procedures and policies are available for
    classification, characterization, processing, packaging, delivery,
    and transportation of waste. Verify that these are approved by
    management and/or the safety committee as specified in the DP. Also
    verify that they are adequate, current, and implemented in
    conformance with 10 CFR Parts 20, 61,and 71 and 49 CFR Parts
    173-178.

69013‑03 INSPECTION GUIDANCE

[General Guidance]{.underline}

Safety concerns during the decontamination and dismantlement of a
research reactor are significantly different from the concerns during
operations. Radiological hazards may dominate the early phases of work,
when the inventory of radioactive material is large, but lessen as
decontamination and dismantlement progresses. The work environment will
be similar to a prolonged maintenance outage.

While varying levels of radiological hazards are usually present,
industrial hazards may become more safety significant during certain
work. Industrial safety hazards will depend on the type of work in
progress. Examples of these hazards include: elevated work from
scaffolding or ladders, cutting, burning, grinding, soil excavation and
trenching, asbestos removal, electrical and fire loading hazards. The
licensee staff may include industrial hygienists and safety specialists
who ensure that an effective industrial safety program is implemented
and personnel protective equipment (e.g. hard hats, safety glasses,
hearing protection, safety shoes) is available and properly used. In
addition, careful attention to housekeeping and fire prevention is
essential.

The equipment needed for initial response to industrial accidents is
found in Occupational Safety and Health Administration (OSHA)
requirements in 29 CFR 1910 and 1926, or the licensee=s industrial
safety manual. This may include eye wash stations, first aid kits, fire
extinguishers, evacuation alarms, and rescue extraction equipment for
work in confined spaces.

The NRC inspector is not an OSHA inspector but should review the
NRC-OSHA Memorandum of Understanding and be alert for weaknesses in the
licensee=s industrial safety program. Significant observations in this
area should be reported to licensee management and the region-based
NRC/OSHA Liaison Officer in accordance with IP 93001, AOSHA Interface
Activities.@ Furthermore, industrial hazards can be expected to change
as work progresses.

Routine NRC inspections should be frequent during the early phases of
the project. Initial inspections should verify that the decommissioning
organization is in place and all key positions filled prior to the start
of dismantlement and decontamination. The safety committee providing
oversight and review of changes must be fully operational. General
policies and procedures, as well as detailed work instructions, where
appropriate, should be available. Arrangements for contractor support
(tradesmen, supervision, engineering, HP, etc), specialized equipment,
and waste disposal should be completed. The inspector should maintain
frequent contact with the licensee and identify the schedule of safety
significant work to be inspected. As-needed inspections should be
conducted to observe preparation and performance of this work. A
flexible inspection schedule will be needed for the duration of
decommissioning.

[Specific Guidance]{.underline}

03.01 [Organization and Staffing]{.underline}. This inspection element
should be performance-based on the licensee\'s effectiveness at
fulfilling commitments and meeting regulatory requirements.
Performance-based elements characterizing adequate staffing could
include the licensee\'s success at: (1) timely review, implementation,
or revision of procedures, drawings, and design changes; (2) managing
engineering backlogs; (3) supervising decommissioning activities; (4)
performing facility tours to assess housekeeping and material condition;
and, (5) conducting technical evaluations and insightful management
reviews.

The inspector could also assess whether the licensee had performed a
task analysis of their staff by functional description, job comparison,
and qualification considerations. A possible licensee analysis could
balance an estimation of future site activities with staff experiences,
expertise, or work load. The licensee may or may not have documented
these management decisions.

The inspector could also assess whether the site staff can effectively
implement site procedures for accidents, events, fire emergencies, etc.
This includes the inter-actions between site personnel performing
decommissioning activities, security, and radiation protection
functions.

a.  [Organization]{.underline} The DP provides specific requirements for
    the decommissioning organization and staffing. The facility license
    and technical specifications may also provide additional
    requirements. The inspector should verify that licensee personnel
    and contractors satisfy the qualifications, responsibilities, and
    authorities requirements. The inspector should interview supervisory
    personnel to verify that they posses a clear understanding of their
    role in the organization. Qualifications and experience of the
    radiation safety officer should be adequate to deal with the
    anticipated radiological hazards. The composition of the safety
    committee and qualifications of the members should enable the
    committee to provide effective oversight of dismantlement work.
    Efforts by licensee management to retain operating staff with
    knowledge of reactor system layout and historical events with
    potential to impact decontamination and dismantlement should be
    reviewed. The licensee could consider the implementation of
    promotion, retention, or rotation programs to provide assurance that
    the experienced staff is retained.

b.  [Temporary Staffing]{.underline}. Contracted personnel may be hired
    to perform tasks beyond the capabilities of the reactor staff and
    management. This may include labor, supervision, rental equipment
    and equipment operators. Examples of this are health physics
    technicians, radioactive waste brokers, quality assurance
    inspectors, large crane operators, welders, scabble machine
    operators, and concrete cutters. Contractors who occupy positions
    described in the DP should be interviewed to determined if
    requirements for training and experience are met. The licensee
    retains prime responsibility for regulatory compliance by contracted
    workers and should provide the quality assurance and oversight of
    contractors that can be verified by the inspector.

c.  The licensee\'s organization may undergo changes as the facility
    transitions through decommissioning. For example, health physics
    technicians supporting remediation work will be replaced by
    personnel qualified to perform the final status survey. The number
    and qualifications of the licensee staff should be adjusted to meet
    regulatory requirements and maintain safety commensurate with the
    conduct of decommissioning activities.

03.02 [Work Control]{.underline}

a.  The DP may require detailed work instructions if the contamination
    found during site characterization requires complex decontamination
    and dismantlement. The selected methods, techniques, surveillance,
    maintenance and equipment to be used for decontamination and
    dismantlement will be described in the DP and the TS. The inspector
    should verify that these are implemented as required. Except for
    work that is clearly within the skills of a trade, the inspector
    should verify the availability and use of detailed work
    instructions. The instructions may include reference to a job
    specific radiation work permit and other safety precautions or
    permits. For example, scaffolding should be inspected for proper
    installation before use, a flame permit and fire watch should be
    posted before flame cutting or arc gouging.

> The DP may require work sequencing to ensure that required safety
> systems remain available for as long as required. For example,
> effluent monitors will be required until near the end of the project.
> The work schedule should also be used to avoid conflicts or work
> interferences. For example, a section of floor is removed that was
> needed to support a forklift designated to remove a large piece of
> equipment. The inspector should review management coordination and use
> of work schedules .

b.  Licensee programs to provide oversight and control of the
    contractors should ensure that work performance and coordination are
    in accordance with regulatory requirements.

c.  If major demolition or dismantlement of load bearing structures or
    soil is required to gain access to radioactive material, an
    engineering safety evaluation may be required. This should be
    completed by a competent and licensed professional engineer in
    accordance with local building codes and generally accepted
    practices. The evaluation may determine the need for bracing,
    shoring, or other temporary support structures. The inspector should
    determine if safety evaluations are completed and the recommended
    precautions are implemented.

03.03 [Health Physics]{.underline}

a.  The radiation protection and ALARA programs should be based on an
    assessment of the projected hazards during decommissioning and
    describe the methods to be used to limit personnel exposures.
    Program documentation and annual review requirements are specified
    in 10 CFR 20.1101 and program records in 10 CFR 20.2102.

> The licensee may choose to use selected radiation protection policies
> and procedures in effect during the reactor operations due to their
> familiarity to the staff. Enhancements to the program in selected
> areas should be reviewed. For example, use of full protective
> clothing, alarming personnel dosimeters, respirators, whole body
> counting, and planned special exposures are not routinely used during
> research reactor operations but may need to be added to the program
> during decommissioning.
>
> The DP should include an estimate of the total person-REM for the
> project. Licensee progress towards this goal should be reviewed to
> assess the effectiveness of ALARA efforts. Management support of and
> commitment to the ALARA program may be verified through a review of
> radiological planning and observation of work practices.

b.  Requirements for radiation surveying, sampling and monitoring can be
    found in Subpart F of 10 CFR Part 20 and licensee procedures. Record
    keeping requirements can be found in 10 CFR 20.2103. Radiation
    protection personnel should be observed performing radiological
    surveys of high radiation areas, contaminated areas, and during
    on-going work. Survey techniques and equipment should conform to
    generally accepted practices. A review of records of radiation
    surveys, sampling, and monitoring activities will provide an
    indication that radiological conditions are adequately assessed.
    Independent radiological surveys and split sampling may also be
    performed by the inspector to confirm the accuracy of licensee data.

> Radiological conditions will change significantly as highly activated
> components are removed, packaged, stored, and shipped off site.
> Frequent surveys should be completed during this work and area
> postings adjusted as necessary.
>
> During the latter phases of the project, technicians will conduct
> surveys of residual contamination in support of license termination.
> These surveys will generally use more sensitive techniques and
> equipment than were used during decontamination and dismantlement due
> to the lower levels of radioactivity. The inspector should review the
> equipment, survey techniques, and data reduction for conformance with
> licensee procedures and good practices. A thorough review of the final
> status surveys is performed using inspection procedure IP 69014.

c.  An annual review of the radiation protection programs is required by
    10 CFR 20.1101(c). A record of this review is required by 10 CFR
    20.2102. The inspector should review the audit results and
    corrective actions with the RSO. All program areas may be audited
    annually, or if decommissioning extends over several years, selected
    areas are audited such that all aspects are reviewed after two or
    three years. Experience with past decommissioning indicate that
    unanticipated radiological situations will occur in spite of the
    care exercised during the radiological characterization stage.
    Therefore, reviews of program implementation and identification of
    needed improvements may be more frequent than specified in 10 CFR
    20.1101(c).

> Radiological incidents should be recorded in accordance with licensee
> procedures. Corrective actions related to these incidents should be
> approved by the radiation safety officer (RSO) and the safety
> committee as specified in the DP.

d.  A review of about 50 percent of the calibration records for portable
    instruments is an acceptable sample for this inspection requirement.
    Instruments should be calibrated at the specified intervals using
    the manufacturers recommended techniques. The calibration of fixed
    area monitors, laboratory analytical equipment, airborne effluent,
    and liquid effluent monitors, if required by the DP, should be
    verified as current. An operational check of portable equipment by
    the technician prior to its use is an accepted good practice.

e.  Verify that facility personnel have been issued dosimetry and are
    wearing it as required. Multiple whole body dosimeters, finger
    rings, and wrist badges may be appropriate for certain work. Face
    shields or goggles may be used to protect the lens of the eye from
    beta radiation. If neutron sources or reactor fuel is handled,
    neutron dosimeters should be issued. The dosimetry processor must be
    National Voluntary Laboratory Accreditation Program (NVLAP)
    accredited in accordance with 10 CFR 20.1501. Verify that the
    licensee=s methods for determining internal dose are in conformance
    with 10 CFR 20.1204. Review records for all workers and verify that
    internal and external doses are properly summed as required by 10
    CFR 20.1202 and the TEDE dose is within limits specified in 10 CFR
    20.1201.

f.  Commensurate with item b. above, verify that area postings are kept
    up-to-date as decontamination and dismantlement progresses and
    radiological conditions change. Temporary storage of waste may
    create transitory radiation areas. Some areas will be de-posted
    after decontamination and remediation are completed. Verify that
    materials removed from the controlled area and containers of waste
    are labeled as specified in 10 CFR 20.1904. Posting of radiation,
    high radiation, very high radiation, airborne radioactivity, and
    radioactive material storage areas is specified in 10 CFR 20.1902.

g.  Verify that only personnel authorized by the licensee will have
    access to the radioactive materials area. Installation of fencing,
    gates, and locks may be required, depending on the site and need for
    staging or storage areas. For large sites, the NRC may authorize the
    unrestricted release of a portion of the site if the conditions
    specified in 10 CFR 20.1402 are satisfied. If reactor fuel and
    special nuclear material are still on-site, refer to the NRC
    approved security plan for detailed requirements.

h.  The licensee=s radioactive material control program is to ensure
    that licensed radioactive material is controlled in accordance with
    the requirements of 10 CFR Part 20. 10 CFR Part 20 does not contain
    release limits for the release of contaminated material to
    unrestricted areas; thus, the licensee=s criteria should be that no
    detectable licensed radioactive material (radioactive gaseous and
    liquid effluents excepted) is released for unrestricted use or as
    waste into an unrestricted area. The licensee=s equipment should
    ensure the radiation detection sensitivities are consistent with the
    NRC guidance contained in IE Circular 81-07 and IE Information
    Notice 85-92 for surface contamination and HPPOS-221 for
    volumetrically contaminated material. If applicable, radiation
    surveys for hard-to-detect radio nuclides that decay by electron
    capture should be as discussed in HPPOS-250. The licensee=s
    radiation detection instrumentation should be used at its typical
    sensitivity level based on appropriate counting parameters (i.e.,
    counting times and background radiation levels). The licensee should
    not establish a Arelease limit@ by altering the instrument=s typical
    sensitivity through such methods as raising the discriminator level
    or locating the instrument in a high radiation background area.

> Verify that practices to prevent the spread of contamination are
> effective. This may include painting or coating surfaces to fix any
> loose contamination. If protective clothing is used, workers should be
> trained in its use. Contaminated components should be wrapped before
> removal from the work area. The use of barriers, signs, and step off
> pads may be employed. Exit frisking of personnel and equipment should
> be in a low background area using sensitive monitors.

i.  The inspector should verify that the general employee training (GET)
    remains an effective tool in assuring personnel and facility safety.
    This training, at a minimum, should incorporate the material in 10
    CFR 19.12 through 19.16, and Part 20, facility security
    requirements, and emergency preparedness actions, if any. GET should
    be updated as required to reflect the current state of
    decommissioning and site organization. GET should include the
    information in RG 8.29 with periodic refresher training. An
    effective training program could also include, in part, safety
    evaluations, modifications, radiation protection, effluent controls,
    and design changes. Observe a pre-job safety briefing to verify that
    radiological hazards and protective measures are described to
    workers in detail. Full time job coverage by HP technicians may be
    appropriate for work crews with little prior nuclear experience.
    Note if briefings also include a discussion of industrial hazards
    and precautions.

j.  Select a job with a potential for significant dose to the workers.
    Verify that ALARA techniques are effectively employed such as mockup
    training, special tooling, or use of robots. Assess the commitment
    to ALARA through discussions with workers and supervision. Review
    the percentage of work completion relative to dose projections as
    described in the DP and discuss any anomalies with the licensee.

k.  Many aggressive decontamination techniques, such as scabbling and
    grinding, will generate airborne radioactivity. Verify that
    engineering controls, such as vacuum systems, local high‑efficiency
    particulate air ventilation, or wet working, are adequate to control
    airborne concentrations as required by 20.1701. Determine if
    airborne concentrations are monitored using the preferred method of
    breathing zone air sampling of workers. When individual respirators
    are used, verify that the respiratory protection program elements
    specified in 20.1703 are properly implemented following the guidance
    in Regulatory Guide 8.15 Rev 1 (October 1999). The protection
    factors afforded by respirators listed in 10 CFR 20 Appendix A may
    be used when calculating the internal doses to workers.

l.  Review training records and verify that the information in RG 8.13
    or a copy of the guide was provided to all fertile females and their
    supervisors. Declaration of pregnancy is voluntary and is required
    to be in writing; the declaration is revocable by the woman, who
    does not need to provide any \"medical proof\" of pregnancy. For
    declared pregnant workers, verify that the licensee=s dose records
    conform to 10 CFR 20.1502 and guidance in RG 8.7 and RG 8.36.
    Confirm that the dose to the embryo/fetus (not to the worker) is
    within the limit specified in 10 CFR 20.1208

03.04 [Effluent Control and Environmental Monitoring]{.underline}

a.  Review the records for radioactive airborne releases to the
    environment since the last inspection and verify that the constraint
    limit (10 millirem/year) specified in 10 CFR 20.1101(d) was
    satisfied. If exceeded, verify that reporting was completed as
    specified in 10 CFR 20.2203 and the limits specified in 10 CF
    20.1301 and 10 CFR 20 Appendix B, Table 2 were satisfied. Acceptable
    methods to estimate the dose from air emissions are described in
    Regulatory Guide 4.20, \"Constraint on Releases of Airborne
    Radioactive Materials to the Environment for Licensees Other than
    Power Reactors.\" Of these methods, the computer program COMPLY
    (available free from EPA) is most often used.

> Review records of radioactive liquid releases since the last
> inspection and verify that the limits in 10 CFR 20 Appendix B, Table
> 2, the TS or DP conditions were met. For releases to the sanitary
> sewer system, verify that the solubility and other requirements in 10
> CFR 20.2003 were satisfied. Solubility assessment techniques can be
> found in Information Notice 94-07.
>
> Review the most recent calibration results for the effluent radiation
> monitoring systems (RMS) and determine whether the licensee calibrated
> the RMS appropriately for: (1) the isotope mix (or average gamma
> energy); (2) concentration level; and (3) licensee procedures. The
> isotopic mix may be different from the normal mix found during reactor
> operations. Confirm that batch releases were properly sampled,
> analyzed, and authorized for release by the RSO or designated
> alternate. Verify that controls are effective in preventing releases
> through unmonitored release pathways.

b.  Review the techniques used to implement the effluent releases ALARA
    program. This may include storage for decay, filtering,
    solidification of liquids, and other techniques as appropriate.

c.  Verify that on-site and off-site environmental samples are taken at
    the locations and at the frequencies specified in the DP. Review the
    sample preparation and laboratory measurement techniques to verify
    that adequate sensitivities were achieved. Verify the measurement
    accuracy through the laboratory quality control program. Review the
    records and verify that the projected doses to the public from site
    activities were within the limits specified in 10 CFR 20.1301 and
    ALARA.

03.05 [Change Control]{.underline}. Changes to the DP are likely when
unanticipated events or conditions develop as dismantlement progresses.
For example, contamination from a hidden leak may require significant
additional decontamination and dismantlement efforts. In such a case,
major changes may be required to the completion schedule, cost and dose
projections, volume of waste, and the need for new procedures and
specialized equipment. Verify that changes are identified, reviewed,
approved, and recorded in accordance with the administrative process
described in 10 CFR 50.59, and the DP or TS. The inspector should (1)
assess the rigor of engineering and management reviews; (2) determine
whether the proposed activities are bounded by the general
decommissioning safety evaluation; and, (3) identify whether the
changes, tests, or experiments are consistent with 10 CFR 50.59. NRC
inspector/staff review of licensee safety evaluations are conducted
after licensee management final approval and should be inspected prior
to the activity occurring, if possible.

The safety review process used by a licensee during decommissioning
should be comparable to the program in use during reactor operations. If
a licensee enters into long-term storage and implements a 50.59 program
that has not been exercised for a long period of time, the licensee may
experience a loss of expertise in the conduct of the technical or safety
reviews due to staff changes. Regardless, the inspector should assess
whether the program: (1) reflects the licensee\'s current decommissioned
organization and staffing configuration; (2) incorporates the most
recent regulatory requirements and guidance; (3) applies appropriate
management and technical reviews; and, (4) includes appropriate audit
considerations to assure that safety questions, evaluations, and
justifications are performed and documented.

03.06 [Audits, and Reviews]{.underline}

a.  Verify that the decommissioning safety committee members have the
    necessary knowledge and experience in areas important to
    decommissioning. For example, experience with oversight of
    contractors, radiological and industrial safety, and large volume
    waste handling and shipping become important. The composition and
    number of members on the committee should be adjusted by the
    licensee to deal with emerging issues as long as minimum staffing
    specified in the TS or DP is maintained. The required meeting
    frequency will be described in the TS or DP. However, meetings
    should be held as frequently as needed to support a timely
    decommissioning schedule.

b.  The responsibilities of the decommissioning safety committee will be
    described in the TS, DP, and/or the committee charter. A
    performance-based assessment should be performed to assure that
    significant decommissioning activities are independently and
    effectively reviewed, as required by the DP, the TSs, and docketed
    commitments. The inspector and/or PM should observe the conduct of a
    safety review committee and assess the effectiveness of this review
    body at: (1) questioning risks and benefits and the technical
    adequacy of the particular activity; (2) providing an independent
    safety perspective; and, (3) contributing to plant safety through,
    in part, the incorporation of lessons learned and experience.
    Minutes of meeting can also be used to verify that the committee
    functions and oversight are as required. For example, committee
    concurrence with policy changes (but not specific work instructions)
    may be required. Observe a meeting if possible.

> If periodic audits are performed, a review of the depth of findings,
> timeliness and thoroughness of corrective actions for identified
> concerns will indicate the effectiveness of the audit program.

c.  Management should evaluate audit and review findings, events,
    conditions, reportable occurrences, and other issues and problems
    that arise. Management response should be in writing and include a
    root causes analysis and effective corrective actions. Management
    response should meet schedule commitments. The response should
    clearly state the corrective action taken to prevent recurrence. In
    the event that corrective action can not be taken immediately, the
    response should include a scheduled date for initiation and
    completion of corrective actions. Management should report the
    status of corrective action to their management or to the safety
    review committee. Once complete, management should report on the
    effectiveness of the corrective actions. Management or the safety
    review committee should assess the effectiveness of the corrective
    actions. The licensee\'s program should provide contingencies for
    decisions by higher level management should an impasse be reached
    between the reviewing and responsible organizations.

03.07 [Emergency Planning]{.underline}

a.  The NRC-approved Emergency Plan required for reactor operations may
    be eliminated once the reactor fuel is removed from the site.
    However, response to emergencies involving radioactive materials and
    industrial hazards will continue to be required. For example,
    planning for response to a contaminated, injured worker or a fire in
    a radiation area is appropriate. Verify that emergency preparedness
    policies are documented, kept up-to-date, posted or otherwise made
    available to supervisors and workers. Emergency response during
    decommissioning may rely on support from off-site agencies. Verify
    that support agreements with the local fire department, police,
    ambulance service, and hospitals are current.

b.  Drills and exercises may not be required by the DP or TS. If the
    licensee is required to conducts drills, verify that they are
    conducted at the required frequency and that emergency program
    weaknesses are identified and corrected. Review the training
    programs and pre-job worker briefings and verify that the policies
    and procedures are disseminated and understood. Verify that
    emergency response equipment required by the DP and licensee=s
    procedures is available and maintained.

> Fire protection at the decommissioning facility can be part of the
> emergency response. The licensee should determine whether the quality
> and amount of fire prevention, detection, and mitigation features as
> described in licensing basis documentation are sufficient to ensure
> that decommissioning can be performed safely and that the spread of
> contamination can be minimized and controlled in event of a fire.
> Other licensee considerations could include strategies and/or systems
> specifically implemented for insurance and indemnity purposes. The
> inspector should focus on fire protection requirements as described in
> license requirements. These may include: (1) periodic updating of the
> licensee fire plan to reflect the status of decommissioning; (2)
> training; (3) conduct of fire fighting drills; and, (4) fire detection
> and suppression systems. Inspection emphasis should also be placed on
> ensuring that areas undergoing decontamination have fire detection
> and/or suppression systems capable of fulfilling their intended
> function. Verify that the fire protection program or plan addresses
> both transient and permanent fire loading conditions.

03.08 [Solid Radioactive Waste Management and
Transportation]{.underline}

a.  The licensee may dispose of small quantities of LSA and SCO
    byproduct and activated waste by shipment to a disposal site for
    direct burial. This is usually the case for zero power reactors and
    AGN facilities. ( Note that waste with a specific activity less that
    0.002 μCi/g is exempt from 10 CFR Part 71 regulations by 71.10.)
    Disposal of larger quantities of waste will likely be arranged
    through a waste broker. The broker may provide services such as
    sorting, solidification, reclamation, decontamination ,volume
    reduction, metal melt, shipping containers, transportation, and
    ultimate disposal of waste at an off-site burial facility.

> Verify that the licensee has arranged for the disposal of the waste
> type, form, quantity, isotope mix, and mixed waste that will be
> generated during decontamination and dismantlement. Verify that the
> waste broker is licensed to accept these wastes in accordance with 10
> CFR 20.2001.

b.  Ropes and stanchions do not provide adequate control for temporary
    waste storage areas unless under constant surveillance. Review the
    management controls, policies, procedures and physical barriers that
    ensure the safe placement and removal of waste in temporary storage.
    Verify that waste is properly packaged, labeled, and inventoried
    while in storage. Confirm that radiation survey results and posting
    are current.

c.  A graded approach to training is preferred. Review the content and
    depth of training provided to radioactive waste handlers. Review
    training, testing, or certification records of the workers and
    contractors. For training provided on site, review the credentials
    of the instructors. Review the management process to identify
    updates, regulatory changes, and provisions for retraining.

d.  Verify that certified Industrial Packages (IP-1, -2, -3) are used to
    ship low specific activity and surface contaminated objects waste in
    accordance with 49 CFR 173.427. Dose rate and surface contamination
    limits for packages is specified in 49 CFR 173.441 and 443.The
    activity limits for use of Type A and Type B packages is specified
    in 49 CFR 173.431. Users of Type B packages and spent fuel casks
    (highway route controlled quantities) must have a letter from the
    NRC (NMSS) approving their quality assurance program. Review the
    implementation of the elements of the approved program. Additional
    inspection guidance in this area can be found in Inspection
    Procedure 86740.

e.  Review the radioactive waste procedures and ensure that the
    responsibility for classifying (DOT and NRC criteria) and verifying
    the classification of waste is clearly assigned. Acceptable methods
    for identifying the isotopic mix and concentrations are needed to
    classify the waste include waste stream analysis, laboratory
    analysis of individual batch samples, or the site characterization
    report. Review the records for waste prepared for shipment and
    verify the classification is proper and the package labeled with the
    class as required by 10 CFR 61.57. Most waste is expected to be in
    the form of dry active waste (DAW). For mixed waste containing
    radioactive material mixed with chemical, toxic, biological, or
    gaseous agents not meeting the criteria in 10 CFR 61.56 for direct
    burial, verify that acceptable arrangements with a waste processor
    will allow disposal.

f.  Review shipping manifests for completeness and accuracy of the data.
    Burial sites may impose restrictions in addition to NRC and DOT
    record requirements. Verify that the receiver is authorized to
    accept the shipment and acknowledges receipt. Records should also
    include the results of radiation surveys of the truck, incoming and
    outgoing inspections of the truck, placarding, and driver briefings
    for exclusive use shipments.

g.  Documentation that the wastes were processed, controlled, stored,
    and transported in accordance with procedures that were qualified
    through a licensee\'s DP approved program. If procedures were
    modified by the licensee, under 10 CFR 50.59, determine the
    acceptability of the modifications via inspection of the licensee\'s
    records of the changes, including the safety evaluation for each
    change, to ensure that the changes did not result in a product that
    does not possess the long-term structural stability required by 10
    CFR 61.56(b)(1). Additional guidance is in the most recent version
    of the NRC Technical Position on Waste Form.

69013-04 RESOURCE ESTIMATE

Inspection resources for this inspection procedure will vary from
site‑to‑site based on NRC management\'s assessment of licensee
performance. In addition, inspection resources will be dependent on the
phase of decommissioning being implemented. It is estimated that during
active periods of decommissioning approximately 24 onsite inspection
hours will be needed to adequately assess and document licensee
performance quarterly.

69013‑05 REFERENCES

NUREG/CR-6204 Questions and Answers Based on Revised 10 CFR Part 20

NUREG/CR-5569 Health Physics Positions Data Base (HPPOS)

NUREG -1660 U.S. - Specific Schedules of Requirements for Transport of
Specified Types of Radioactive Material Consignments

> (Note; Schedule Summary on page 7 is a useful compilation of shipping
> requirements.)

NUREG -1608 Categorizing and Transporting Low Specific Activity
Materials and Surface Contaminated Objects

ANS - 15.19 Shipment and Receipt of Special Nuclear Material (SNM) by
Research Reactor Facilities

NRC Inspection Procedure 86740 Inspection of Transportation Activities

NRC Information Notice No. 94-23; Guidance to Hazardous, Radioactive and
Mixed Waste Generators on the Elements of a Waste Minimization Program

NRC Information Notice No. 92-72: Employee Training and Shipper
Registration Requirements for Transporting Radioactive Materials

NRC Information Notice No. 85-92: Surveys of Wastes Before Disposal from
Nuclear Reactor Facilities

END