10.0 ENVIRONMENTAL PROTECTION 

10.1 PURPOSE OF REVIEW 

The purpose of this review is to determine whether an applicant for the construction and 

operation of a mixed oxide (MOX) fuel fabrication facility has established environmental 

protection measures that are adequate to protect public health and the environment and comply 

with the regulatory requirements imposed by the Commission in 10 CFR Parts 20, 51, and 70. 

In addition, pursuant to 10 CFR Part 51, the NRC will determine if the applicant has submitted 

an environmental report that is adequate for NRC use in preparation of an Environmental 

Impact Statement (EIS) for licensing a MOX fuel fabrication facility, including the construction 

approval. This determination will be coordinated through the Division of Waste Management 

(DWM) since on May 17, 1999, the Office of Nuclear Material Safety and Safeguards (NMSS) 

assigned DWM the responsibility to prepare each NMSS EIS. As a result, guidance for 

reviewing an environmental report used to prepare an EIS is not provided in this chapter. 

The Division of Fuel Cycle Safety and Safeguards currently retains the responsibility for 

determining if, pursuant to 10 CFR Part 51, an environmental report is adequate to support a 

licensing action that will result in the preparation of an Environmental Assessment (EA) and a 

Finding of No Significant Impact (FONSI). However, this type of licensing action is not 

anticipated until after the MOX fuel fabrication facility is licensed to possess and use special 

nuclear material (SNM). Staff should contact DWM for coordination and guidance and refer to 

the supplementary guidance in Appendix F to this Standard Review Plan (SRP). 

10.2 RESPONSIBILITY FOR REVIEW 

Primary: Environmental Engineer/Scientist 

Secondary: Project Manager 

Supporting: Health Physicist Reviewer 

Chemical Safety Reviewer 

Primary Reviewer of SRP Chapter 15.0 

Environmental Protection Inspector 

10.3 AREAS OF REVIEW 

Regulatory requirements for environmental protection are contained in 10 CFR Parts 20, 51, 

and 70. The NRC staff environmental review under Parts 20 and 70 is focused on that part of 

the applicant's facility-wide safety program that is established to control and assess the level of 

radioactive releases (gaseous, liquid, and solid) to the environment during normal and 

anticipated operations. Therefore, the effluent control portion of the applicant's radiation 

protection program, as well as effluent and environmental monitoring practices, are reviewed.

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This review complements the review conducted under Section 9.2 of this SRP, which addresses 

the radiation protection program as it applies to worker safety. 

An applicant engaged in the fabrication of MOX fuel must perform an integrated safety analyais 

(ISA) in accordance with Subpart H of 10 CFR Part 70, and submit an ISA Summary in 

accordance with 10 CFR 70.65. Guidance on the ISA is covered in Chapter 5.0 of this SRP. 

The environmental review of the ISA Summary should include the identified potential accident 

sequences that result in radiological releases to the environment, the items relied on for safety 

(IROFS) that are specified by the applicant to reduce the risk of these accidents, and the 

associated management measures that provide reasonable assurance that the IROFS will 

perform their designated safety functions as required by 10 CFR Part 70. 

Thus, environmental protection includes three main components: (1) the radiation protection 

program, (2) effluent and environmental monitoring for normal and off-normal operations, and 

(3) the ISA Summary and other ISA documentation, as necessary. 

Areas of review for each of these components should include: 

A. Radiation Safety 

i. As Low As Reasonably Achievable (ALARA) goals for effluent control; 

ii. Effluent controls to maintain public doses ALARA; 

iii. ALARA reviews and reports to management; and 

iv. Waste minimization practices, and for new operations, design plans for waste 

minimization. 

B. Effluent and Environmental Monitoring 

i. In-place filter testing procedures for air cleaning systems; 

ii. Known or expected concentrations of radionuclides in effluents; 

iii. Physical and chemical characteristics of radionuclides in discharges; 

iv. Discharge locations; 

v. Environmental media to be monitored and the sample locations; 

vi. Sampling collection and analysis procedures, including the minimum detectable 

concentrations of radionuclides, equipment used, and calibration information; 

vii. Action levels and actions to be taken when the levels are exceeded; 

viii. Permits, including air discharge and National Pollutant Discharge and Elimination 

System permits; 

ix. Leak detection systems for ponds, lagoons, and tanks; 

x. Pathways analysis methods to estimate public doses; 

xi. Recording and reporting procedures, including event notification; and 

xii. Solid waste handling and disposal programs. 

C. Safety Assessment of the Design Basis or Safety Program Description and ISA Summary

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The safety assessment of the design basis (construction approval review) or the Safety 

Program Description and ISA Summary (application for a license to possess and use SNM) 

address similar material, as follows: 

i. Accident sequences (and associated facility processes) which, if unmitigated, result in 

releases to the environment; 

ii. Likelihood and consequences of these accident sequences; 

iii. Safety controls relied on to reduce the unmitigated risk from high to an acceptable level; 

and 

iv. Availability and reliability of safety controls. 

10.4 ACCEPTANCE CRITERIA 

10.4.1 Regulatory Requirements 

A. 10 CFR Part 20, specifically the effluent control and treatment measures necessary to meet 

the dose limits and dose constraints for members of the public specified in Subparts B, D, 

and F; the requirements for minimization of contamination specified in 10 CFR 20.1406; the 

survey requirements specified in Subpart F; the waste disposal requirements of Subpart K; 

the records requirements of Subpart L; and the reporting requirements of Subpart M. 

B. 10 CFR Part 51, specifically requiring the applicant to establish effluent and environmental 

monitoring systems to provide the information required by 10 CFR 51.60(a). 

C. 10 CFR Part 51, specifically requiring the applicant to submit an environmental report as 

required by 10 CFR 51.60(b), or to support a categorical exclusion as described in 

10 CFR 51.22(c). 

D. 10 CFR Part 70, requiring the application to demonstrate that proposed facilities and 

equipment, including measuring and monitoring instruments and devices for the disposal of 

radioactive effluents and wastes, are adequate to protect public health and the environment 

as specified in 10 CFR 70.22(a)(7). 

E. 10 CFR Part 70, requiring the application for a plutonium processing facility as defined in 

10 CFR 70.4 to include a safety assessment of the design basis of the principal structures, 

systems, and components of the facility, including provisions for protection against natural 

phenomena, as specified in 10 CFR 70.22(f). 

F. 10 CFR Part 70, requiring an application for a facility to fabricate MOX fuel to include an ISA 

Summary that lists the IROFS established by the applicant and other elements as 

described in 10 CFR 70.65(b).

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10.4.2 Regulatory Guidance 

Regulatory guidance for environmental protection is contained in: 

A. American Nuclear Standards Institute (ANSI). ANSI-N13.1-1982, "Guide to Sampling 

Airborne Radioactive Materials in Nuclear Facilities." 

B. . ANSI-N42.18-1980, "Specification and Performance of On-site Instrumentation for 

Continuously Monitoring Radioactive Effluents." 

C. National Council on Radiation Protection and Measurements (NCRPM) Report No. 123, 

"Screening Models for Releases of Radionuclides to Atmosphere, Surface Water, and 

Ground." January 1996. 

D. Nuclear Regulatory Commission (U.S.) (NRC). Information Notice 94-07, "Solubility 

Criteria for Liquid Effluent Releases to Sanitary Sewerage Under the Revised 10 CFR Part 

20." NRC: Washington, D.C. January 28, 1994. 

E. . Information Notice 94-23: "Guidance to Hazardous, Radioactive and Mixed Waste 

Generators on the Elements of a Waste Minimization Program." NRC: Washington, D.C. 

March 1994. 

F. . Regulatory Guide (RG) 4.5, "Measurements of Radionuclides in the Environment 

Sampling and Analysis of Plutonium in Soil." NRC: Washington, D.C. May 1974. 

G. . RG 4.15, Revision 2, "Quality Assurance for Radiological Monitoring Programs 

(Normal Operations)-Effluent Streams and the Environment." NRC: Washington, D.C. 

H. . RG 4.16, Revision 2, "Monitoring and Reporting Radioactivity in Releases of 

Radioactive Materials in Liquid and Gaseous Effluents from Nuclear Fuel Processing and 

Fabrication Plants and Uranium Hexafluoride Production Plants." NRC: Washington, D.C. 

I. . RG 4.20, "Constraint on Releases of Airborne Radioactive Materials to the 

Environment for Licensees Other than Power Reactors." NRC: Washington, D.C. 

December 1996. 

J. . RG 8.37, "ALARA Levels for Effluents from Materials Facilities." NRC: Washington, 

D.C. July 1993. 

10.4.3 Regulatory Acceptance Criteria 

A. Radiation Safety 

In accordance with 10 CFR 20.1101, each licensee must implement a radiation protection 

program, which is discussed in detail in Chapter 9.0 of this SRP. The environmental review 

of the radiation protection program focuses on the applicant's methods to maintain public

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doses ALARA in accordance with 10 CFR 20.1101. NRC guidance on compliance with 

these regulations can be found in Regulatory Guide 8.37, "ALARA Levels for Effluents from 

Materials Facilities," July 1993. 

Specifically, 10 CFR 20.1101(d) requires the applicant to establish a constraint on air 

emissions of radioactive material to the environment, excluding Radon-222 and its decay 

products, such that the individual member of the public likely to receive the highest dose will 

not be expected to receive a total effective dose equivalent (TEDE) in excess of 0.1 mSv 

(10 mrem) per year from these emissions. The applicant must have procedures to report to 

the NRC in accordance with 10 CFR 20.2203 when this dose constraint is exceeded and to 

take prompt appropriate corrective action to ensure against recurrence. NRC guidance on 

compliance with this regulation can be found in Regulatory Guide 4.20, "Constraint on 

Releases of Airborne Radioactive Materials to the Environment for Licensees Other Than 

Power Reactors," December 1996. 

The environmental review of the radiation protection program also focuses on the 

applicant's waste minimization practices. Applicants for new licenses are required to 

comply with 10 CFR 20.1406, which states that the applicant must describe how facility 

design procedures for operation will minimize, to the extent practical, contamination of the 

facility and the environment, facilitate eventual decommissioning, and minimize, to the 

extent practical, the generation of radioactive waste. Applicants requesting amendment or 

renewal of existing licenses must minimize and control waste generation during operations 

as part of the radiation protection program in accordance with 10 CFR 20.1101 

[62 FR 39082]. 

Guidance for waste minimization programs can be found in NRC Information Notice 94-23, 

"Guidance to Hazardous, Radioactive and Mixed Waste Generators on the Elements of a 

Waste Minimization Program," March 25, 1994. 

The proposed radiation protection program is acceptable if, in addition to the acceptance 

criteria outlined in Section 9.2, it satisfies the following criteria: 

i. Radiological (ALARA) Goals for Effluent Control 

ALARA goals for effluent control are set at a modest fraction (10% to 20%) of the values 

in Appendix B, Table 2, Columns 1 and 2 and Table 3 and the external exposure limit in 

10 CFR 20.1302(b)(2)(ii), or the dose limit for members of the public, if the applicant 

proposes to demonstrate compliance with 10 CFR 20.1301 through a calculation of the 

TEDE to the individual likely to receive the highest dose. 

An applicant's constraint approach is acceptable if it is consistent with guidance found in 

Regulatory Guide 4.20 and the applicant's description of the constraint approach 

provides sufficient detail to demonstrate specific application of the guidance to proposed 

routine operations and nonroutine operations, including anticipated events.

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ii. Effluent Controls To Maintain Public Doses ALARA 

The applicant describes and commits to using effluent controls (e.g., procedures, 

engineering controls, and process controls) to maintain public doses ALARA. Common 

control practices include filtration, encapsulation, adsorption, containment, recycling, 

leakage reduction, and the storage of materials for radioactive decay. The applicant 

demonstrates a commitment to reducing unnecessary exposure to members of the 

public and releases to the environment. 

Effluent controls during normal and likely facility conditions: 

a. Are capable of handling the expected volume of potentially radioactive Waste. 

b. Are compartmentalized to reduce the potential for cross-contamination. For 

example, storm water and sanitary sewage lines should be separate from lines 

carrying radioactive effluents. Laundry facilities and personnel decontamination 

facilities should send effluents to radioactive waste. There should be no means by 

which radioactive waste can bypass the effluent controls and be directly released to 

the environment. 

c. Are capable of safe shutdown, consistent with the operating status of the facility. 

d. Are capable of safely handling the chemical characteristics of the effluent. For 

example, effluent controls in contact with strong acids or caustics should be 

corrosion resistant. 

e. Achieve a decontamination factor for each radionuclide sufficient to reduce the total 

radioactivity to an acceptable release level on a "once" through treatment basis. 

Provisions are made to recirculate effluents for further decontamination when 

radioactivity is above an acceptable release level. 

iii. ALARA Reviews and Reports to Management 

As part of the annual review of the content and implementation of the radiation 

protection program as discussed in Section 9.2, the applicant commits to reviewing the 

effluent controls to maintain public doses ALARA. This review includes analysis of 

trends in release concentrations, environmental monitoring data, and radionuclide 

usage; determines whether operational changes are needed to achieve the ALARA 

effluent goals; and evaluates all designs for system installations or modifications. The 

applicant also includes a commitment to report the results to senior management along 

with recommendations for changes in facilities or procedures that are necessary to 

achieve ALARA goals.

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iv. Waste Minimization 

The application contains a description of how facility design and procedures for 

operation will minimize, to the extent practical, contamination of the facility and the 

environment and minimize, to the extent practical, the generation of radioactive waste. 

A waste minimization programs is acceptable if it includes: 

a. Top management support; 

b. Identification of responsibilities for waste minimization activities and assessments; 

c. Methods to characterize waste generation, including types and amounts, and waste 

management costs, including costs of regulatory compliance, paperwork, 

transportation, treatment, storage, disposal, etc.; 

d. Periodic waste minimization assessments to identify waste minimization 

opportunities and solicit employee or external recommendations; 

e. Provisions for technology transfer to seek and exchange technical information on 

waste minimization; 

f. Provisions to incorporate operational experience; and 

g. Methods for implementation and evaluation of waste minimization recommendations. 

B. Effluent and Environmental Monitoring 

The applicant is required to make, or cause to be made, as appropriate, surveys of radiation 

levels in unrestricted and controlled areas and radioactive material in effluents released to 

unrestricted and controlled areas to demonstrate compliance with the dose limits for 

individual members of the public as described in 10 CFR 20.1301. Accordingly, the staff's 

acceptance criteria for the applicant's effluent and environmental monitoring for normal and 

off-normal operations are described in Sections 10.4.3(B)(i) and 10.4.3(B)(ii). 

i. Effluent Monitoring 

The reviewer should find that the applicant's effluent monitoring is acceptable if it meets 

the following criteria: 

a. The known or expected concentrations of radioactive materials in airborne and liquid 

effluents are below the limits in 10 CFR Part 20, Appendix B, Table 2 or below site

specific limits established in accordance with § 20.1302(c) and are ALARA. 

If, in accordance with § 20.1302(c), the applicant proposes to adjust the effluent 

concentrations in Appendix B to 10 CFR Part 20 to take into account the actual 

physical and chemical characteristics of the effluents, the applicant provides

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information related to aerosol size distributions, solubility, density, radioactive decay 

equilibrium, and chemical form. This information is complete and accurate for the 

radioactive materials to justify the derivation and application of the alternative 

concentration limits. 

b. If the licensee proposes to demonstrate compliance with 10 CFR 20.1301 through a 

calculation of the TEDE to the individual likely to receive the highest dose in 

accordance with 10 CFR 20.1302(b)(1), calculation of the TEDE by pathways 

analyses uses appropriate models and codes and assumptions that accurately 

represent the facility, the site, and the surrounding area; assumptions are 

reasonable; input data is accurate; all applicable pathways are considered; and the 

results are interpreted correctly. 

NCRP Report No. 123, "Screening Models for Releases of Radionuclides to 

Atmosphere, Surface Water, and Ground," January 1996, provides acceptable 

methods for calculating the dose from radioactive effluents. Computer codes are 

acceptable tools for pathways analysis if the applicant is able to show that the code 

has undergone validation and verification to demonstrate the validity of estimates 

developed using the code for established input sets. Dose conversion factors used 

in the pathways analyses are acceptable if they are based on the methodology 

described in International Council on Radiation Protection 30, "Limits for Intakes of 

Radionuclides by Workers," as reflected in Federal Guidance Report 11. Such 

methods are acceptable for determining the dose to the maximally exposed 

individual during normal facility operations and anticipated events. 

c. All liquid and airborne effluent discharge locations are identified and monitored. 

Monitoring locations are identified, and for those effluent discharge points that have 

input from two or more contributing sources within the facility, sampling each 

contributing source is evaluated for effective effluent control. 

d. Airborne effluents from all routine operations and nonroutine operations, as well as 

anticipated events associated with the facility, including effluents from areas not 

used for processing special nuclear material such as laboratories, experimental 

areas, storage areas, and fuel element assembly areas, are continuously sampled. 

Effluents are sampled unless the applicant has established, by periodic sampling or 

other means, that radioactivity in the effluent is insignificant and will remain so. In 

such cases, the effluent is sampled at least quarterly to confirm that effluents are not 

significant. For the purposes of this SRP, an effluent is significant if the 

concentration averaged over a calendar quarter is equal to 10% or more of the 

appropriate concentration listed in Table 2 of Appendix B to 10 CFR Part 20. 

e. The sample collection and analysis methods and frequencies are appropriate for the 

effluent medium and the radionuclide(s) being sampled. Sampling methods provide 

reasonable assurance that representative samples are obtained by use of 

appropriate sampling equipment and sample collection and storage procedures. For

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liquid effluents, representative samples are taken at each release point for the 

determination of concentrations and quantities of radionuclides released to an 

unrestricted area, including discharges to sewage systems. For continuous 

releases, samples are continuously collected at each release point. For batch 

releases, a representative sample of each batch is collected. If periodic sampling is 

used in lieu of continual sampling, the applicant shows that the samples are 

representative of actual releases. Monitoring instruments are calibrated at least 

annually, or more frequently if suggested by the manufacturer. 

f. Radionuclide specific analyses are performed on selected composited samples 

unless either: 

"* The gross alpha and gross beta activities are so low that individual radionuclides 

could not be present in concentrations greater than 10 % of the concentrations 

specified in Table 2 or 3 of Appendix B to 10 CFR Part 20; or 

"* The radionuclide composition of the sample is known through operational data, 

such as the composition of the feed material. 

Monitoring reports in which estimates of quantities of individual radionuclides are 

based on methods other than direct measurement include an explanation and 

justification of how the results were obtained. 

Examples of cases in which operational data may not be adequate for the 

determination of radionuclide concentration are (1) facilities processing uranium in 

which extraction, ammonium diuranate precipitation, ion exchange, or other 

separation processes could result in concentration of thorium isotopes (principally 

Th-234); (2) facilities in which uranium of varying enrichments is processed; and (3) 

facilities processing plutonium in which significant variation in the Pu-238/Pu-239 

ratio among batches and the continuous in-growth of Am-241 would preclude the 

use of feed material data to determine the radionuclide composition of effluents. 

Radionuclide analyses are performed more frequently than usual under three 

circumstances: (1) at the beginning of the monitoring program until a predictable 

and consistent radionuclide composition in effluents is established; (2) whenever 

there is a significant unexplained increase in gross radioactivity in effluents; or (3) 

whenever a process change or other circumstance might cause a significant 

variation in the radionuclide composition. 

g. The minimum detectable concentration (MDC) for sample analyses is not more than 

5% of the concentration limits listed in Table 2 of Appendix B to 10 CFR Part 20. If 

the actual concentrations of radionuclides in samples are known to be higher than 

5% of the 10 CFR Part 20 limits, the analysis methods need only be adequate to 

measure the actual concentration. However, in such cases, the MDC is low enough 

to accommodate fluctuations in the concentrations of the effluent and the uncertainty 

of the MDC.

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h. The laboratory quality control (QC) procedures are adequate to support the validity 

of the analytical results. These QC procedures include the use of established 

standards such as those provided by the National Institute of Standards and 

Technology, as well as standard analytical procedures, such as those established by 

the National Environmental Laboratory Accreditation Conference. 

i. The proposed action levels, and actions to be taken if the action levels are 

exceeded, are appropriate. The action levels are incremental, such that each 

increasing action level results in a more aggressive action to assure and control 

effluents. A slightly higher than normal concentration of a radionuclide in effluent 

triggers an investigation into the cause of the increase. An action level is specified 

that will result in the shutdown of an operation if this level is exceeded. These action 

levels are selected based on the likelihood that a measured increase in 

concentration could indicate potential violation of the effluent limits. 

j. The descriptions of applicable Federal and/or State standards for discharges and 

any permits issued by local, State, or Federal governments for gaseous and liquid 

effluents are complete and accurate. 

k. The systems for the detection of leakage from ponds, lagoons, and tanks are 

adequate to detect and assure against any unplanned releases to groundwater, 

surface water, or soil. 

1. Releases to sewer systems are controlled and maintained to meet the requirements 

of 10 CFR 20.2003, including (i) the material is water soluble; (ii) known or expected 

discharges meet the effluent limits of 10 CFR 20 Appendix B, Table 3; and (iii) the 

known or expected total quantity of radioactive material released into the sewer 

system in a year does not exceed 5 Ci (185 GBq) of 3H, 1 Ci (37 GBq) of 14C, 

and 1 Ci (37 GBq) of all other radioactive materials combined. Solubility is 

determined in accordance with the procedure described in NRC Information Notice 

94-07. 

m. Reporting procedures comply with the requirements of 10 CFR 70.59 and the 

guidance specified in Regulatory Guide 4.16. Reports of the concentrations of 

principal radionuclides released to unrestricted areas in liquid and gaseous effluents 

are provided and include the MDC for the analysis and the error for each data point. 

n. The applicant's procedures and facilities for solid and liquid waste handling, storage, 

and monitoring result in safe storage of the material and timely disposition. 

ii. Environmental Monitoring 

The applicant's environmental monitoring is acceptable if it is commensurate with the 

scope of activities at the facility and the expected impacts of operations as identified in 

the environmental report and if it meets the following criteria:

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a. Background and baseline concentrations of radionuclides in environmental media 

have been established through sampling and analysis. 

b. A preoperational monitoring program is initiated prior to operation. The 

preoperational program should be of sufficient length to allow an adequate database 

for comparison with operational data. 

c. Monitoring includes sampling and analyses for important pathways for the 

anticipated types of radionuclides released from the facility into the environment 

from routine and nonroutine operations, including anticipated events. The pathways 

include air, surface water, groundwater, soil, sediments, and vegetation, as 

appropriate. Important environmental media are sampled to estimate radionuclide 

concentrations in important biota. 

d. The description of monitoring identifies adequate and appropriate sampling locations 

and frequencies for each environmental medium, the frequency of sampling, and the 

analyses to be performed on each medium. Sampling methods ensure that 

representative samples are obtained by use of appropriate sampling equipment, 

sample collection, and sample storage procedures. 

e. Monitoring procedures employ acceptable analytical methods and instrumentation to 

be used, and monitoring procedures and analytical methods are subject to quality 

controls. The applicant commits to a program of instrument maintenance and 

calibration appropriate to the instrumentation, as well as participation in round-robin 

measurement comparisons if the applicant proposes use of its own analytical 

laboratory for analysis of environmental samples. 

f. Appropriate action levels and actions to be taken if the levels are exceeded are 

specified for each environmental medium and radionuclide. 

Action levels are selected based upon a pathways analysis that demonstrates that 

below those concentrations, doses to the public will be below the limits in 

10 CFR Part 20, Subpart B, and are ALARA. The action levels specify the 

concentrations at which an investigation would be performed and levels at which 

process operations would be shut down. 

g. MDCs are specified for sample analyses and are at least as low as those selected 

for effluent monitoring in air and water. MDCs for sediment, soil, and vegetation are 

selected based upon the action levels to ensure that sampling and analytical 

methods are sensitive and reliable enough to support application of the action levels. 

h. Data analysis methods and criteria to be used for evaluating and reporting the 

environmental sampling results are appropriate and will indicate when an action level 

is being approached in time to take corrective actions.

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i. The description of the status of all licenses, permits, and other approvals of facility 

operations required by Federal, State, and local authorities is complete and 

accurate. 

j. Environmental monitoring is adequate to assess impacts to the environment from 

potential radioactive and nonradioactive releases as identified in high- and medium

risk accident sequences in the ISA. 

C. Safety Assessment of the Design Bases and the Safety Program Description and ISA 

Summary 

i. Safety Assessment of the Design Bases 

In accordance with 10 CFR 70.22(f), an applicant for a MOX fuel fabrication facility is 

required to submit a safety assessment of the design bases to allow the NRC to make a 

determination regarding construction approval. The design bases includes the principal 

structures, systems, and components (SSCs) that the applicant has identified to protect 

against the consequences of accidents and natural phenomena. The applicant's safety 

assessment of the design bases should be acceptable in the area of environmental 

protection if: 

a. Based on the level of design, the accidents analyzed in the applicant's safety 

assessment of the design bases, as described in Chapter 5.0, bound the types, 

consequences, and likelihoods of the accidents that could result in radiological 

releases to the unrestricted area. 

b. The applicant used acceptable methods for estimating consequences from accident 

sequences that result in radiological releases to the environment. The estimates are 

bounding; assumptions necessary from the level of design are appropriately 

conservative. For the purposes of this review, consequences include dose to the 

public and the 24-hour averaged release of radioactive material outside the 

restricted area as defined in 10 CFR 70.61. 

c. The applicant identified principal SSCs as part of the design bases that provide 

reasonable assurance that the applicant can construct a facility that will meet the 

performance requirements of 10 CFR 70.61. 

ii. ISA Summary 

As part of the license application for possession and use of SNM, pursuant to Subpart H 

to the 10 CFR Part 70, the applicant for a MOX fuel fabrication facility is required to 

submit a Safety Program Description and an ISA Summary. The ISA Summary includes 

IROFS identified to prevent or mitigate against accidents. The applicant's treatment of 

environmental protection in the Safety Program Description and ISA Summary (see 

Chapter 5.0 of this SRP) is acceptable if the applicant:

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a. Provides a complete list of accident sequences that result in radiological releases to 

the unrestricted area. 

b. Provides a reasonable estimate for the likelihood of each accident sequence . identified. 

c. Uses acceptable methods for estimating consequences from accident sequences 

that result in radiological releases to the environment. For the purposes of this 

review, consequences include dose to the public and the 24-hour averaged release 

of radioactive material outside the restricted area as defined in 10 CFR 70.61. 

Acceptable methods are described in NUREG/CR-6410, "Nuclear Fuel Cycle 

Facility Accident Analyses Handbook." 

d. Identifies IROFS for each accident sequence that results in consequences greater 

than the limits defined in 10 CFR 70.61. The IROFS prevent or mitigate risk 

sequences to an acceptable level of protection. 

e. Affords adequate levels of assurance to the IROFS to ensure that they will be 

reliable and available to perform their safety functions. This may be accomplished 

through configuration management, training, maintenance activities, or other 

management measures as appropriate. 

10.5 REVIEW PROCEDURES 

10.5.1 Acceptance Review 

The primary reviewer should perform an acceptance review to determine if the application 

adequately addresses the items in Section 10.3, "Areas of Review," for either the construction 

approval review or the review for a license to possess and use SNM. 

Guidance specific to the construction approval review and the review for a license to possess 

and use SNM is provided below. 

A. Construction Approval 

Specifically, the safety assessment of the design basis should address Section 10.3(C) 

consistent with the level of design. Sections 10.3(A) and (B) should be addressed to the 

extent that the material therein supports information provided in the environmental report, 

such as environmental monitoring as a mitigation measure. Where information is under 

development or not yet available, the applicant may use a commitment to providing the 

material with the application for a license to possess and use SNM in lieu of the actual 

material. 

B. License To Possess and Use SNM

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Specifically, the safety assessment of the license application should address 

Sections 10.3(A) through (C) in full. 

If the primary reviewer verifies that environmental protection is adequately addressed 

(construction or license to possess and use SNM), the primary reviewer should accept the 

application for the safety evaluation in Section 10.5.2. If the primary reviewer identifies 

significant deficiencies in the material provided, the primary reviewer should request that the 

applicant submit additional information prior to the start of the safety evaluation. 

10.5.2 Safety Evaluation 

After determining that the application is acceptable for review in accordance with 

Section 10.5.1, the primary reviewer should perform a safety evaluation against the acceptance 

criteria described in Section 10.4. On the basis of its review, the staff may request that the 

applicant provide additional information or modify the application to meet those acceptance 

criteria. 

Guidance specific to the construction approval review and the review for a license to possess 

and use SNM is provided below. 

A. Construction Approval 

The primary reviewer should establish that the applicant's facility design as described in the 

safety assessment of the design basis and other commitments, as they relate to 

environmental protection, meet or exceed the regulatory acceptance criteria in Section 

10.4.3(C)(i). 

As described in Appendix E to this SRP, the primary reviewer should coordinate with DWM 

during the preparation of the EIS. 

B. License To Possess and Use SNM 

The primary reviewer should establish that the applicant's facility design, operations, and 

chemical safety items provide reasonable assurance that they will function as intended and 

provide for the safe handling of licensed materials at the facility. The primary reviewer 

should identify the mechanisms that will allow the applicant to identify and correct potential 

problems. 

In support of the primary reviewer for Chapter 9.0, the environmental protection reviewer 

should determine whether the acceptance criteria in Chapter 9.0 have been met as they 

relate to environmental review of the radiation protection program. The primary reviewer 

should also support the primary reviewer for Chapter 8.0 to ensure that the acceptance 

criteria for Chapter 8.0 have been met as they relate to effluent controls to maintain public 

doses ALARA.

NUREG-1718 10.0-14
Environmental Protection

In support of the primary reviewer for Chapter 5.0, the environmental reviewer should 

review the ISA Summary. All accident sequences identified in the ISA that can have 

significant consequences due to releases to the unrestricted area should be reviewed to 

determine that the list of potential accidents is complete and properly identified. Detailed 

review should only be conducted of the accident sequences which, when left unmitigated, 

are rated as "high-consequence" events by the applicant, as well as approximately 10% of 

the "intermediate-consequence" events and a smaller number of the lower risk sequences. 

However, additional "high-consequence" and "intermediate-consequence" events may be 

evaluated based on the results of the initial review. 

The primary reviewer should provide input on the ISA Summary to the primary reviewer of 

Chapter 5.0 and input on management measures (if any) to the primary reviewer of 

Chapter 15.0. 

In addition, for renewal and amendment applications, review of environmental protection by the 

primary reviewer will include coordination with the inspector responsible for environmental 

protection (supporting reviewer). Any comments or concerns that the inspector identifies will be 

addressed and resolved, and the Safety Evaluation Report (SER) (described in Section 10.6.1) 

for the licensing action will contain a statement indicating if the inspection staff has any 

objections to approval of the proposed licensing action. In addition, if applicable, the primary 

reviewer will review inspection reports and semiannual effluent reports submitted in accordance 

with 10 CFR 70.59 to assure licensee performance in environmental protection. 

When the safety evaluation is complete, the primary reviewer, with assistance from the other 

reviewers, should prepare the environmental protection input for the SER as described in 

Section 10.6 using the acceptance criteria from Section 10.4. 

10.6 EVALUATION FINDINGS 

Documentation of the evaluation findings for the environmental protection review is contained in 

two types of products. The SER documents the review of the environmental protection 

measures and the design bases (construction approval) or the Safety Program Description and 

ISA Summary (license to possess and use SNM). The EA or EIS documents the staff's 

independent assessment of the environmental impacts of the proposed action. 

Environmental protection measures may be summarized in the EA or EIS. However, the EA or 

EIS does not become part of the license. The primary reviewer should document the safety 

evaluation by preparing material suitable for inclusion in the SER. The primary reviewer should 

describe the review, explain the basis for the findings, and state the conclusions (see the 

supplementary guidance in Appendix E to this SRP). 

If an EA or EIS was prepared for the licensing action, the date the document was issued should 

be reported in the environmental protection section of the SER. If the EA resulted in a FONSI, 

the FONSI's publication date in the Federal Register should be included in the SER. If an EIS 

was prepared, the SER would include the Federal Register publication date for the Record of

10.0-15 NUREG-1718
Environmental Protection

Decision. When applicable, the SER also documents the determination that an action meets a 

categorical exclusion. 

The staff could document the safety evaluation for the construction approval review as follows: 

The staff prepared an Environmental Impact Statement (EIS) on [publication date] for the 

construction approval for [insert name of facility]. Based on the EIS, the NRC stated in its 

Record of Decision [publication date in the Federal Register] that the preferred option was 

[state preferred option here]. 

For the preferred option, the staff reviewed the environmental protection measures for 

construction approval for [insert facility name] according to Chapter 10.0 of NUREG-1 718. 

The staff evaluated [state what was evaluated] and found [state what was found]. The staff 

concluded that the applicant's design basis has adequate environmental protection 

measures to protect the public and the environment against natural phenomena and the 

consequences of potential accidents in accordance with the regulatory requirements 

imposed by the Commission in 10 CFR Part 70. 

The staff could document the safety evaluation for the review for the license to possess and 

use SNM as follows: 

The staff prepared an Environmental Impact Statement (EIS) on [publication date] for this 

licensing action as required by 10 CFR Part 51.20. Based on the EIS, the NRC stated in its 

Record of Decision [publication date in the Federal Register] that the preferred option was 

[state preferred option here]. 

For the preferred option, the staff reviewed the environmental protection measures for 

issuing a license to possess and use SNM for [insert facility name] according to 

Chapter 10.0 of NUREG-1718. The staff evaluated [state what was evaluated] and found 

[state what was found]. The staff concluded that the applicant has adequate environmental 

protection measures, including: (1) environmental and effluent monitoring and (2) effluent 

controls to maintain public doses ALARA as part of the radiation protection program to 

protect public health and the environment and comply with the regulatory requirements 

imposed by the Commission in 10 CFR Parts 20, 51, and 70. 

10.7 REFERENCES 

American National Standards Institute (ANSI). ANSI-N13.1-1982, "Guide to Sampling 

Airborne Radioactive Materials in Nuclear Facilities." ANSI: La Grange, Illinois. 1982 

* ANSI-N42.18-1980, "Specification and Performance of On-site Instrumentation for 

Continuously Monitoring Radioactive Effluents." ANSI: La Grange, Illinois. 1980

NUREG-1 718 10.0-16
Environmental Protection

National Council on Radiation Protection and Measurements (NCRPM). Report No. 123 I & II, 

"Screening Models for Releases of Radionuclides to Atmosphere, Surface Water, and Ground." 

NCRPM: Bethesda, Maryland. January 1996. 

Nuclear Regulatory Commission (U.S.), "Domestic Licensing of Special Nuclear Material 

(10 CFR Part 70)," Federal Register: Vol. 64, No. 146. pp. 41338-41357. July 30, 1999. 

Nuclear Regulatory Commission (U.S.) (NRC). Information Notice No. 94-23, "Guidance to 

Hazardous, Radioactive and Mixed Waste Generators on the Elements of a Waste Minimization 

Program." NRC: Washington, D.C. March 25, 1994. 

Information Notice No. 94-07, "Solubility Criteria for Liquid Effluent Releases to 

Sanitary Sewerage Under the Revised 10 CFR Part 20." NRC: Washington, D.C. January 28, 

1994. 

NUREG-1 520, "Draft Standard Review Plan for the Review of a License Application for 

a Fuel Cycle Facility." NRC: Washington, D.C. April 1998. 

NUREG/CR-6410, "Nuclear Fuel Cycle Facility-Accident Analysis Handbook." NRC: 

Washington, D.C. March 1998. 

RG 3.52, Revision 2, "Standard Format and Content for the Health and Safety Sections 

of License Applications for Fuel Cycle Facilities." NRC: Washington, D.C. January 1995. 

RG 4.15, Revision 1, "Quality Assurance for Radiological Monitoring Programs 

(Normal Operations)-Effluent Streams and the Environment." NRC: Washington, D.C. 

February 1979. 

RG 4.16, Revision 1, "Monitoring and Reporting Radioactivity in Releases of 

Radioactive Materials in Liquid and Gaseous Effluents from Nuclear Fuel Processing and 

Fabrication Plants and Uranium Hexafluoride Production Plants." NRC: Washington, D.C. 

December 1985. 

RG 4.20, "Constraint on Releases of Airborne Radioactive Materials to the 

Environment for Licensees other Than Power Reactors." NRC: Washington, D.C. December 

1996. 

RG 8.37, "ALARA Levels for Effluents from Materials Facilities." NRC: Washington, 

D.C. July 1993. 

Rev. 6, "Materials Licensing Procedures Manual." NRC: NMSS/FCSS/Fuel Cycle 

Licensing Branch, Washington, D.C. April 1998.

10.0-17 NUREG-1718
11.0 PLANT SYSTEMS

11.1 PURPOSE OF REVIEW 

The purpose of this review is to determine that the plant systems-systems that are identified 

as items relied on for safety (I ROFS) pursuant to 10 CFR Part 70 and encompassed by the 

hazard and accident analyses of the integrated safety analysis (ISA)-will be available and 

reliable to perform their intended safety function when needed. Examples of plant systems are: 

(a) a ventilation system necessary to provide certain decontamination functions for normal, off

normal, and accident conditions and (b) an electrical distribution system necessary to support 

various IROFS. 

11.2 RESPONSIBILITY FOR REVIEW 

Primary: Discipline-specific engineers 

Secondary: Chemical Process Engineer, Health Physicist, Fire Protection 

Specialist, Human Factors Engineer 

Supporting: Primary Reviewers of SRP Sections 1.1 and 13.1 and 

Chapters 2.0, 3.0, 4.0 and 14.0 

Primary Reviewers of Applicable Sections of Standard Review Plan 

(SRP) Chapter 15.0 

11.3 AREAS OF REVIEW 

The review for the construction approval should focus on the layout and design of the plant 

systems, their components, and any related information considering the present stage of the 

applicant's design process. The review for the license application should focus on design 

modifications and any other system features not adequately described during the construction 

approval review. 

Also, the review for the licensee application for operations should encompass the adequacy of 

the design and operation of plant systems identified in the ISA Summary as IROFS such as 

electrical and ventilation systems. 

The license application for operations documentation, to be reviewed by the staff, should 

include specific items listed below for each system. The documentation for construction 

approval should address the following items to the extent practical considering the stage of 

design information available. 

A. Safety Function 

i. Identification of safety function as related to the performance requirements of § 70.61 

and the ISA; and

11.0-1 NUREG-1 718
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ii. Functional requirements stemming from the baseline design criteria (BDC) and the ISA 

process, including environmental design considerations (temperature, pressure, 

humidity, etc., resulting from normal, off-normal, and accident operating conditions) for 

IROFS with site factors (including natural phenomena that occur infrequently and 

conditions that are continuously present), defense-in-depth, reliability/availability goals 

(including continued operation of plant systems that perform essential utility services), 

and design features such as redundancy and independence (as appropriate) to ensure 

that reliability and availability goals are met. 

B. System Description 

i. Purpose (safety and nonsafety); 

ii. System design, including performance features; 

iii. Structures (including their materials, shielding, and physical protection) and 

components; 

iv. Instrumentation and controls (manual and automatic); 

v. System interfaces; 

vi. Drawings (including arrangements, plans, elevations, and sections for structures), 

specifications, and procedures; and 

vii. Assurance measures, including applicable industrial codes and standards, 

environmental qualification, quality assurance, inspection, testing, and maintenance. 

C. Safety Analysis 

i. How functional requirements are satisfied by system design; 

ii. How nonsafety features, as appropriate, do not prevent the plant system from 

performing its intended safety function; 

iii. How long-term performance, testing, and maintenance features are addressed; 

iv. How potential failure modes are analyzed, including consideration of communication 

failures, common-mode failures, and human errors; 

v. How material-related failure modes are analyzed to include the effects of corrosion, 

erosion, and fatigue under normal, off-normal, and accident conditions; and 

vi. How data, information, and evaluations are developed as a result of site-related 

investigations, studies of historical data, and any newly developed information

NUREG-1718 11.0-2
Plant Systems

addressing the geology, seismology, hydrology, meteorology, and geotechnical aspects 

of the site as well as site-proximity events considered as natural phenomena events 

(such as earthquakes, high winds, tornadoes, tornado missiles, and floods) and other 

external events (such as nearby transportation accidents, airplane crashes, and fires 

external to the facility) that may produce conditions that could influence the performance 

of plant facilities required to protect health and minimize danger to life or property. 

Because the ISA results identify the IROFS that form the safety functions discussed above, the 

primary reviewer should also review the ISA Summary (see SRP Chapter 5.0) to determine 

which plant systems have been identified as IROFS, their safety categories, their assumed 

operating modes and conditions, the impact of their inoperability, and any related limiting 

operations or plant mode restrictions. The review should also encompass any additional 

assumptions used in ISA qualitative/quantitative evaluations related to performance 

requirements for plant systems, such as redundancy, independence, reliability, quality, etc. 

11.4 ACCEPTANCE CRITERIA 

As part of the application for construction approval, the applicant should commit to providing 

plant systems that meet or exceed the acceptance criteria in the following subsections of this 

SRP section. 

11.4.1 Regulatory Requirements 

The staff's requirements applicable to all plant systems are the following: 

10 CFR Part 70.22, specifically relating to the requirement that the applicant is to provide a 

description of the equipment and facilities and propose procedures to protect health and 

minimize danger to life and property. 

10 CFR Part 70.23, specifically relating to the requirement that the Commission determine that 

the proposed equipment, facilities, and procedures are adequate to protect health and minimize 

danger to life and property. 

10 CFR Part 70.61 (e), specifically relating to the requirement that each engineered or 

administrative control or control system that is needed to meet the performance requirements 

be designated as an item relied on for safety and relating to the safety program that ensures 

each item relied on for safety will be available and reliable to perform its intended function when 

needed. 

10 CFR Part 70.62, specifically relating to the establishment and maintenance of a safety 

program and to the performance of an ISA. 

10 CFR Part 70.64, specifically relating to the application of BDC and defense-in-depth 

practices to new facilities or new processes at existing facilities.

11.0-3 NUREG-1 718
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11.4.2 Electrical Systems 

11.4.2.1 Regulatory Guidance 

Staff and industry documents that may provide useful background information for consideration 

in the design of MOX fuel fabrication facilities are listed in Section 11.7. 

11.4.2.2 Regulatory Acceptance Criteria 

The reviewer should find the applicant's electrical systems' design and operation acceptable if 

they satisfy the requirements listed in Section 11.4.1. The requirements and guidelines for 

electrical systems are those related to the BDC and defense-in-depth. The electrical systems' 

design and operation should fulfill the functional requirements determined from the ISA, and the 

electrical systems should be available and reliable to perform their intended safety function 

when needed. 

Typically, specific design considerations for electrical systems include two physically 

independent offsite power sources with redundant and independent onsite ac and dc power 

sources that should be designed with the following: 

A. Provisions so that components of the electrical systems can be tested periodically for 

operability and required functional performance; 

B. Electrical and physical separation to ensure that any required independence is maintained; 

C. No single failure vulnerability; 

D. Sufficient capacity and capability to ensure the IROFS supported by the electrical systems 

perform their intended functions; 

E. Adequate protective relaying and breaker control to ensure required functional performance 

and adequate response to electrical fault/overload conditions; 

F. Status monitoring of the behavior of the systems and components that are identified as 

IROFS; 

G. System capability to maintain functionality when subjected to tornadoes, tornado missiles, 

earthquakes, floods, and any other appropriate severe natural phenomena as established in 

the ISA.

NUREG-1718 11.0-4
Plant Systems

11.4.3 Instrumentation and Control Systems 

11.4.3.1 Regulatory Guidance 

Staff and industry documents that may provide useful background information for consideration 

in the design of MOX fuel fabrication facilities are listed in Section 11.7. 

11.4.3.2 Regulatory Acceptance Criteria 

The reviewer should find the applicant's instrumentation and control (W&C) systems' design and 

operation acceptable if they satisfy the requirements listed in Section 11.4.1. The requirements 

and guidelines for I&C systems are those related to the BDC and defense-in-depth. The I&C 

systems' design and operation should fulfill the functional requirements determined from the 

ISA, and the I&C systems should be available and reliable to perform their intended safety 

function when needed. 

Typically, specific design considerations for I&C systems include redundant and/or diverse 

instrument channels with coincident logic providing automatic actuation with additional manual 

operation capability. The instrument channels and associated logic should be designed with the 

following: 

A. Provisions so that I&C system components can be tested periodically for operability and 

required functional performance; 

B. Electrical, physical, and control/protection separation to ensure that any required 

redundancy and independence are maintained; 

C. No single failure vulnerability; 

D. Adequate instrument spans, setpoints, and control ranges to ensure proper monitoring and 

control of IROFS; 

E. Provisions so that I&C system components fail in a safe failure mode; 

F. Status monitoring of the behavior of the systems and components that are identified as 

IROFS; 

G. System capability to maintain functionality when subjected to tornadoes, tornado missiles, 

earthquakes, floods, and any other appropriate severe natural phenomena as established in 

the ISA.

11.0-5 NUREG-1 718
Plant Systems

11.4.4 Cooling Water System 

11.4.4.1 Regulatory Guidance 

None. 

11.4.4.2 Regulatory Acceptance Criteria 

The reviewer should find the applicant's cooling water system's design and operation 

acceptable if they satisfy the requirements listed in Section 11.4.1. For cooling water systems 

identified in the ISA as items relied on for safety, the requirements and guidelines are those 

related to the BDC and defense-in-depth. The cooling water system's design and operation 

should fulfill the functional requirements determined from the ISA, and the cooling water system 

should be available and reliable to perform its intended safety function when needed. 

Typically, specific design considerations for the cooling water system should demonstrate the 

following: 

A. Transfer of heat loads to an appropriate heat sink under normal, off-normal, and accident 

conditions; 

B. Adequate water supply under normal, off-normal, and accident conditions; 

C. Adequate component redundancy; the capability to isolate components, systems, or piping 

for maintaining system safety function under varying system configuration; and the 

capability of integrated system control; 

D. Supporting management measures (including tests and other verification methods) that 

ensure the structural integrity and system leak tightness (including the prevention of cross

contamination (radioactive and chemical)), the operability and adequate performance of 

active system components, and the capability of the system to perform required functions 

during normal and accident situations; 

E. Capability for withstanding environmental hazards resulting from pipeline breaks and 

dynamic effects associated with flow instability and attendant loads such as water hammer 

or cavitation and measures to prevent such dynamic conditions from occurring; 

F. Capacity and capability for detecting leaks and cross-contamination (radioactive and 

chemical), for inservice component inspection and system maintenance, and for operational 

functional testing of the system and its components; 

G. System capability to maintain functionality when subjected to tornadoes, tornado missiles, 

.earthquakes, floods, and any other appropriate severe natural phenomena as established in 

the ISA.

NUREG-1718 11.0-6
Plant Systems

11.4.5 Ventilation Systems 

11.4.5.1 Regulatory Guidance 

Regulatory guidance for the MOX fuel fabrication facilities for implementing and satisfying the 

regulatory requirements and acceptance criteria for ventilation systems is provided in Nuclear 

Regulatory Commission (NRC) Regulatory Guide 3.12, "General Design Guide for Ventilation 

Systems of Plutonium Processing and Fuel Fabrication Plants." 

Additional staff and industry documents that may provide useful background information for 

consideration in the design of MOX fuel fabrication facilities are listed in Section 11.7. 

11.4.5.2 Regulatory Acceptance Criteria 

The reviewer should find the applicant's ventilation systems' design and operation acceptable if 

they satisfy the requirements listed in Section 11.4.1, and the applicant has considered the 

relevant guidelines mentioned under Section 11.4.5.1. The requirements and guidelines for 

ventilation systems determined to be IROFS are those related to the BDC and defense-in

depth. The ventilation systems' design and operation should fulfill all of the functional 

requirements determined from the ISA, and the systems should be available with adequate 

reliability to perform all of their intended safety functions when needed. 

Typically, specific design considerations for ventilation systems should include the following: 

A. Confinement of radioactive contamination by zones and pressure differentials: 

i. Confinement of radioactive material is provided by multiple zones, with each zone 

bounded by barriers such as vessel, glovebox, building, and internal room walls. 

ii. The systems have the capability to direct ventilation air from areas of low radioactivity to 

areas of progressively higher radioactivity. Devices are provided to control and indicate 

pressure differentials between confinement zones. Alarms are provided to indicate 

when pressure differentials are not maintained in a prescribed range. 

iii. The systems have the capability to detect the need for isolation and to isolate portions of 

the systems relied on for safety in the event of failures or malfunctions elsewhere in the 

systems. The isolated systems have the capability to function under such conditions. 

iv. Supply air fans are interlocked with an exhaust air plenum pressure sensor to prevent 

supply fan operation unless the exhaust fans are running. This will prevent 

pressurization of any process room or area if exhaust ventilation fails. 

B. Test, calibration, and inservice surveillance capabilities: 

i. Provisions are made so that components of ventilation systems can be tested 

periodically for operability and required functional performance. Provisions include

11.0-7 NUREG-1718
Plant Systems

capability for periodic measurement of air flows in exhaust ducts and/or at equipment, 

hoods, and exhaust ducts. 

ii. The capability is also provided to test, under conditions as close to design as practical, 

the operating sequence that would bring ventilation systems into action, including the 

transfer to alternate power sources and the design airflow delivery capability. 

C. Redundancy of fans, dampers, and power supplies and no single failure vulnerability: 

i. There are two automatically operated isolation dampers in series to separate 

nonessential portions of the system from essential portions. 

ii. Essential components and subsystems are able to function in the event of loss of offsite 

power. In the event of failure of a single active component (equipment or control device) 

or loss of offsite power, the resulting systems flow capacity will not cause the loss of 

preferred direction of air flow from areas of low potential radioactivity to areas of higher 

potential radioactivity. 

iii. The systems are capable of automatically actuating components not operating under 

normal conditions or actuating standby components (redundant equipment) in the event 

of failure or malfunction, as needed. 

D. Sufficient capacity and capability: 

i. The heating and cooling functions of the ventilation systems are sufficient to maintain a 

suitable temperature range in the areas serviced, assuming proper performance of 

equipment contained in those areas. 

ii. Equipment identified as IROFS are capable of functioning under the worst anticipated 

ventilation systems' conditions. 

iii. The systems are capable of preventing the accumulation of flammable or explosive 

gases from processes within the facility. 

iv. The systems are capable of controlling airborne particulate material (dust) accumulation. 

v. Ventilation systems are capable of operating during a normal power outage at capacities 

required to maintain confinement of contaminants. 

E. Monitoring and alarms: 

L All exhausting ducts and stacks that may contain plutonium contaminants are provided 

with two monitoring systems: a continuous air monitoring system (CAMS) and a fixed 

sampler. The probes for sampling purposes are designed for isokinetic sampling and 

located to obtain representative samples. Each system is connected to an emergency 

power supply. The continuous stack sampler alerts cognizant personnel through an

NUREG-1718 11.0-8
Plant Systems

audible and visual annunciator if the airborne radioactive effluents reach prescribed 

limits. 

ii. Air monitoring and warning systems (including CAMS) are installed in areas where 

radioactive material is handled. Air sampling heads provide a representative of the 

potential airborne radioactivity being breathed. 

iii. Duct runs and flow distributors assure uniform representative air flow past monitoring 

and sampling stations as well as through filter installations. 

iv. Acceptance criteria for air monitoring and warning systems specific to radiation safety 

for design features, the radiation protection program, and effluent monitoring can be 

found in Sections 9.1.4.4.3(C), 9.2.4.5, and 10.4.3.B of this SRP. 

F. Environmental qualification: 

i. The ventilation systems, including detectors, monitoring systems, and controls, are 

qualified for all expected and credible severe environments in which the systems are 

expected to function. 

G. Design for natural phenomena: 

i. The ventilation systems are designed to maintain functionality when subjected to 

tornadoes, tornado missiles, earthquakes, floods, and any other appropriate severe 

natural phenomena as established in the ISA. 

ii. Design considerations are also made for protection from offsite releases of toxic 

chemicals as a result of natural phenomena, if appropriate. 

H. Fire protection and smoke control: 

i. The ventilation systems are designed to withstand any credible fire and explosion and 

continue to act as confinement barriers. 

ii. Ventilation systems are capable of operating during a fire in the areas they ventilate and 

safely handle products of combustion through appropriate ventilation channels. A 

supply air system remains operational; however, the option to discontinue air supply to 

the involved spaces is maintained. 

iii. The materials of construction for the ventilation systems are fire resistant to protect 

against fires occurring within or outside the systems. Approved smoke and heat 

detectors are provided in the system. 

I. Safe air supply to the control room and other occupied areas:

11.0-9 NUREG-1718
Plant Systems

L The ventilation systems confine and prevent uncontrolled release of radioactive 

aerosols, noxious fumes, and vapors into rooms and areas normally occupied by 

personnel. 

ii. Provisions are made for continuous monitoring of recirculated air to occupied areas and 

diversion of contaminated air to a once-through exhaust system if allowable radiation 

standards are exceeded. 

iii. The control room heating and cooling subsystems are capable of maintaining a suitable 

ambient temperature for control room personnel and equipment. 

iv. Portions of the system are isolated in the event of fires, failures, and malfunctions. 

v. The ventilation systems are capable of keeping essential equipment in the control room 

operational under the worst anticipated degraded conditions of the ventilation system. 

vi. The control room ventilation has an internal recirculation filtering mode or can discharge 

airborne contaminants from the control room area using a once-through ventilation 

mode, as applicable. 

J. Removal and replacement of filters and other expected maintenance designed to permit 

only minimum exposure of personnel to radioactivity: 

i. Ventilation systems allow for routine in-place testing of high efficiency particulate air 

(HEPA) filtration systems. 

ii. Potential doses from expected maintenance of ventilation systems can be minimized by 

providing ready access to the systems, by providing space to permit the activities to be 

accomplished expeditiously, by separating filter banks and components to reduce 

exposures to radiation from adjacent banks and components, and by providing sufficient 

space to accommodate auxiliary ventilation of shielding of components. 

K. Gloveboxes and process enclosures: 

i. Gloveboxes are constructed using high quality materials and workmanship to assure 

total containment and minimize leakage. Gloveboxes are constructed of 

noncombustible materials. (see Chapter 7.0 of this SRP). 

ii. The design of enclosures is based on downdraft ventilation flow to minimize the spread 

of fire. Heat detectors and combustible gas and vapor detection meters are provided on 

gloveboxes or enclosures where fire or explosion hazards exist. An inerting 

environment or automatic suppression are provided in these boxes or enclosures. 

Where automatic suppression is not provided, fire detectors are installed and manual 

fire suppression capability provided. (See Chapter 7.0 of this SRP.)

NUREG-1718 11.13-1 0
Plant Systems

iii. Small gloveboxes or enclosure systems supplied with gases under positive pressure 

have positive-acting pressure-relief devices (discharging into an exhaust system) to 

prevent overpressurization. Further, should these systems be recirculating, all 

necessary cleanup and detection equipment for noxious, corrosive, or explosive vapors 

or gases are considered. 

iv. The minimum instrumentation for a glovebox or enclosure ventilation system includes 

devices to indicate the pressure differential between the box or enclosure and the 

surrounding work area, the filter resistance, and the exhaust flow rate from the box or 

enclosure. (The applicant should specify the maximum operable pressure differential.) 

When box operations are not in full-time attendance for a continuous process, a sensor 

is provided to monitor abnormal pressure or temperature and alarm at a point where 

cognizant personnel are stationed. 

11.4.6 Civil-Structural Systems 

Civil-structural systems include the buildings and support structures of the facilities that are to 

house, support, confine, or contain the various other plant systems, components, and 

equipment associated with licensed nuclear materials or hazardous chemicals associated with 

licensed nuclear materials that may adversely affect I ROFS. 

11.4.6.1 Regulatory Guidance 

Industry standards that provide guidance for implementing and satisfying the regulatory 

requirements and acceptance criteria for civil-structural systems are: 

American Society of Civil Engineers (ASCE). ASCE-1 -82, "Guidelines for Design and Analysis 

of Nuclear Safety-Related Earth Structures." 

ASCE-4-86, "Seismic Analysis of Safety-Related Nuclear Structures." 

ASCE-7-95, "Minimum Design Loads for Buildings and Other Structures." 

For most of the structural materials that will be utilized in the civil-structural systems, there are 

existing design codes or standards that are based on using allowable stresses or on using a 

strength approach with load or resistance factors. The list of industry codes and standards that 

may have been used by the applicant is provided below: 

American Concrete Institute (ACI). 349-97, "Code Requirement for Nuclear Safety-Related 

Concrete Structures." 

American Institute of Steel Construction (AISC). N690-84, "Specification for the Design, 

Fabrication, and Erection of Steel Safety-Related Structures for Nuclear Facilities." 

Additional staff and industry documents that may provide useful background information for 

consideration in the design of MOX fuel fabrication facilities are listed in Section 11.7.

11.0-11 NUREG-1718
Plant Systems

The governing building code should be considered as a guidance document in that it will 

prescribe absolute minimum parameters for the civil-structural systems independent of the 

facility requirements resulting from the ISA. Embedded within the building codes or other 

standards and documents that are incorporated by reference, there can be guidance regarding 

the design, analysis, construction, and testing portions of all of the elements for consideration 

described above. Listed below are the major national building codes that may become a single 

building code in 2000. One of these documents, or a local building code, will govern as the 

minimum requirement for all civil-structural systems at the facility. These building codes are 

listed below: 

BOCA, Building Officials and Code Administrators International, Inc. 

SBC, Southern Building Code Congress International, Inc. 

UBC, International Conference of Building Officials 

IBC, International Code Council, Inc. (to release the International Building Code 2000 that will 

replace the three major U.S. building codes) 

11.4.6.2 Regulatory Acceptance Criteria 

The reviewer should find the applicant's civil-structural systems' design and operation 

acceptable if they satisfy the requirements listed in Section 11.4.1, and the applicant has 

considered the relevant guidelines mentioned in Section 11.4.6.1. The requirements and 

guidelines for the civil-structural systems are those related to the BDC and defense-in-depth. 

The civil-structural systems' design and operation should fulfill all of the functional requirements 

determined from the ISA, and the systems should be available with adequate reliability to 

perform all of their intended safety functions when needed. 

Typically, specific design considerations for civil-structural systems should include the following: 

A. Site and environmental design conditions: 

The parameters defining the site and environmental conditions should be identified along 

with the magnitude or range of values that are to form part of the basis of the design for the 

structural systems that are IROFS based on the ISA results. The civil-structural systems to 

which these parameters are to be applied should be those that are involved in the 

prevention or mitigation of the consequences of any of the events, accidents, or conditions 

of operation that are identified as a result of the ISA and are therefore IROFS. These 

civil-structural systems would also include those that could affect systems and components 

that are IROFS. The basis; e.g., field observations and data, test data, analytical results, or 

other sources; for the parameters and the values being used should be provided. Some of 

the parameters may be deterministic while others may result from the ISA or other 

nondeterministic studies. For example, the return period or frequency of a specific site or 

environment-related design parameter should be reflected in, or derived from, the ISA.

NUREG-1718 11.0-12
Plant Systems

B. Analytical models and material properties: 

The analytical models that are used to represent the IROFS in the civil-structural systems in 

the analysis, evaluation, and design of those systems should be identified. The degree of 

sophistication and the level of detail in the models should reflect the ISA results in 

identifying the relative importance of the various civil-structural systems. Where several 

levels of model detail are used, the linkage and logic used to obtain reliable results should 

be identified. The bases of the specific models utilized should be provided. The material 

properties of the individual elements in the civil-structural systems should be provided, 

including the bases for the properties and the methods that will be used to assure that the 

range of the future as-built properties are adequately represented in the analysis, evaluation 

and design process. 

C. Civil-structural systems classification and loading combinations: 

Since many of the input parameters may occur simultaneously, there should be identifiable 

combinations that link these parameters in the form of the loading combinations defined for 

the facility and the various classes of civil-structural systems. These loading combinations 

should be linked to the ISA results. The resulting load combinations should be clearly 

identified as representing the unique set of loading functions for the facility at the site for the 

various classes of civil-structural systems. These form part of the design bases of the 

facility. In addition, the reviewer should verify that the minimum requirements of the 

governing building code for the facility have been incorporated into the design bases. The 

reviewer should verify the acceptability of the loading combinations and the classifications 

are consistent with the ISA results. 

D. Design of civil-structural elements and systems: 

The reviewer should verify that the application provides the bases for sizing the various 

structural elements and members of the civil-structural systems. This aspect of the design 

basis can be used to quantify the safety margins that may be provided based on the loads 

and load combinations identified as a result of the ISA for the various natural hazards, 

operational, and accident event scenarios. The reviewer should ascertain that these safety 

margins are consistent with the analysis in Chapter 5.0 of this SRP relative to the ISA that 

was performed. 

E. Evaluation and verification: 

The reviewer should that ensure that the civil-structural systems are adequately designed to 

maintain functionality when subjected to tornadoes, tornado missiles, earthquakes, floods, 

and any other appropriate severe natural phenomena as established in the ISA. 

Overall, the reviewer must ensure that all the relevant parameters have been incorporated into 

the design, and that the design reflects the supplemental design bases of the other plant 

systems as well as the requirements of facility operations and the BDC. The reviewer should

11.0-1 3 NUREG-1718
Plant Systems

ascertain that the final design as reflected in the construction documents has undergone a final 

structural evaluation, with the results being provided to verify that the ISA results are correct. 

11.4.7 Material Transport System (Pumps and Valves) 

11.4.7.1 Regulatory Guidance 

Staff and industry documents that may provide useful background information for consideration 

in the design of MOX fuel fabrication facilities are listed in Section 11.7. 

11.4.7.2 Regulatory Acceptance Criteria 

The reviewer should find the design and operation of the material transport systems identified 

as IROFS acceptable if they satisfy the requirements listed in Section 11.4.1. The 

requirements and guidelines for such systems are those related to the BDC and defense-in

depth. The material transport system's design and operation should fulfill all of the functional 

requirements determined from the ISA, and the system should be available with adequate 

reliability to perform all of its intended safety functions when needed. 

Typically, specific design considerations for the material transport system should demonstrate 

the following: 

A. Adequate capacity exists to handle the expected volume of radioactive material during 

normal operating and accident conditions. 

B. There is redundancy or diversity of components required to prevent the release of 

radioactive materials to the environment or needed for the safe operation of the material 

transport system. 

C. The material transport system can be safely shut down during normal operations and 

accident conditions. Provisions for emergency power are included for critical process 

components. 

D. Tank and piping systems are of welded construction to the fullest extent possible. 

E. Tank and piping systems are designed to take advantage of gravity flow to reduce the 

potential for contamination associated with pumping and pressurization. 

F. The design of the material transport system assures that accidental criticality will not occur 

under normal operating conditions or under credible accident conditions. 

G. All system components expected to be in contact with strong acids or caustics are corrosion 

resistant. 

H. Use of traps is avoided, and the piping is designed to minimize entrapment and buildup of 

solids in the system.

NUREG-1718 11.0-1 4
Plant Systems

I. Systems and devices are evaluated to determine the need for hoods, gloveboxes, and 

shielding for personnel protection. Generally, wet processing operations involving gram 

quantities of plutonium and operations involving 50 micrograms or more of plutonium in 

respirable form are conducted in a glovebox. (See Chapter 9.0 of this SRP.) 

J. Surface finishes in the work area are of materials that have satisfactory decontamination 

characteristics for their particular application. 

K. Material transport systems are adequately designed to maintain functionality when 

subjected to tornadoes, tornado missiles, earthquakes, floods, and any other appropriate 

severe natural phenomena as established in the ISA. 

11.4.8 Heavy Lift Cranes 

11.4.8.1 Regulatory Guidance 

Staff and industry documents that may provide useful background information for consideration 

in the design of MOX fuel fabrication facilities are listed in Section 11.7. 

11.4.8.2 Regulatory Acceptance Criteria 

The reviewer should find the design and operation of cranes for lifting heavy loads acceptable if 

they satisfy the requirements listed in Section 11.4.1. The requirements and guidelines for 

heavy lift cranes identified as IROFS are those related to the BDC and defense-in-depth. The 

design and operation of heavy lift cranes should fulfill all of the functional requirements 

determined from the ISA, and the heavy lift cranes should be available with adequate reliability 

to perform all of their intended safety functions when needed. 

Typically, specific design considerations for heavy lift cranes should demonstrate the following: 

A. The handling equipment is designed in accordance with the American National Standard for 

Overhead and Gantry Cranes (Top Running Bridge, Single or Multiple Girder, Top Running 

Hoist), American National Standards Institute/American Society of Mechanical Engineers 

(ANSI/ASME) ANSI/ASME-B30.2-1983 and the American National Standard for Overhead 

Hoists, ANSI/ASME-B30.16-1987. 

B. The purchase of equipment and materials is based on the codes and standards which 

represent a level of capability to meet the design requirements specified in American 

National Standard Lightning Protection Code, American National Standards 

Institute/National Fire Protection Association (ANSI/NFPA) ANSI/NFPA-78-1986, and the 

Specifications for Overhead Traveling Cranes, Crane Manufacturers Association of America 

(CMAA) Specification 70. 

C. Cranes capable of carrying heavy loads are prevented, preferably by design rather than by 

interlocks, from moving over safety and containment systems.

11.0-15 NUREG-1718
Plant Systems

D. Cranes are designed to provide single failure-proof handling of heavy loads, so that a single 

failure will not result in loss of capability of the crane-handling system to perform its safety 

function. 

E. The crane structures and their support equipment are designed to withstand all design 

loads while remaining in place. 

F. The crane system design is based on an analysis that considers the confinement of 

radioactive material under conditions of system failure and misoperation. 

G. Heavy lift cranes are adequately designed to maintain functionality when subjected to 

tornadoes, tornado missiles, earthquakes, floods, and any other appropriate severe natural 

phenomena as established in the ISA. 

11.5 REVIEW PROCEDURES 

11.5.1 Acceptance Review 

The primary reviewer should perform an acceptance review to determine if the application 

adequately addresses the items in Section 11.3, "Areas of Review," for either the construction 

approval review or the review for a license to possess and use special nuclear material (SNM). 

Guidance specific to the construction approval review and the review for a license to possess 

and use SNM is provided below: 

A. Construction Approval 

Specifically, the material submitted for the construction approval review should contain the 

applicant's commitments to provide plant systems that satisfy the acceptance criteria in 

Section 11.4 and should also address the layout and design of the plant systems, their 

components, and any related information considering the current stage of the applicant's 

design process. 

B. License To Possess and Use SNM 

Specifically, the application for a license to possess and use SNM should address the items 

described in Section 11.3 in full and update the information submitted for the construction 

approval to encompass design modifications and any other system features not adequately 

described during the construction approval review. 

If the primary reviewer verifies that plant systems are adequately addressed, the primary 

reviewer should accept the application for the safety evaluation in Section 11.5.2. If the primary 

reviewer identifies significant deficiencies in the material provided, the primary reviewer should

NUREG-1718 11.0-16
Plant Systems

request that the applicant submit additional information prior to the start of the safety 

evaluation. 

The secondary and supporting reviewers should confirm that the described plant systems are 

consistent with descriptions in other sections of the application. Information provided for plant 

systems should be of comparable quality and detail and should not contradict or adversely 

impact information contained in other sections of the application. 

11.5.2 Safety Evaluation 

After determining that the application is acceptable for review in accordance with either 

Section 11.5.1 (A) (construction approval review) or Section 11.5.1 (B) (review for a license to 

possess and use SNM), the primary reviewer should perform a safety evaluation against the 

acceptance criteria described in Section 11.4. On the basis of its review, the staff may request 

that the applicant provide additional information or modify the application to meet those 

acceptance criteria. 

Guidance specific to the construction approval review and the review for a license to possess 

and use SNM is provided below. 

A. Construction Approval 

The primary reviewer should verify the applicant's commitment to providing plant systems 

that meet or exceed the acceptance criteria in Section 11.4. The primary reviewer should 

focus on the layout and design of the plant systems, their components, and any related 

information considering the current stage of the applicant's design process. 

B. License To Possess and Use SNM 

The primary reviewer should establish that the applicant's plant systems' designs and 

operations provide reasonable assurance that the plant systems satisfy the acceptance 

criteria in Section 11.4 and will be available and reliable to perform their intended safety 

functions when needed. The primary reviewer also should ensure that adequate 

documentation is provided in the ISA Summary for all plant systems that are identified as 

IROFS. 

Secondary and supporting reviewers should confirm that the provisions made in the 

application for plant systems are in accordance and consistent with their specified sections 

of the SRP. For example, the review performed by the primary reviewer of Chapter 15.0 of 

this SRP-as a supporting reviewer-should encompass the adequacy of management 

measures applied to plant systems. The reviewer of radiation safety under Chapter 9.0 

should evaluate the design and operation of plant systems, such as the ventilation systems 

and certain instrumentation and controls, with regard to adequate radiation protection. The 

reviewer of human factors under Chapter 12.0 should confirm that the principles of human 

factors engineering are applied to the instrumentation and control systems' design. Also, 

the primary reviewer of Chapter 5.0 should determine the adequacy of IROFS (including

11.0-1 7 NUREG-1 718
Plant Systems

plant systems) to assure that the likelihood and consequences of identified accidents meet 

the performance requirements of 10 CFR 70.61. 

For an existing facility being reviewed for a license amendment or renewal, the NRC reviewers 

may wish to visit the site and facility personnel to gain a better understanding of the 

represented plant systems and their intended safety functions. For a planned facility, the NRC 

reviewers may wish to meet with the design team to gain a better understanding of the process, 

its potential hazards, and safety approaches. 

When the safety evaluation is complete, the primary reviewer-with assistance from the other 

reviewers-should prepare the plant systems input for the Safety Evaluation Report (SER) as 

described in Section 11.6 using the acceptance criteria from Section 11.4. The secondary 

reviewers should coordinate the plant systems input with the balance of the reviews and the 

SER. 

11.6 EVALUATION FINDINGS 

The primary reviewer should document the safety evaluation by preparing material suitable for 

inclusion in the SER. The primary reviewer should describe the review, explain the basis for the 

findings, and state the conclusions. 

The staff could document the safety evaluation for the construction approval review by stating 

that the applicant has committed to providing plant systems that meet or exceed the 

acceptance criteria in Section 11.4. 

The staff could document a safety evaluation for the license application for operations as 

follows: 

The staff evaluated [insert a summary statement of what was evaluated and why the 

reviewer finds the submittal acceptable]. Based on the review of the license application, the 

NRC staff concluded that the applicant's plant systems' designs and operations satisfy the 

staff's acceptance criteria and are adequately available and reliable to perform their 

intended safety functions when needed. In doing so, the applicant has satisfactorily 

addressed the applicable regulatory requirements, including the performance requirements, 

the baseline design criteria, and the defense-in-depth practices contained in 

10 CFR Part 70. 

11.7 REFERENCES 

American Concrete Institute (ACI). 349-97, "Code Requirement for Nuclear Safety-Related 

Concrete Structures." 

American Institute of Steel Construction (AISC). N690-84, "Specification for the Design, 

Fabrication, and Erection of Steel Safety-Related Structures for Nuclear Facilities."

NUREG-1718 11.0-1 8
Plant Systems

* American National Standards Institute/American Nuclear Society (ANSI/ANS). 

ANSI/ANS-2.8-1992, "Determining Design Basis Flooding at Power Reactor Sites." 

* ANSI/ANS-2.11-1978(R1 989), "Guidelines for Evaluating Site-Related Geotechnical 

Parameters at Nuclear Power Plant Sites." 

* ANSI/ANS-2.19-1981 (R1 990), "Guidelines for Establishing Site-Related Parameters 

for Site Selection and Design of an Independent Spent Fuel Storage Installation (Water Pool 

Type)." 

American National Standards Institute/American Society of Mechanical Engineers 

(ANSI/ASME). ANSI/ASME-B30.2-1983, "Overhead and Gantry Cranes (Top Running Bridge, 

Single or Multiple Girder, Top Running Hoist)." 

. ANSI/ASME-B30.16-1987, "Overhead Hoists." 

ANSI/ASME-N509-1980, "Testing of Nuclear Air-Cleaning Systems." 

. ANSI/ASME-N510-1980, "Nuclear Power Plant Air Cleaning Units and Components." 

American National Standards Institute/National Fire Protection Association (ANSI/(NFPA). 

ANSI/NFPA-780-1997, "Lightning Protection." 

American Society of Civil Engineers (ASCE). ASCE-1-82, "Guidelines for Design and Analysis 

of Nuclear Safety-Related Earth Structures." 

ASCE-4-86, "Seismic Analysis of Safety-Related Nuclear Structures." 

ASCE-7-95, "Minimum Design Loads for Buildings and Other Structures." 

Code of Federal Regulations, Title 29, Labor, Chapter XVII, "Occupational Safety and Health 

Administration, Department of Labor." 

Crane Manufacturers Association of America (CMAA). Specification 70, "Specifications for 

Electric Overhead Traveling Cranes." 

Department of Energy (U.S.) (DOE). Standard 1128-98, "Guide of Good Practices for 

Occupational Radiological Protection in Plutonium Facilities," Section C.4.3.1, "Piping and 

Valves." 

Institute of Electrical and Electronics Engineers (IEEE) Nuclear Power Standards Collection. 

IEEE: Piscataway, New Jersey. 

Nuclear Regulatory Commission (U.S.) (NRC). "Domestic Licensing of Special Nuclear 

Material (10 CFR Part 70)," Federal Register: Vol. 64, No. 146. pp. 41338-41357. July 30, 

1999.

11.0-1 9 NUREG-1718
Plant Systems

* NUREG-0800, Standard Review Plan, Chapter 7, "Instrumentation and Controls." 

NRC: Washington, D.C. 

NUREG-0800, Standard Review Plan, Chapter 8, "Electric Power," Table 8-1, 

Acceptance Criterial and Guidelines for Electric Power Systems. NRC: Washington, D.C. 

• NUREG-0800, Standard Review Plan, Chapter 9, "Auxiliary Systems." NRC: 

Washington, D.C. 

Regulatory Guides for Division 1, Power Reactors. 

Regulatory Guide 8.8, Revision 3, "Information Relevant to Ensuring that Occupational 

Exposures at Nuclear Power Stations will be as Low as is Reasonably Achievable." NRC: 

Washington D.C. June 1978.

NUREG-1718 11.0-20
12.0 HUMAN FACTORS ENGINEERING FOR PERSONNEL ACTIVITIES 

12.1 PURPOSE OF REVIEW 

The purpose of this review is to establish that human factors engineering (HFE) is applied to 

personnel activities identified as safety-significant, consistent with the findings of the integrated 

safety analysis (ISA), and the determination of whether an item relied on for safety has special 

or unique safety significance. A graded approach commensurate with the complexity and 

integration and operation of the control systems is appropriate. The application of HFE to 

personnel activities ensures that the potential for human error in the facility operations was 

addressed during the design of the facility by facilitating correct, and inhibiting wrong, decisions 

by personnel and by providing means for detecting and correcting or compensating for error. 

For the purposes of this chapter, the phrase "personnel activities" represents personnel 

activities identified as items relied on for safety (IROFS) and personnel activities that support 

safety, such as maintenance. 

12.2 RESPONSIBILITY FOR REVIEW 

Primary: Human Factors Specialist 

Secondary: ISA Reviewer 

Primary Reviewer of SRP Section 15.4, "Training and Qualification 

of Plant Personnel" 

Primary Reviewer of SRP Section 15.5, "Plant Procedures" 

Instrumentation and Control (I&C) Reviewer 

Supporting: Fuel Cycle Facility Inspector 

12.3 AREAS OF REVIEW 

The mixed oxide (MOX) facility relies heavily on automated systems employing advanced digital 

instrumentation and control technology. These systems may be complex, with potential 

negative impacts on human performance activities in both operations and maintenance. The 

scope of review.for the HFE for personnel activities should be consistent with the results of the 

ISA and include, as appropriate: 

A. A description of the safety-significant personnel actions, the associated human systems 

interfaces (HSIs), and the consequences of incorrectly performing or omitting actions for 

each personnel activity. 

B. The applicant's plans for HFE design review, including the: 

i. Goals and scope; 

ii. Team composition, organizational authority, and responsibilities;

12.0-1 NUREG-1718
Human Factors Engineering for Personnel Activities

iii. Process and procedures; 

iv. Issues tracking; and 

v. Functional description. 

C. Operating experience review; 

D. Function and task analysis; 

E. HSI design, inventory, and characterization; 

F. Staffing; 

G. Procedure development; 

H. Training program development; and 

I. Human factors verification and validation (V&V). 

All nine areas of review (A-I) may not be necessary for a specific application. Areas of review 

should be based on the applicant's provisions to address personnel activities consistent with the 

ISA findings; the similarity of the associated HFE issues for similar type plants; and the 

determination of whether an item relied on for safety has special or unique safety significance. 

12.4 ACCEPTANCE CRITERIA 

12.4.1 Regulatory Requirements 

Regulatory requirements for HFE for personnel activities are: 

A. 10 CFR 70.61(e), which requires a safety program to ensure that each item relied on for 

safety will be available and reliable to perform its intended function when needed. 

B. 10 CFR 70.64(b)(2), which requires features that enhance safety by reducing challenges to 

IROFS. 

12.4.2 Regulatory Guidance 

Staff and industry documents that may provide useful background information for consideration 

in applying HFE to MOX fuel fabrication facilities are listed in Section 12.7. 

12.4.3 Regulatory Acceptance Criteria 

The HFE for personnel actions should be acceptable if: 

A. Identification of Personnel Activities 

NUREG-1718 12.0-2
Human Factors Engineering for Personnel Activities

The applicant appropriately identified the personnel activities such that the reviewer can 

understand the actions, the HSIs involved, and the consequences. 

B. HFE Design Review Planning 

The applicant's approach for planning HFE design review includes: 

i. Identification of appropriate goals and scope to ensure that HFE practices and 

guidelines are implemented during design, construction, and operation of the facility. 

ii. Implementation by an HFE team that has the appropriate composition, experience, and 

organizational authority to ensure that HFE is considered in the design of HSI for 

personnel activities. The HFE team's responsibilities include ensuring the proper 

development, execution, oversight, and documentation of the HFE function. Depending 

on the identification of personnel activities, it may be appropriate for the HFE team to 

consist of a single individual. 

iii. An HFE team that attains the HFE goals and scope through established processes and 

procedures and that tracks HFE issues. 

iv. An HFE function that ensures that all aspects of the personnel activities including the 

HSI are developed, designed, and evaluated on the basis of a structured approach 

using HFE. 

C. Operating Experience Review (OER) 

The applicant identified safety-related HFE events or potential events that have occurred in 

existing facilities that are similar to the proposed facility. The applicant: 

i. Reviewed the HFE-related events or potential events for relevance; 

ii. Analyzed the HSI technology employed for the relevant HFE events or potential events; 

and 

iii. Conducted (or reviewed existing) operator interviews and surveys on the HSI technology 

for the relevant HFE events or potential events. 

D. Functional Allocation Analysis and Task Analysis 

i. Functional allocation analysis: The functional allocation analysis is based on the OER. 

Personnel activities are functionally allocated to take advantage of human strengths and 

to avoid demands that are not compatible with human capabilities. 

ii. Task analysis: The task analysis includes the task analysis scope, identification and 

analysis of critical tasks; detailed description of personnel demands (e.g., input,

12.0-3 NUREG-1718
Human Factors Engineering for Personnel Activities

processing, and output); iterative nature of the analysis; and incorporation of job design 

issues. The task analysis addresses each operating mode for each personnel activity 

(e.g., startup, normal operations, emergency operations, and shutdown). The task 

analysis results support the functional allocation. 

E. HSI Design, Inventory, and Characterization 

The HSI design incorporates the functional allocation analysis and task analysis into the 

detailed design of safety-significant HSI components (e.g., alarms, displays, controls, and 

operator aids) through the systematic application of HFE. The HSI design includes the 

overall work environment, the work space layout (e.g., control room and remote shutdown 

facility layouts), the control panel and console design, the control and display device layout, 

and information and control interface design details. The HSI design process ensures the 

application of HFE to the HSI required to perform personnel activities. The HSI design 

process excludes the development of extraneous controls and displays. The HSI design 

documentation includes a complete HSI inventory and the basis for the HSI 

characterization. 

F. Staffing 

Staffing is based on a review of the number and qualifications of personnel for each 

personnel activity during all plant operating conditions. The applicant conducts this review 

in a systematic manner that incorporates the functional allocation and task analysis results. 

Categories of personnel are based on the types of personnel activities. Staffing 

considerations include issues identified in the OER, functional allocation, HSI design, 

procedure development, and V&V. 

G. Procedure Development 

The applicant's procedure development for personnel activities incorporates HFE principles 

and criteria, along with all other design requirements, to develop procedures that are 

technically accurate, comprehensive, explicit, easy to utilize, and validated consistent with 

the acceptance criteria in Section 15.5.4 of this SRP. Because procedures are considered 

an essential component of the HSI design, they are derived from the same design process 

and analyses as the other components of the HSI (for example, displays, controls, operator 

aids) and subject to the same evaluation processes. Procedures include, as needed to 

support the personnel activity: generic technical guidance, plant and system operations, 

abnormal and emergency operations, tests (for example, preoperational, startup, and 

surveillance), and alarm response. 

H. Training Program Development 

The applicant's training program development addresses all personnel activities. The 

training program development indicates how the knowledge and skill requirements of 

personnel will be evaluated, how the training program development is coordinated with the 

other activities of the HFE design process, and how the training program will be

NUREG-1718 12.0-4
Human Factors Engineering for Personnel Activities

implemented in an effective manner consistent with human factors principles and practices. 

The training program development should address the areas of review and acceptance 

criteria described in Section 15.4.4 of this SRP and should result in a training program that 

provides personnel with the qualifications commensurate with the personnel activities. 

I. Verification&Validation 

V&V confirms that the design incorporates HFE to HSI in a manner that enables the 

successful completion of personnel activities. The V&V should be applied to personnel 

activities (see Item A) and HSI design (see Item E). The V&V process should consist of the 

following: 

L HSI task support verification: HSI components are appropriately provided for personnel 

activities through HSI task support verification. The verification shows that each HSI 

identified the task analysis (see Item D(ii)) and that the HSI design (see Item E) is 

appropriately provided, yet minimizes the incorporation of information, displays, controls, 

and decorative features that unnecessarily complicate personnel activities. 

ii. HFE design verification: The HFE design verification shows that each HSI identified for 

a personnel activity incorporated HFE into the design. Deviations from accepted HFE 

principles and guidelines should be justified or documented for resolution/correction. If 

all HSI components are not addressed by HFE design verification, then an alternative 

multidimensional sampling methodology should be used to assure comprehensive 

consideration of the safety significance of HSI components. The sample size should be 

sufficient to identify a range of significant safety issues. 

iii. Integrated system validation: The applicant commits to a performance-based evaluation 

of the integrated design to ensure that the HFE/HSI supports safe operation of the plant. 

Integrated system validation is performed after HFE problems identified in HFE design 

activities are resolved or corrected because these may negatively affect performance 

and, therefore, validation results. Validation is performed by evaluating personnel 

activities using appropriate measurement tools. All personnel activities should be tested 

and found to be adequately supported in the design, including personnel activities 

outside the control room. 

iv. Human factors issue resolution verification: The applicant verifies that HFE issues 

identified during the design process were addressed and resolved. Issue resolution 

verification should be documented in the HFE issue tracking system established by the 

HFE team (see Item B). Issues that cannot be resolved until the HSI design is 

constructed, installed, and tested should be identified and incorporated into the final 

HFE/HSI design verification. 

v. Final HFE/HSI design verification: The applicant should commit to performing a final 

HFE/HSI design verification if the applicant cannot demonstrate that it has fully 

evaluated the actual installation of the final HSI design in the plant through the V&V 

activities described above. Final HFE/HSI design verification should demonstrate that

12.0-5 NUREG-1718
Human Factors Engineering for Personnel Activities

in-plant HFE design implementation conforms to the HFE design (see Item E) as 

modified V&V activities. 

V&V activities should be performed in the order listed above, as necessary. However, the 

applicant may find that it is necessary to iterate in order to address design corrections and 

modifications that occur during V&V. 

12.5 REVIEW PROCEDURES 

12.5.1 Acceptance Review 

The primary reviewer should perform an acceptance review to determine if the application 

adequately addresses the items in Section 12.3, "Areas of Review," for either the construction 

approval review or the review for a license to possess and use special nuclear material (SNM). 

Guidance specific to the construction approval review and the review for a license to possess 

and use SNM is provided below. 

A. Construction Approval 

Specifically, the safety assessment of the design basis should address Section 12.3(A)-(E) 

consistent with the level of design and the consequences of incorrectly performing the 

personnel activity consistent with the safety assessment of the design basis. Where 

information is under development or not yet available, the applicant may use a commitment 

to providing the material with the license application in lieu of the actual material. 

B. License To Possess and Use SNM 

Specifically, the safety assessment of the license application should fully address 

Section 12.3(A)-(I) consistent with the consequences of incorrectly performing the 

personnel activity. 

If the primary reviewer verifies that the HFE for personnel activities is adequately addressed in 

either the construction approval review or the review for the license to possess and use SNM, 

the primary reviewer should accept the application for the safety evaluation in Section 12.5.2. If 

the primary reviewer identifies significant deficiencies in the material provided, the primary 

reviewer should request that the applicant submit additional information prior to the start of the 

safety evaluation. 

12.5.2 Safety Evaluation 

After determining that the application is acceptable for review in accordance with either 

Section 12.5.1 .A (construction approval review) or 12.5.1 .B (review for a license to possess and 

use SNM), the primary reviewer should perform a safety evaluation against the acceptance 

criteria described in Section 12.4. On the basis of its review, the staff may request that the

NUREG-1 718 12.0-6
Human Factors Engineering for Personnel Activities

applicant provide additional information or modify the application to meet those acceptance 

criteria. 

The primary reviewer should use a tiered approach for evaluating HFE for personnel activities. 

The upper tier is the program description level, such as missions or goals. The middle tier is 

when functions are allocated to tasks (personnel activities) for the purposes of specifying the 

alarms, information, and controls. The tasks are arranged into meaningful jobs and the HSI 

should be designed to best support job task performance. The lower tier is the detailed design 

(of the HSI, procedures, and training) and how they are incorporated into the facility design. 

Evaluation of the HFE design should be broad-based and include aspects of normal and 

emergency operations, testing, maintenance, etc., consistent with findings in the safety 

assessment of the design basis (application for construction approval) or in the ISA Summary 

(license application for operations). 

Guidance specific to the application for construction approval and the license application for 

operations is provided below. 

A. Construction Approval 

In general, the primary reviewer should perform an upper-tier review for the safety 

assessment of the design basis. As the level of design permits, the primary reviewer should 

perform a middle-tier review on those personnel activities that are identified as preventing or 

mitigating accident consequences. 

B. License To Possess and Use SNM 

In general, the primary reviewer should perform a lower-tier review for personnel activities 

that prevent or mitigate "high-consequence" events, a middle-tier review for personnel 

activities that prevent or mitigate "intermediate-consequence" events, and a high-level 

review for any remaining HFE activities. 

The primary reviewer should review the ISA Summary to ensure that personnel activities 

have been suitably characterized as IROFS. The extent to which HFE elements are applied 

should be based on the number, type, complexity, and potential consequences of the 

personnel activities. 

The secondary reviewer should ensure that the types of personnel activities relied on for 

safety are appropriate. The primary reviewer should coordinate with the I&C reviewer for 

Chapter 11.0, "Plant Systems," to confirm that HFE principles are appropriately addressed 

in the I&C approach. 

The supporting reviewers should assist in the tiered approach of the review so they may 

look at more specific examples of human factors engineering application. 

When the safety evaluation is complete, the primary reviewer, with assistance from the other 

reviewers, should prepare the HFE input for the Safety Evaluation Report (SER), as described

12.0-7 NUREG-1718
Human Factors Engineering for Personnel Activities

in Section 12.6 using the acceptance criteria from Section 12.4. The secondary reviewer 

should coordinate the chemical safety input with the balance of the reviews and the SER. 

12.6 EVALUATION FINDINGS 

The primary reviewer should document the safety evaluation by preparing material suitable for 

inclusion in the SER. The primary reviewer should describe the review, explain the basis for the 

findings, and state the conclusions. 

The staff could document the safety evaluation for the construction approval review as follows: 

The staff reviewed the application of human factors engineering (HFE) to personnel 

activities for the application for construction approval for [insert facility name] according to 

Chapter 12.0 of NUREG-1718. The staff evaluated [insert a summary statement of the 

evaluation] and found [insert a summary of the findings]. 

The staff concluded that the applicant has established an adequate design basis, as it 

relates to HFE, that meets the requirements for construction approval in 10 CFR Part 70. 

The staff could document the safety evaluation for the review for a license to possess and use 

SNM as follows: 

The staff reviewed the application of human factors engineering (HFE) to personnel 

activities for the license application to possess and use SNM at [insert facility name] 

according to Chapter 12.0 of NUREG-1718. The staff evaluated [insert a summary 

statement of the evaluation] and found [insert a summary of the findings]. 

The staff concluded that the applicant applied HFE to personnel activities identified as items 

relied on for safety, consistent with the results of the ISA, and that its personnel activities 

meet the requirements associated with human factors given in 10 CFR Part 70. 

12.7 REFERENCES 

Code of Federal Regulations, Title 10, Energy, Part 70, "Domestic Licensing of Special Nuclear 

Material." 

Department of Defense (U.S.) (DOD). MIL-STD-1472D, "Human Engineering Design Criteria 

for Military Systems, Equipment and Facilities." DOD: Washington, D.C. 

Nuclear Regulatory Commission, (U.S.) (NRC). NUREG-0700, Rev.1, Vol.1-3, "Human

System Interface Design Review Guideline." NRC: Washington, D.C. 1996. 

NUREG-071 1, "Human Factors Engineering Program Review Model." NRC: 

Washington, D.C. 1994.

NUREG-1718 12.0-8
Human Factors Engineering for Personnel Activities

* NUREG/CR-6633, "Advanced Information Systems Design: Technical Basis and 

Human Factors Review Guidance." NRC: Washington, D.C. March 2000. 

* NUREG/CR-6634, "Computer-Based Procedure Systems: Technical Basis and Human 

Factors Review Guidance." NRC: Washington, D.C. March 2000. 

- NUREG/CR-6635, "Soft Controls: Technical Basis and Human Factors Review 

Guidance." NRC: Washington, D.C. March 2000. 

NUREG/CR-6636, "Maintainability of Digital Systems: Technical Basis and Human 

Factors Review Guidance." NRC: Washington, D.C. March 2000. 

• NUREG/CR-6637, "Human Systems Interface and Plant Modification Process: 

Technical Basis and Human Factors Review Guidance." NRC: Washington, D.C. March 2000. 

Nuclear Regulatory Commission (U.S.), Washington, D.C. "Domestic Licensing of Special 

Nuclear Material (10 CFR Part 70)." Federal Register: Vol. 64, No. 146. pp. 41338-41357. 

July 30, 1999.

12.0-9 NUREG-1718
13.0 SAFEGUARDS 

13.1 PHYSICAL PROTECTION 

13.1.1 PURPOSE OF REVIEW 

The purpose of this review is to determine with reasonable assurance that the applicant has 

committed to having a physical protection system that provides high assurance that activities 

involving special nuclear material (SNM) are not inimical to the common defense and security 

and do not constitute an unreasonable risk to the public health and safety. The physical 

protection system should be designed to protect against the design basis threats of theft or 

diversion of formula quantities of strategic special nuclear material (SSNM) and radiological 

sabotage as stated in 10 CFR 73.1 (a). Physical protection requirements for applicants 

possessing formula quantities of SSNM are found in 10 CFR 73.20, 73.45, and 73.46. 

13.1.2 RESPONSIBILITY FOR REVIEW 

Primary: Physical Protection Specialist 

Secondary: None 

Supporting: Regional Physical Protection Inspector 

13.1.3 AREAS OF REVIEW 

The reviewer should review the applicant's submittal for an acceptable physical protection 

system that protects against the design basis threats of both theft or diversion of formula 

quantities of SSNM and radiological sabotage. The reviewer should ensure that the applicant 

has described how the general performance requirements of 10 CFR 73.20, the performance 

capabilities outlined in 10 CFR 73.45, and the specific measures included in 10 CFR 73.46 will 

be met through development, implementation, and maintenance of a physical protection 

system. 

13.1.4 ACCEPTANCE CRITERIA 

13.1.4.1 Regulatory Requirements 

Specific references are as follows: 

A. In 10 CFR 73.20, the general performance objective and requirements for fixed-site physical 

protection systems are defined. 

B. In 10 CFR 73.45, the performance capabilities for fixed-site physical protection systems are 

defined.

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Safeguards

C. In 10 CFR 73.46, specific measures for fixed-site physical protection systems, subsystems, 

components, and procedures are detailed. 

D. Appendices B, C, G, and H to 10 CFR 73.46 provide additional requirements applicable to 

the mixed oxide (MOX) facility. 

13.1.4.2 Regulatory Guidance 

The regulatory guidance for physical protection includes: 

International Atomic Energy Agency (IAEA), Information Circular 225, Rev.4 (corrected), "The 

Physical Protection of Nuclear Material and Nuclear Facilities." IAEA: Vienna, Austria. June 

1999. 

Nuclear Regulatory Commission, (U.S.) (NRC). NUREG-BR0252, "User's Guide to Physical 

Protection Documents Published by the NRC." NRC: Washington, D.C. November 1998. 

Regulatory Guide (RG) 5.52, Rev. 3, "Standard Format and Content of a Licensee 

Physical Protection Plan for Strategic Special Nuclear Material at Fixed Sites (Other than 

Nuclear Power Plants)." NRC: Washington, D.C. December 1994. 

RG 5.44, "Perimeter Intrusion Alarm Systems." NRC: Washington, D.C. October 

1997. 

RG 5.55, "Standard Format and Content of Safeguards Contingency Plans for Fuel 

Cycle Facilities (for comment)." NRC: Washington, D.C. March 1978. 

13.1.4.3 Regulatory Acceptance Criteria 

The reviewer will find the applicant's physical protection system acceptable if the physical 

protection plan commitments are consistent with 10 CFR 73.20, 73.45, and 73.46. The physical 

protection plan for the MOX facility shall contain inspectable commitments that shall be the 

basis for the NRC physical protection inspection program. Therefore, it is imperative that 

commitments be expressed in unambiguous terms. NRC has determined that public disclosure 

of the details of the physical protection system for a MOX facility could affect common defense 

and security and should be classified as Confidential National Security Information. 

13.1.4.3.1 Introduction and Schedule for Implementation 

The applicant should state its corporate name, the facility name, and the location of the facility. 

The applicant should describe the MOX facility and the type of SNM that will be utilized, its 

general layout, its surrounding area, and the surrounding terrain. The reviewer should ensure 

that the applicant has included a map of the entire facility and other maps and illustrations, as 

appropriate. The applicant should indicate on these maps the owner-controlled area; the 

location of all buildings; the locations of physical protection systems, subsystems, and major 

components; the protected area and all entry/exit points; vehicle barriers; all material access

NUREG-1718 13.1-2
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areas; vital areas (if applicable); controlled access areas; vaults; entry/exit control points; alarm 

stations; security posts; and response force staging areas. 

The applicant should describe the schedule for implementing the physical protection plan. 

SSNM may not be stored or used at the MOX facility until the physical protection system is fully 

implemented and operational. 

13.1.4.3.2 General Performance Objectives 

The reviewer will determine that the applicant's commitments in this section are consistent with 

10 CFR 73.46. In addition, the reviewer should verify the following: 

A. The applicant has described in general terms how the physical protection system will have 

as its objective to provide high assurance that activities involving SNM are not inimical to the 

common defense and security and do not constitute an unreasonable risk to the public 

health and safety. 

B. The applicant has described how, through the development, implementation, and 

maintenance of a physical protection system, the general performance objective and 

requirements outlined in 10 CFR 73.20 and the performance capability requirements of 

10 CFR 73.45 will be met. 

Further, the reviewer should ensure that the applicant has identified and described those 

portions of the physical protection system for which redundant and diverse components, as well 

as redundant and diverse subsystems and components, are necessary to ensure adequate 

performance as required by 10 CFR 73.20(b)(2). In general terms, the applicant should 

describe the subsystems and components to be used to provide this redundancy and diversity 

and the ways in which these subsystems and components are redundant and diverse. 

Finally, the reviewer should verify that the applicant has described how the physical protection 

system is designed, tested, and maintained to ensure its continuous effectiveness, reliability, 

and availability. This verification should be conducted onsite by the reviewer prior to plan 

approval. 

13.1.4.3.3. Design Basis Threat (10 CFR 73.1(a)) 

The applicant has affirmed the intent to protect against the design basis threats of both theft or 

diversion of formula quantities of SSNM and radiological sabotage, as described in 

10 CFR 73.1 (a). For a MOX fuel fabrication facility, it is important that the physical protection 

system be designed both to protect against radiological sabotage and to prevent theft of 

formula quantities of SSNM. With respect to radiological sabotage, the applicant is expected to 

establish a defensive strategy that would deny unauthorized access to areas of the plant that 

contain plutonium. The reviewer should ensure that the applicant has committed to maintain 

and update the physical protection plan to reflect any changes that are necessary to ensure the 

continuous ability to protect against the design basis threats.

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Safeguards

13.1.4.3.4 Security Organization (10 CFR 73.46(b)) 

The performance objective of the security organization is to manage, control, and implement 

the physical protection system in a manner that is consistent with the physical protection plan 

and continuously maintains its effectiveness. The reviewer should ensure that the applicant 

has clearly described the security organization that will be used at the facility. The security 

organization should be acceptable if the applicant's commitments are consistent with the 

requirements in 10 CFR 73.46(b), and associated Appendices B, C, G, and H of 

10 CFR Part 73, and the following criteria: 

A. The applicant has stated whether the security organization is employed directly by the 

applicant or is a contractor to the applicant. If a contractor, the reviewer should ensure that 

the applicant described the written agreements between the applicant and contract guard 

force management that pertain to how the guard force will meet NRC's requirements in 

10 CFR 73.46(b)(1) and in Appendix B, "General Criteria for Security Personnel," and 

Appendix H, "Weapons Qualification Criteria," to 10 CFR Part 73. 

B. The applicant has described the structure and management of the security organization, 

including both uniformed security personnel and other persons responsible for security

related functions, consistent with 10 CFR 73.46(b)(1). This discussion should include a 

description of each supervisory and management position, including responsibilities and 

lines of authority to facility and corporate management. 

C. The applicant has affirmed that at least one full-time member of the security organization 

will be onsite at all times with the authority to direct the physical protection activities of the 

security organization, consistent with 10 CFR 73.46(b)(2). The plan should also affirm that 

written security procedures will be used and that provisions for written approval of such 

procedures, and any revision thereto, are developed and used, consistent with 

10 CFR 73.46(b)(3). 

D. The applicant has affirmed that an approved Guard Force Training Plan will be in effect in 

accordance with Appendix B to 10 CFR Part 73. The physical protection plan should 

commit to having all members of the security organization trained, equipped, and qualified 

to perform each assigned security duty per 10 CFR Part 73 Appendix B and Appendix H as 

appropriate, consistent with 10 CFR 73.46(b)(4). 

E. The applicant has described how the security personnel, licensee employees, or contractor 

employees will carry out their assigned duties or responsibilities upon the request of the 

NRC. The applicant should also affirm that, within any given period of time (e.g., at least 

one work shift or 8 hours), a member of the security organization will not be assigned to or 

have direct operational control over more than one of the redundant elements of a physical 

protection subsystem if such assignment or control could result in the loss of effectiveness 

of the subsystem, consistent with 10 CFR 73.46(b)(5).

NUREG-1718 13.1-4
Safeguards

F. The applicant has affirmed that every guard, armed response person, and Tactical 

Response Team (TRT) member will be armed and should describe the armament assigned 

to members of the security force by position title, consistent with 10 CFR 73.46(b)(6). The 

applicant should include a description of the qualification and requalification program for 

guard and TRT members in night firing with assigned weapons, and, for TRT members 

only, a description of the training program in response tactics, consistent with 

10 CFR 73.46(b)(7) and (8). In addition, the applicant should describe the equipment to be 

used by members of the security force in providing effective response capabilities. 

G. The applicant has described how scenarios for force-on-force exercises are developed, the 

design goals for conducting such exercises, and the frequency of exercises. The applicant 

should affirm that as a licensee it will permit NRC to observe one force-on-force exercise 

each year and that the NRC will receive a 60-day notice of the planned exercise, consistent 

with 10 CFR 73.46(b)(9). 

H. The applicant has affirmed that the records required by 10 CFR 73.46(b)(3)(i), (4), (7), (8) 

and (9) will be maintained/retained and has described how they will be maintained/retained. 

I. The applicant has described the physical fitness training program and medical examination 

for each guard, armed response person, or TRT member consistent with 

10 CFR 3.46(b)(10)-(12) to ensure that these personnel are able to perform their assigned 

duties under conditions of strenuous tactical engagements. 

13.1.4.3.5 Physical Barrier Subsystems (10 CFR 73.46(c)) 

A performance objective of physical barriers is to define areas within which authorized activities 

and conditions are permitted. Other performance objectives of barriers are to channel persons, 

vehicles, and material to or from entry/exit control points; to delay or deny unauthorized 

penetration attempts by persons, vehicles, or material; and to delay any unauthorized SSNM 

removal attempts sufficient to assist detection and assessment and permit a timely response by 

the security force to prevent the intended act. The reviewer should ensure that the applicant 

has clearly described the physical barrier subsystems that will be used at the facility. This 

section should be acceptable if the applicant's commitments are consistent with the 

requirements in 10 CFR 73.46(c) and the following criteria: 

A. The applicant has described the facility's protected, controlled access, material access, and 

vital (if applicable) area barriers; discussed the purpose of each barrier; and described the 

spatial relationship between the protected area and material access or vital areas, 

consistent with 10 CFR 73.46(c)(2). 

B. The applicant has affirmed that the perimeter of the protected area will be provided with two 

physical barriers, as defined in 10 CFR 73.2. The inner barrier must be positioned, 

constructed, and maintained to enhance assessment of penetration attempts and to delay 

attempts at unauthorized exit from the protected area, consistent with 10 CFR 73.46(c)(1). 

The applicant should commit to installing the protected area barrier fence so that it cannot 

be lifted to allow an individual to crawl under. The applicant should describe any access

13.1-5 NUREG-1718
Safeguards

points in the protected area barrier, their use, and how they are controlled and protected to 

ensure the integrity of the barrier. 

C. The applicant has described the location and size of all isolation zones at the facility. 

Affirmation should be given that the isolation zones adjacent to the physical barriers at the 

perimeter of the protected area should be at least 6.1 m (20 ft) wide and be maintained 

clear of obstacles or structures on either side of the barriers to permit assessment, 

consistent with 10 CFR 73.46(c)(3). 

D. The applicant has affirmed that the location and placement of vehicle barriers will provide 

protection against radiological sabotage by the design basis explosive (classified) or the use 

of a vehicle for transporting personnel and their equipment into the protected area to aid in 

the theft of SSNM. The applicant should include the physical description of the barrier 

system along with a commitment that the barrier can adequately counter the design basis 

vehicle (classified), consistent with 10 CFR 73.46(c)(1). If other than a commercially 

available barrier is used, the applicant should discuss any testing conducted to validate the 

penetration resistance of the barrier. 

E. The applicant has described the lighting system provided to ensure illumination for all 

required monitoring, observation, and assessment activities for all exterior areas within the 

protected area. The commitment for illumination should be not less than 2.15 lumen per 

meter2 (0.2 footcandle) measured horizontally at ground level, consistent with 

10 CFR 73.46(c)(4). The applicant should discuss emergency backup power for protected 

area lighting and assessment capability if normal power is lost. 

F. The applicant has described the purpose of each process material access area at the 

facility and the protection afforded SSNM (other than alloys, fuel elements, or fuel 

assemblies) while in these material access areas. Both physical and procedural protective 

measures should be described, consistent with 10 CFR 73.46(c)(5). 

G. The applicant has affirmed that physical barrier subsystems will be in place to ensure that 

SSNM is stored or processed only in a material access area, vital equipment is located only 

within a vital area, and both vital and material access areas are located within a protected 

area. Physical barriers will be maintained for vital or material access areas that are 

separated from any physical barrier at the perimeter of the protected area. The applicant 

should describe the level of physical hardening for the walls, floors, and ceilings of these 

areas. The number, location, and types of entry/exit portals should be described. Methods 

used to provide hardening of the portals (during open and closed conditions) should be 

described. Hardening for ventilation and other openings greater than 619.4 cm 2 (96 square 

inches), with the smaller dimension of 15.2 cm (6 inches) or greater, should be described. 

Access to vital equipment or SSNM will require passage through at least three physical 

barriers. 

H. The applicant has affirmed that SSNM other than alloys, fuel elements, or fuel assemblies 

shall be stored in a vault when not undergoing processing if the material can be used 

directly in the manufacture of a nuclear explosive device, consistent with

NUREG-1718 13.1--6
Safeguards

10 CFR 73.46(c)(5). The applicant should describe the purpose; the construction of the 

walls, ceiling, and floor; and the location and type of entry portal to each vault. The 

penetration delay time for the vault should be estimated by the applicant based on the vault 

construction method and materials and considering penetration by both tools and explosive 

techniques. The applicant should affirm that the penetration delay time will be greater than 

the time required for the TRT to respond. 

I. The applicant has described the construction and use of tamper-indicating containers for 

the storage of SSNM (other than alloys, fuel elements, or fuel assemblies), consistent with 

10 CFR 73.46(c)(5). 

J. The applicant has described how fuel elements and fuel assemblies will be stored and 

protected. 

13.1.4.3.6 Access Control Subsystems and Procedures (10 CFR 73.46(d)) 

The performance objective of access authorization controls and procedures is to provide 

current authorization lists and entry criteria. The performance objectives of entry controls and 

procedures are to verify the identity of persons, vehicles, and materials; assess such identities 

against current authorization lists and entry criteria before permitting entry; and initiate timely 

response measures to deny unauthorized entries. The reviewer should ensure that the 

applicant has clearly described the access control subsystems that will be used at the facility. 

This section should be acceptable if the applicant's commitments are consistent with the 

requirements in 10 CFR 73.46(d) and the following criteria: 

A. The applicant has described the numbered picture badge identification system used at the 

facility, consistent with 10 CFR 73.46(d)(1). This description should include a discussion of 

procedures used for badging individuals authorized access to the protected area and for 

individuals whose are not employed by the applicant but require frequent and extended 

access to the protected area. Instructions that badged individuals receive in proper badge 

procedures should also be discussed, along with procedures for accommodating 

nonbadged emergency response individuals during emergency situations. Verification of 

authorization can be accomplished by use of systems such as biometrics, personal 

identification numbers, card readers, or combinations thereof. Badges should not be taken 

offsite unless the applicant commits to using a highly reliable method of verifying personal 

identity such as biometrics. The applicant should affirm that blank badge material will be 

controlled. The applicant should affirm that the badge of an employee terminated for cause 

should be immediately retrieved or deleted from the computerized access system. 

B. The applicant has affirmed that badges will be required to be displayed by all individuals 

while inside the protected area, consistent with 10 CFR 73.46(d)(1). 

C. The applicant has committed to procedures for determining an individual's need for access 

to a vital area, material access area, or controlled access area; procedures for distributing 

and maintaining lists of authorized individuals; procedures for ensuring the maintenance of 

the two-man rule within material access areas and vaults; procedures for ensuring that no

13.1-7 NUREG-1718
Safeguards

activities other than those that require access to SSNM or necessary maintenance are 

permitted within material access areas; and methods for visually identifying individuals 

authorized unescorted access to vital areas, material access areas, or controlled access 

areas, consistent with 10 CFR 73.46(d)(2). This discussion should note differences in 

procedures, if any, between working and nonworking hours (nights, weekends, and 

holidays) and normal versus emergency conditions. The applicant should commit that 

access to material access areas and vaults requires a minimum of two individuals to be 

present. The applicant should commit to allowing unescorted access to vital, material 

access, or controlled access areas only to individuals with a government security clearance 

and a need to know. 

D. The applicant has described how it will control all points of personnel access into the 

protected area under both normal and emergency conditions, consistent with 

10 CFR 73.46(d)(4). This description should include a discussion of methods used to 

identify individuals and to verify individuals' authorization; methods used to verify 

emergency conditions; and procedures for conducting searches of individuals for firearms, 

explosives, and incendiary devices. The search function for detecting both firearms and 

explosives must use detector equipment. The equipment used should represent the current 

state-of-the-art equipment that is commercially available. The capabilities of the search 

equipment should be described. The applicant should also describe what actions it takes, 

including the use of pat-down searches, if it suspects an individual of trying to introduce 

contraband into the protected area or if the search equipment is not operating satisfactorily. 

The applicant should describe how it will determine that the equipment is operating properly. 

The applicant should commit to having the individual responsible for the last access control 

point prior to the protected area to be protected by a bullet-resisting structure hardened to at 

least the Underwriters Laboratories Inc. (UL) 752/Class IV level and preferably to the 

7.62 mm level of protection. 

E. The applicant should identify individuals exempted from any of the aforementioned access 

controls. The distribution and maintenance of authorization lists should also be described. 

F. The applicant has affirmed that it will establish and follow written procedures that will permit 

access-control personnel to identify materials in hand-carried packages that are not 

authorized entry to the protected area during both normal and emergency conditions, 

consistent with 10 CFR 73.46(d)(5). Further, the applicant should describe procedures for 

searching hand-carried packages at personnel and vehicle access points for firearms, 

explosives, and incendiary devices. 

G. The applicant has affirmed that it will establish and follow written procedures that will permit 

access-control personnel to identify materials in delivered packages that are not authorized 

entry to the protected area during both normal and emergency conditions, consistent with 

10 CFR 73.46(d)(6). Further, methods used to check for proper identification and 

authorization should be described, along with search procedures for firearms, explosives, 

and incendiary devices. Any activities exempted from the above procedures should be 

described. The development, distribution, and maintenance of authorized (or unauthorized) 

materials lists should be described.

NUREG-1718 13.1-8
Safeguards

H. The applicant has described procedures used for controlling all points of vehicle access 

(nonemergency and emergency) into the protected area and how written procedures are 

established and followed that will permit access-control personnel to identify vehicles that 

are authorized entry to the protected area, consistent with 10 CFR 73.46(d)(3). The 

distribution and maintenance of these procedures should be described. Search procedures 

of all vehicles requiring entry to the protected area for firearms, explosives, and incendiary 

devices should also be described. Any vehicles exempt from the aforementioned 

procedures should be described, consistent with 10 CFR 73.46(d)(7). Procedures used in 

escorting vehicles within the protected area, and areas where vehicles may have access, 

along with the purpose for the access, should be discussed. 

I. The applicant has described the control and use of designated licensee vehicles within the 

protected area, consistent with 10 CFR 73.46(d)(8). 

J. The applicant has described the methods it proposes to use to control all points of 

personnel access to material access areas, vital areas, and controlled access areas, 

including methods used to verify identification and authorization, consistent with 

10 CFR 73.46(d)(9). The applicant shall affirm that at least two armed and appropriately 

trained guards shall be posted at each material access area control point whenever it is in 

use. Personnel exit searches from material access areas should also be discussed, and the 

applicant should affirm that at least two individuals who are not authorized access to that 

material access area will conduct separate, independent searches for concealed SSNM. 

The applicant should affirm that material access area exit searches for SNM and metal can 

detect within established standards, consistent with NRC classified criteria. 

K. The applicant has described procedures for verifying material entry authorizations and 

procedures for verifying quantity and type of material, consistent with 10 CFR 73.46(d)(9). 

These descriptions should include the components to be used in the detection of 

unauthorized materials that are hand-carried by authorized individuals, or mailed or 

otherwise shipped as part of an authorized shipment. The applicant has described how 

normal conditions differ between regular working hours and nonworking hours (nights, 

weekends, and holidays). 

L. The applicant has described methods used to control all points of vehicle access 

(nonemergency and emergency) to material access areas, vital areas, and controlled 

access areas, including the establishment and maintenance of written procedures that will 

permit access control personnel to identify those vehicles that are authorized entry to 

material access and vital areas, consistent with 10 CFR 73.46(d)(9). Vehicle exit searches 

should also be described, and the applicant should affirm that searches will be conducted 

by a team of at least two individuals. 

M. The applicant has described procedures and areas used for searching contaminated wastes 

coming from a material access area, consistent with 10 CFR 73.46(d)(1 0).

13.1-9 NUREG-1 718
Safeguards

N. The applicant has described containers, procedures, and areas used for shipping SSNM 

offsite, consistent with 10 CFR 73.46(d)(1 1) and (12). The applicant should affirm that the 

packaging and shipping process will be conducted by a team of at least two individuals. 

0. The applicant has described by job function individuals who may be designated as escorts 

and procedures used for escorting individuals during both routine and emergency situations. 

Such procedures should describe individuals requiring escort, escort/visitor ratios, badging 

procedures, and escort training and recordkeeping, consistent with 10 CFR 73.46(d)(1 3). 

The applicant should commit to a maximum escort/visitor ratio of at least one escort to five 

visitors. 

P. The applicant has described procedures for controlling all keys, locks, combinations, and 

related equipment used to control access to protected, material access, vital, and controlled 

access areas. The discussion should describe the circumstances under which such keys, 

locks, combinations, and related equipment are changed and procedures followed when the 

employment of an employee with access to such keys, locks, combinations, etc., is 

terminated, consistent with 10 CFR 73.46(d)(1 4). The applicant should commit to changing 

keys, locks, combinations, and related equipment at least when there is evidence of 

compromise to any of the items to which a terminated employee had access. 

Q. The applicant has committed to controlling information regarding the presence of NRC 

safeguards inspectors, consistent with 10 CFR 73.46(d)(1 5). 

R. The applicant should describe recordkeeping procedures for: (1) current written procedures 

that permit access-control personnel to identify vehicles that are authorized and those 

materials that are not authorized entry to protected, material access, and vital areas; (2) 

findings of drum-scanning and tamper-sealing of containers of contaminated wastes exiting 

from material access areas; and (3) the required log of escorted individuals, consistent with 

10 CFR 73.46(d)(3), (10), and (13). 

13.1.4.3.7 Detection, Surveillance and Alarm Subsystems and Procedures 

(10 CFR 73.46(e)) 

The performance objectives of detection, surveillance, and alarm subsystems and procedures 

are to detect, assess, and communicate any unauthorized access or penetrations or such 

attempts by persons, vehicles, or materials at the time of the act or attempt so that the 

response can be such as to prevent unauthorized access or penetration. The reviewer should 

ensure that the applicant has clearly described the detection, surveillance, and alarm 

subsystems that will be used at the facility. This section should be acceptable if the applicant's 

commitments are consistent with the requirements in 10 CFR 73.46(e) and the following 

criteria: 

A. The applicant has described the intrusion detection system that will be installed in the 

isolation zone between the two barriers at the protected area perimeter, consistent with 

10 CFR 73.46(e)(1). The applicant should commit to providing a volumetric intrusion 

detection system, which is capable of detecting an individual weighing a minimum of

NUREG-1718 13.1-10
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35 kg (77 Ib) whether the individual is running, walking, crawling, jumping, or rolling through 

the isolation zone of the protected area. The capabilities, installation, and testing of the 

intrusion detection equipment should be consistent with Revision 3 of Regulatory 

Guide 5.44. 

B. The applicant has described the location of all emergency exits and described the protection 

afforded them, consistent with 10 CFR 73.46(e)(2). The applicant should commit to 

maintaining all emergency exits in the protected, material access, vital, and controlled 

access areas locked to prevent entry from outside and to equipping them with local audible 

and visible alarms. 

C. The applicant has described the protection and surveillance afforded: (1) unoccupied 

material access and vital areas; (2) the location of SSNM within process material access 

areas; and (3) vaults and process areas that contain SSNM that has not been alloyed or 

encapsulated, including a description of procedures for access to these particular vaults and 

process areas, consistent with 10 CFR 73.46(e)(3). Equipment used to provide this 

protection, along with associated detection capabilities, should also be described. The 

applicant should commit to having all unoccupied material access areas where plutonium is 

located equipped with volumetric intrusion detection equipment and closed circuit television 

(CCTV) for remote assessment. The applicant should affirm that access to unoccupied 

vaults and process areas requires that an individual other than the alarm station operator be 

present or have knowledge of access. 

D. The applicant has described how all security stations and individuals (by job position), 

consistent with 10 CFR 73.46(e)(4), will be provided with duress alarms. The type of duress 

alarms used, where they are monitored, and emergency backup power should be 

described. 

E. The applicant has described the location, construction, and characteristics of the central 

and secondary alarm stations, consistent with 10 CFR 73.46(e)(5). The applicant should 

commit to having all required alarms annunciate in a continuously manned central alarm 

station located within the protected area and in at least one other continuously manned 

independent onsite station. Continuous manning of alarm stations and methods used for 

annunciation of required alarms should be described, along with protection afforded the 

stations (both procedural and physical), so that a single act cannot remove the capability of 

calling for assistance or responding to an alarm. Affirmation also should be provided that 

the alarm stations are bullet-resisting to at least the UL 752/Class IV level, and preferably to 

the 7.62 mm level of protection. If other than commercially available armoring material is 

used, any testing or engineering studies conducted to validate the penetration resistance of 

the barrier should be described. Affirmation should be given that access to the alarm 

stations is controlled on a strict need-to-know basis and the central alarm station not contain 

any operational activities that would interfere with the execution of alarm response 

functions. The applicant should describe the annunciation systems at the alarm stations 

and commit to indicating the status of all alarms and alarm zones in both alarm stations.

13.1-11 NUREG-1718
Safeguards

F. The applicant has described (1) how detection equipment and alarm annunciation shall 

remain operable from independent emergency power sources, (2) the duration of operation 

in the event of loss of normal power, and (3) the indications given upon loss of normal 

power and transfer to standby power, consistent with 10 CFR 73.46(e)(6). The applicant 

should also affirm that switchover to standby power will be automatic and not cause false 

alarms. 

G. The applicant has described the physical protection afforded alarm systems, including 

transmission media, to ensure that the system is not being tampered with, compromised, or 

on standby power without the knowledge of the licensee, consistent with 

10 CFR 73.46(e)(7). The applicant should affirm that all tamper alarms will annunciate in 

either the access or the secure mode. 

H. The applicant has described methods used to monitor all exterior areas within the protected 

area for unauthorized persons, vehicles, materials, and activities, and the duration or 

periodicity of such monitoring, consistent with 10 CFR 73.46(e)(8). Criteria used in defining 

authorized and unauthorized activities and conditions within the protected area should be 

described, along with methods for developing, maintaining, and distributing lists of 

authorized activities and conditions. The applicant should commit to monitoring or 

conducting random patrols within the protected area at least several times each shift. 

1. The applicant has described methods used to observe individuals within material access 

areas to ensure that SSNM is not moved to unauthorized locations or moved in an 

unauthorized manner, consistent with 10 CFR 73.46(e)(9). The duration or periodicity of 

such monitoring should be described, along with criteria used in defining authorized and 

unauthorized activities and conditions within the material access area. Methods for 

developing, maintaining, and distributing lists of authorized activities and conditions should 

be described. The applicant should commit to using CCTV to observe these areas 

periodically during working hours and for remote access during nonworking hours. 

13.1.4.3.8 Communication Subsystems (10 CFR 73.46(f)) 

The performance objective of communication subsystems is to provide for notification of an 

attempted unauthorized or unconfirmed removal of SSNM or attempted act of radiological 

sabotage so that response can be such as to prevent the unauthorized act. The reviewer 

should ensure that the applicant has clearly described the communication subsystems that will 

be used at the facility. The communication subsystems should be acceptable if the applicant's 

commitments are consistent with the requirements in 10 CFR 73.46(f) and the following criteria: 

A. The applicant has described how each guard, watchman, armed response person,. or TRT 

member on duty will be capable of maintaining continuous communications with the 

individuals in each continuously manned alarm station, consistent with 10 CFR 73.46(f)(1). 

The applicant should also describe how the individuals in each continuously manned alarm 

station will be capable of Calling for assistance from other guards, watchmen, armed 

response personnel, or TRT members and from local law enforcement authorities.

NUREG-1718 13.1-12
Safeguards

B. The applicant has described the redundant and diverse systems used to ensure the 

capability of communications with the local law enforcement authorities, consistent with 

10 CFR 73.46(f)(2). Cellular phone service may be an acceptable alternative method of 

communications if the service is reliable and provides complete coverage of the area of 

concern. 

C. The applicant has described methods used to keep the nonportable communications 

equipment it uses operable in the event of loss of normal power, consistent with 

10 CFR 73.46(f)(3). The applicant should discuss the length of time the equipment will 

operate on the emergency power source. All sources of emergency power should be 

protected and located within the protected area. 

13.1.4.3.9 Test and Maintenance Programs (10 CFR 73.46(g)) 

The performance objective of test and maintenance programs is to provide confidence that 

security equipment will be available and reliable to perform its function when needed. The 

review should ensure that the applicant has clearly described the test and maintenance 

programs that will be used at the facility. The test and maintenance programs should be 

acceptable if the applicant's commitments are consistent with the requirements in 

10 CFR 73.46(g) and the following criteria: 

A. The applicant has described the test and maintenance programs for (1) intrusion alarms; 

(2) emergency exit alarms; (3) communications equipment; (4) physical barriers; and 

(5) other physical protection-related devices and equipment such as CCTV, locks, 

emergency power sources, alarm annunciators, duress alarms, search equipment, etc. 

used pursuant to 10 CFR 73.46 during the installation and construction, preoperational and 

operational tests of the physical protection subsystems and components, consistent with 

10 CFR 73.46(g)(1)-(3). This discussion should also include the purpose for and intended 

level of testing and maintenance programs. In addition, specific methods for testing each 

type of equipment should be discussed, along with periodicity of testing, consistent with 

10 CFR 73.46(g)(3). The applicant should commit to having a testing program for the 

perimeter intrusion detection system consistent with Revision 3 of Regulatory Guide 5.44. 

The applicant should describe the sensitivity of the SNM, metal, explosive, and x-ray search 

equipment and the device used for calibration. The applicant should commit to using a 

device comparable to one which meets the American Society for Testing and Materials 

(ASTM) ASTM-F792 standard, "Standard Practice for Design and Use of Ionizing Radiation 

Equipment for the Detection of Items Prohibited in Controlled Access Areas," consistent 

with NRC classified criteria. 

B. The applicant has described the preventive maintenance program established to ensure the 

maintenance of all physical protection-related subsystems and components in operable and 

reliable condition, consistent with 10 CFR 73.46(g)(4) and (5). The applicant should 

describe corrective actions or compensatory measures used in the event of component 

failure within physical protection systems.

13.1-13 NUREG-1718
Safeguards

C. The applicant has described procedures used in performing repairs and maintenance of 

physical protection systems, consistent with 10 CFR 73.46(g)(5). The applicant should 

commit that all repairs and maintenance will be performed by two individuals working as a 

team and that performance verification tests will be conducted after maintenance has been 

completed. 

D. The applicant has described how it will review and audit its security program, consistent with 

10 CFR 73.46(g)(6). This discussion should include the periodicity of the review and audit, 

a description of who will conduct the review and audit, items to be covered by the review 

and audit, how the review and audit will be documented, to whom the review and audit 

findings will be provided for review, and the recordkeeping associated with the review and 

audit program. 

E. The commitment for the frequency of the annual audit should not vary by more than plus or 

minus 1 month. 

13.1.4.3.10 Contingency and Response Plans and Procedures (10 CFR 73.46(h)) 

The performance objective of contingency and response plans and procedures is to provide for 

predetermined response to safeguards contingency events so that the adversary will be 

engaged and impeded until offsite assistance arrives. The reviewer should ensure that the 

applicant has clearly thought out potential contingencies and has clearly described contingency 

and response plans that will be used by the facility. The contingency and response plans 

should be acceptable if the applicant's plans are consistent with the requirements in 

10 CFR 73.46(h) and developed in accordance with the criteria in Appendix C to Part 73 and 

the following criteria: 

A. The applicant has established a safeguards contingency plan for dealing with threats, thefts, 

and radiological sabotage related to SSNM and its facility and commits to maintain and 

follow the plan, consistent with 10 CFR 73.46(h)(1). 

B. The applicant has described the documented response arrangements it has made with local 

law enforcement agencies, consistent with 10 CFR 73.46(h)(2). This description should 

include estimated number of response individuals with specific response times of arrival that 

are consistent with NRC classified criteria. 

C. The applicant has described the number of TRT members immediately available for 

response and the duties they will be assigned. TRT members may be physically located at 

the facility or at a nearby facility such that their response is timely, effective, and not easily 

interdicted to ensure protection against the design basis threats defined in 10 CFR 73.1(a). 

In addition, the required force of guards or armed responders available onsite to assist the 

TRT should be described, along with a discussion of the rationale for determining the 

number of individuals in this force of guards or armed responders and the availability of this 

force, consistent with 10 CFR 73.46(h)(3).

NUREG-1718 13.1-14
Safeguards

D. The applicant has described its planned response procedures for dealing with the detection 

of abnormal presence or activity of persons or vehicles within an isolation zone, the 

protected area, a material access area, or a vital area; or evidence or indication of intrusion 

into the protected area, material access area, or a vital area should be described. The 

applicant has also described the methods for assessing the threat and responding to the 

threat, consistent with 10 CFR 73.46(h)(4). The applicant should establish a defensive 

strategy that would deny unauthorized access to areas of the facility that contain plutonium. 

The applicant should commit to requiring guards to interpose themselves between vital and 

material access areas and any adversary attempting entry for purposes of radiological 

sabotage or theft of SSNM, to intercept any persons exiting with SSNM, and to inform local 

law enforcement of the threat and request assistance. 

E. The applicant has described the instructions that guards and armed responders will receive 

in the use of force, including the use of deadly force, in preventing or impeding theft of 

SSNM, consistent with 10 CFR 73.46(h)(5). 

F. The applicant has described the methods that will be used for providing assessment of all 

protected area alarms. The applicant should commit to using CCTV or other suitable 

means that limit exposure of responding personnel to possible attack and permit 

assessment of the protected area barrier and associated isolation zones, consistent with 

10 CFR 73.46(h)(6). The applicant should commit to the CCTV providing unobstructed view 

of the protected area barrier and isolation zones with no blind spots. 

G. The applicant has described methods that will be used for assessing alarms occurring within 

unoccupied vaults and unoccupied material access areas containing plutonium, and the 

timeliness of assessment. The applicant should commit to using at least two security 

personnel to assess alarms by CCTV or other remote means that occur within unoccupied 

vaults and unoccupied material access areas, consistent with 10 CFR 73.46(h)(7). 

H. The applicant has described methods that will be used for assessing alarms occurring within 

unoccupied vaults and unoccupied material access areas containing alloyed or 

encapsulated SSNM, and the timeliness of assessment. The applicant should commit to 

using at least two security personnel to remotely assess alarms by CCTV, or by at least two 

security personnel who are searched before exiting the material access areas, consistent 

with 10 CFR 73.46(h)(8). 

I. The applicant has described how it will establish, maintain, and retain as a record the 

current safeguards contingency plan, consistent with 10 CFR 73.46(h)(1) and (2). 

13.1.4.3.11 Reporting of Safeguards Events (10 CFR 73.71) 

Acceptance should be based on the fact that the applicant adequately addresses how and 

when it will report safeguards events to the NRC and follows the criteria in 10 CFR Part 73, 

Appendix G, "Reportable Safeguards Events."

13.1-15 NUREG-1718
Safeguards

13.1.5 REVIEW PROCEDURES 

13.1.5.1 Acceptance Review 

The primary reviewer should perform an acceptance review to determine if the application 

adequately addresses the items in Section 13.1.3, "Areas of Review," for either the construction 

approval review or the review for a license to possess and use SNM. 

Guidance specific to the construction approval review and the review for a license to possess 

and use SNM is provided below. 

A. Construction Approval 

Although the. applicant is not expected to submit a physical protection plan for construction 

approval, the applicant should commit to developing and implementing a physical protection 

system that meets or exceeds the acceptance criteria in Section 13.1.4. If provided by the 

applicant, the primary reviewer should evaluate the proposed location and construction 

technique and materials of the buildings; protected, vital, material access, and controlled 

access area barriers; vehicle barriers; alarm stations; security search or control points; and 

vaults to ensure that the commitments and program goals, as described in Section 13.1.3, 

are appropriate for physical protection at the design stage. 

B. License To Possess and Use SNM 

Specifically, the license application should address Section 13.1.3 in full. The applicant is 

expected to provide a physical protection plan with the application for a license to possess 

and use SNM. 

If the primary reviewer verifies that physical protection is adequately addressed (construction 

approval or license to possess and use SNM), the primary reviewer should accept the 

application for the safety evaluation in Section 13.1.5.2. If the primary reviewer identifies 

significant deficiencies in the material provided, the primary reviewer should request that the 

applicant submit additional information prior to the start of the safety evaluation. 

13.1.5.2 Safety Evaluation 

After determining that the application is acceptable for review in accordance with either 

Section 13.1.5.1 (A) (construction approval) or 13.1.5.1 (B) (license to possess and use SNM), 

the primary reviewer should perform an evaluation against the acceptance criteria described in 

Section 13.1.4. On the basis of that review, the reviewer may request that the applicant provide 

additional information or modify the application to meet those acceptance criteria. 

Guidance specific to the construction approval review and the review for a license to possess 

and use SNM is provided below.

NUREG-1718 13.1-16
Safeguards

A. Construction Approval 

The primary reviewer should establish that the applicant's proposed design, location, 

construction technique, and material for elements of the physical protection system and 

related commitments will lead to a physical protection plan that will meet or exceed the 

regulatory acceptance criteria in Section 13.1.4. 

B. License to Possess and Use SNM 

The primary reviewer should verify that sufficient information has been provided under 

Section 13.1.4.3 with respect to the physical protection plan and that the information 

provided is consistent with the guidance in this section. 

When the evaluation is complete, the primary reviewer should prepare the physical protection 

input for the Safety Evaluation Report (SER), as described in Section 13.1.6 using the 

acceptance criteria from Section 13.1.4. 

13.1.6 EVALUATION FINDINGS 

The primary reviewer should document the physical protection evaluation by preparing material 

suitable for inclusion in the SER. The primary reviewer should describe the review, identify any 

alternative measures that will be used, explain the basis for the findings, and state the 

conclusions. 

The reviewer could document the safety evaluation for the construction approval review as 

follows: 

The reviewer reviewed the license application for construction approval for [insert name of 

facility] according to Section 13.1 of NUREG-1718. The reviewer evaluated [Insert a 

summary statement of what was evaluated] and found that [summarize the findings]. 

The reviewer concluded that the applicant provided adequate commitments and goals for 

the design of a physical protection system and that these commitments and goals should 

result in a physical protection plan that will meet or exceed the requirements in 

10 CFR 73.20, 73.45 and 73.46 and guidance outlined in NUREG-1718. As a result, the 

applicant meets the requirements under 10 CFR Part 70 for construction approval of the 

facility in the area of physical protection. 

The reviewer could document the safety evaluation for the review for the license to possess and 

use SNM as follows: 

The reviewer reviewed the license application for [insert facility name] according to 

Section 13.1 of NUREG-1 718. The reviewer evaluated [Insert a summary statement of 

what was evaluated] and found [insert a description of the findings]. Based on the review of 

the license application, the reviewer concluded that the applicant adequately described and

13.1-17 NUREG-1718
Safeguards

documented the physical protection system and provided a plan to address the regulations 

in 10 CFR 73.20, 10 CFR 73.45, and 10 CFR 73.46. Meeting the requirements given above 

provides an acceptable basis for the finding that, insofar as physical protection is 

concerned, the applicant meets the associated requirements in 10 CFR 73.20, 73.45, and 

73.46 and therefore the physical protection plan is acceptable to support licensed operation 

under 10 CFR Part 70. 

13.1.7 REFERENCES 

Code of Federal Regulations, Title 10, Energy, Part 11, "Criteria and Procedures for 

Determining Eligibility for Access to Or Control Over Special Nuclear Material." 

. Title 10, Energy, Part 25, "Access Authorization for Licensee Personnel." 

Title 10, Energy, Section 73.20, "General Performance Objectives and Requirements." 

Title 10, Energy, Section 73.45, "Performance Capabilities for Fixed Site Physical 

Protection Systems." 

* Title 10, Energy, Section 73.46, "Fixed Site Physical Protection Systems, Subsystems, 

Components, and Procedures." 

Title 10, Energy, Section 73.71, "Reporting of Safeguards Events." 

Title 10, Energy, Part 95, "Security Facility Approval and Safeguarding of National 

Security Information and Restricted Data." 

Nuclear Regulatory Commission (U.S.) (NRC). NUREG/CR-5081, "Tactical Exercise Planning 

Handbook." NRC: Washington, D.C. April 1989. 

NUREG/CR-5172, "Tactical Training Reference Manual." NRC: Washington, D.C. 

April 1989. 

* RG 5.7, Rev. 1, "Entry/Exit Control for Protected Areas, Vital Areas, and Material 

Access Areas." NRC: Washington, D.C. May 1980. 

RG 5.44, Rev. 3, "Perimeter Intrusion Alarm Systems." NRC: Washington, D.C. 

October 1997. 

-. RG 5.52, Rev. 3, "Standard Format and Content of a Licensee Physical Protection Plan 

for Strategic Special Nuclear Material at Fixed Sites (Other than Nuclear Power Plants)." NRC: 

Washington, D.C. December 1994. 

* RG 5.55, "Standard Format and Content of Safeguards Contingency Plans for Fuel 

Cycle Facilities." NRC: Washington, D.C. March 1978.

NUREG-1718 13.1-18
13.0 SAFEGUARDS

13.2 MATERIAL CONTROL AND ACCOUNTING (MC&A) 

13.2.1 PURPOSE OF REVIEW 

The purpose of this review is to ensure that the Fundamental Nuclear Material Control Plan 

(FNMCP) submitted by the applicant describes how an material control and accounting (MC&A) 

system will be established, implemented, and maintained, and to ensure that the FNMCP is 

adequate to protect against, detect, and respond to the loss or theft of strategic special nuclear 

material (SSNM) by achieving the following five performance objectives stated in 

10 CFR 74.51 (a): 

A. Prompt investigation of anomalies potentially indicative of SSNM losses; 

B. Timely detection of the possible abrupt loss of 5 or more formula kilograms (FKG) of SSNM 

from an individual unit process; 

C. Rapid determination of whether an actual loss of 5 or more FKG occurred; 

D. Ongoing confirmation of the presence of SSNM in assigned locations; and 

E. Timely generation of information to aid in the recovery of SSNM in the event of an actual 

loss. 

These objectives will be achieved by meeting the system capabilities requirements stated in 

10 CFR 74.51(b). 

13.2.2 RESPONSIBILITY FOR REVIEW 

Primary: Safeguards Technical Analyst (MC&A Specialist) 

Secondary: Project Manager 

Supporting: MC&A Physical Scientist (MC&A Inspector) 

Physical Protection Reviewer 

13.2.3 AREAS OF REVIEW 

The staff should review the applicant's FNMCP to ensure that the plan, in meeting the five 

performance objectives stated in Section 13.2.1, addresses: 

A. Process Monitoring Program: For each unit process, the applicant's establishment of a 

production quality control program capable of monitoring the status of material in process;

13.2-1 NUREG-1718
Safeguards

B. Item Monitoring Program: The applicant's establishment of a process to verify the presence 

and integrity of SSNM items on a statistical sampling basis; 

C. Alarm Resolution Program: The applicant's establishment of an alarm resolution program 

that is capable of: 

i. Resolving the nature and causes of any MC&A alarm within approved time periods; 

ii. Notifying NRC of any MC&A alarms that remain unresolved beyond the time periods; 

iii. Determining the amount of actual SSNM lost and taking corrective actions; 

iv. Providing an ability to rapidly assess the validity of alleged thefts; and 

v. Taking appropriate actions when the abrupt loss detection estimate exceeds 2 Kg of 

plutonium. 

D. Quality Assurance and Accounting Programs: The applicant's establishment of a quality 

assurance and accounting capability to address the following 11 elements: management 

structure, personnel qualification and training, measurement, measurement control, physical 

inventory, accounting, shipping and receiving, scrap control, human error, independent 

assessment, and SSNM custodianship. 

13.2.4 ACCEPTANCE CRITERIA 

13.2.4.1 Regulatory Requirements 

Regulatory requirements applicable to the MC&A program and the FNMCP are specified in 

10 CFR Part 74, "Material Control and Accounting of Special Nuclear Material." Subpart E, 

"Formula Quantities of Strategic Special Nuclear Material," particularly applies to MOX fuel 

fabrication facilities. 

13.2.4.2 Regulatory Guidance 

Nuclear Regulatory Commission (U.S.) (NRC). NUREG-1280, Rev. 1, "Standard Format and 

Content Acceptance Criteria for the Material Control and Accounting (MC&A) Reform 

Amendment." NRC: Washington, D.C. April 1995. 

13.2.4.3 Regulatory Acceptance Criteria 

The performance objectives and acceptance criteria discussed below pertain to plutonium, both 

before and after processing into MOX. MOX contains uranium in the form of either depleted, 

natural, or low-enriched uranium (LEU). The reviewer must be aware of which type of uranium 

will be processed into MOX and verify that this is stated in the process description section of the 

FNMCP. If the applicant uses LEU to produce MOX, the reviewer must verify that, up until the

NUREG-1718 13.2-2
Safeguards

time of processing, the LEU feed material is adequately controlled to enable the MC&A system 

to meet the objectives and requirements of 10 CFR Part 74.31. 

It is important that the applicant establish the basis for determining the formula quantity of 

SSNM for a facility processing MOX. FKG means SSNM in any combination in a quantity of 

1,000 grams computed by the formula: 

grams = (grams contained U-235) + 2.5 (grams U-233 + grams plutonium). 

Formula quantity means SSNM in any combination in a quantity of 5,000 grams or more 

computed by the above formula. Where the uranium oxide used in the process has an 

enrichment level lower than 20 percent, the determination of FKG is based on the amount of 

plutonium only. Therefore, 2 Kg of plutonium yields 5 FKG or a formula quantity of SSNM. 

13.2.4.3.1 Performance Objectives 

Reviewers should use a risk-informed, performance-based approach to review the applicant's 

program and capability in meeting the performance objectives in 10 CFR 74.51(a). The 

reviewers should give high priority to the overall timely detection and resolution program. The 

reviewers should evaluate whether the applicant appropriately considered and incorporated a 

collusion protection program in the MC&A system (i.e., threats from an insider and potential 

diversion strategies during fuel processing, in material storage, or from recovery/recycling 

products). The primary reviewer of this section should coordinate with the primary reviewer of 

Section 13.1 where the applicant designed the detection program to be complimentary to the 

physical protection requirements in 10 CFR Part 73 to minimize redundant systems while 

maintaining adequate safeguards assurance. 

13.2.4.3.2 Process Monitoring 

Section 74.53 requires that licensees monitor internal transfers, storage, and processing of 

SSNM. The applicant's process monitoring program should be capable of (1) promptly 

detecting a significant abrupt loss, diversion, or theft of 2 Kg of plutonium with 95 percent power 

of detection; and (2) monitoring the status of material in process. The "prompt" detection is 

dependent upon the classification of the materials (i.e., Category IA or IB), as specified in 

10 CFR 74.53. 

The applicant's process monitoring program should at least consist of: 

A. Clearly defined process subdivisions and measurement points to satisfy unit detection 

criteria and the category of material being processed; 

B. Adequate material control tests for each unit process for detecting abrupt losses with at 

least 95 percent power of detection, evaluation and update of the action threshold on 

semiannual basis, and ability to detect losses involving material substitution; 

C. Basis for material classification (i.e., Category IA and IB materials);

13.2-3 NUREG-1718
Safeguards

D. Clear classification of inaccessible locations; 

E. Identification of all credible substitute materials and the methods of preventing substitution; 

F. A listing of material types exempted from the abrupt loss detection tests with their locations 

and bases for exemption; 

G. Adequate trend analysis techniques and decision criteria, especially for the indication of 

trickling diversions; and 

H. Adequate material balance tests and evaluation for research and development operations. 

It is necessary for the applicant to submit a study of potential diversion scenarios as supporting 

information. Such a study should include, but not be limited to, abrupt losses, trickle diversion, 

insider and/or outsider diversion, unauthorized production, and material substitution. 

The applicant's process monitoring program should be found acceptable if it meets the criteria 

specified in Chapter 1 of Revision 1 to NUREG-1280. In addition, NUREG/CR-4604, 

"Statistical Methods for Nuclear Material Management," provides guidance on statistical tests in 

providing 95 percent power of detection. 

13.2.4.3.3 Item Monitoring 

Section 74.55 requires that licensees establish an item monitoring program capable of providing 

timely plant-wide detection of the loss of items that total 2 Kg of plutonium with 99 percent 

power of detection. The "timely" detection is dependent upon the classification of the material 

(i.e., Category IA or IB), and the degree of tamper-safing that is employed, as specified in 

10 CFR 74.55. The applicant's item monitoring program should at least consist of: 

A. A clear item identification system; 

B. A basis for item classification (i.e., Category IA and IB materials); 

C. A tamper-safing procedure and system; 

D. Accessibility control; 

E. Accounting and control procedures; 

F. Item measurement systems; 

G. Item verification procedures; and 

H. Item sampling techniques. 

The applicant's item monitoring program should be found acceptable if it meets the criteria 

specified in Chapter 2 of Revision 1 to NUREG-1 280. In addition, NUREG/CR-4604, 

"Statistical Methods for Nuclear Material Management," provides guidance on statistical tests in 

providing 99 percent power of detection.

NUREG-1718 13.2-4
Safeguards

13.2.4.3.4 Alarm Resolution 

Section 74.57 requires that the licensees' alarm resolution and reporting programs assure: 

A. Resolution of the nature and cause of any MC&A alarm within approved time periods; 

B. Reports to the NRC within 24 hours of any unresolved MC&A alarm beyond the specified 

time period; 

C. Determination of the amount of SSNM lost and corrective actions when a material loss has 

occurred; 

D. The ability to rapidly assess the validity of alleged thefts; and 

E. The taking of appropriate actions when an abrupt loss detection estimate exceeds 2 Kg of 

plutonium. 

Specifically, the programs should address alarm resolution procedures, decision rules and their 

basis, and response time. 

The applicant's programs for resolving and reporting indications of missing SSNM should be 

found acceptable if they meet the criteria specified in Chapter 3 of Revision 1 to NUREG-1280. 

In addition, the applicant should establish the capability to respond rapidly to alarms occurring 

externally to the MC&A system, as stipulated in Chapter 3.3 of Revision 1 to NUREG-1 280. 

13.2.4.3.5 Quality Assurance and Accounting Programs 

Section 74.59 requires that licensees establish a quality assurance and accounting capability to 

address the 11 areas discussed in Sections 13.2.4.3.5 (A) through (K). 

A. Management Structure 

Section 74.59(b) establishes requirements for the licensees' MC&A management structure, 

organization, responsibilities, procedures, etc. The applicant's MC&A program's 

management structure should demonstrate the checks and balances of the program to 

ensure effective functioning of the MC&A program by providing: 

i. Clear overall responsibility for MC&A responsibilities; 

ii. Independence of MC&A functions from production responsibilities; 

iii. Separation of key MC&A responsibilities from each other to provide controls and 

checks; and 

iv. Adequate review, approval, and use of approved written MC&A procedures.

13.2-5 NUREG-1718
Safeguards

The applicant's organization for developing and implementing the MC&A program and 

procedures should be found acceptable if it meets the criteria specified in Chapter 4.1 of 

Revision 1 to NUREG-1280. (SRP Chapter 2.0 provides additional guidance on 

organization and administration, and SRP Chapter 11.5 provides additional guidance on 

procedures.) 

B. Personnel Qualification and Training 

Section 74.59(c) establishes qualifications and training requirements for key MC&A 

personnel. The applicant's personnel qualification and training programs should ensure that 

qualified and adequately trained personnel are implementing and maintaining an effective 

MC&A program by ensuring that: 

L. Personnel who work in key positions where mistakes could degrade the effectiveness of 

the MC&A program are trained to maintain a high level of safeguards awareness and 

are qualified to perform their duties and/or responsibilities; 

ii. Continuing qualification of key personnel will be verified on an ongoing basis or at least 

every 2 years; and 

iii. The training program emphasizes the job purposes and scope and provides a balance 

between theory and practice. 

The applicant's personnel qualification and training programs should be found acceptable if 

they meet the criteria specified in Chapter 4.2 of Revision 1 to NUREG-1280. (SRP 

Chapter 11.4 provides additional guidance on training and qualification.) 

C. Measurements 

Section 74.59(d) requires that licensees establish and maintain a system of measurements. 

The applicant's measurement program should ensure that: 

i. All source material, SNM, and SSNM information in accounting records is based on 

measured values; 

ii. Key measurement systems and measurement points are identified; 

iii. At each measurement point, the appropriate measurement method and system are used 

for the accurate and precise determination of the material type; 

iv. The MC&A system enables the estimation of the standard deviation associated with 

each measured quantity; and 

v. Necessary data are provided for performing material control tests.

NUREG-1718 13.2-6
Safeguards

The applicant's measurement program should be found acceptable if it meets the criteria 

specified in Chapter 4.3 of Revision 1 to NUREG-1 280. The following documents also 

provide additional guidance/information on measurement methods: NUREG-0228, 

"Calorimetric Assay of Plutonium;" NUREG-0256, "Methods for the Accountability of Mixed 

Oxide;" NUREG/CR-0602, "Active Nondestructive Assay of Nuclear Materials;" 

NUREG/CR-2078, "Handbook of Nuclear Safeguards Measurement Methods," September 

1983; and NUREG/CR-5550, "Passive Nondestructive Assay of Nuclear Materials." 

D. Measurement Control 

Section 74.59(e) requires that licensees ensure the quality of measurement systems and 

material processing practices. The applicant's measurement control program should 

include: 

i. Performing engineering analyses and evaluations on all MC&A measurement systems; 

ii. Establishing and verifying procedures for mixing and sampling source material, SNM, 

and SSNM and maintaining sample integrity during transport and storage; 

iii. Generating current data on the performance of measurement processes; 

iv. Using the measurement control data for the estimation of standard errors of inventory 

difference (SEID) and the standard deviation associated with the process differences; 

v. Ensuring SEID is less than 0.1 percent of the active inventory; 

vi. Applying bias corrections in accordance with approved written procedures; 

vii. Investigating and taking corrective actions when the associated measurement biases 

exceed limits; and 

viii. Establishing and maintaining a statistical control system to monitor the quality of each 

type of program measurement. 

The measurement control program applies to measurement systems used for inventory, 

shipper-receiver measurement, monitoring cumulative shipper-receiver differences, and 

detection and response purposes. In addition, the applicant should ensure the traceability 

of calibration and control standard measurements to a national standard or nationally 

accepted measurement system. 

The applicant's measurement control program should be found acceptable if it meets the 

criteria specified in Chapter 4.4 of Revision 1 to NUREG-1280. NUREG/CR-4604 and TID 

26298, "Statistical Methods in Nuclear Material Control," 1973, provide additional guidance 

on measurement error standard deviation. 

E. Physical Inventory

13.2-7 NUREG-1718
Safeguards

Section 74.59(f) contains the basic requirements for scheduling, performing, and evaluating 

physical inventories. The applicant's physical inventory program should ensure that it 

provides for: 

i. Performing a physical inventory at least every 6 calendar months (unless otherwise 

required to satisfy 10 CFR Part 75); 

ii. Within 45 days after the start of the ending inventory: 

a. Calculating inventory difference (ID) and estimating SEID; 

b. Investigating, resolving, and reporting excessive ID and SEID; 

c. Reconciling and adjusting the book inventory; and 

d. Performing reinventory as necessary. 

iii. Implementing policies, practices, and procedures designed to ensure the quality of 

physical inventories; and 

iv. Controlling and maintaining records and documentation associated with the physical 

inventories. 

The applicant should appropriately describe the procedures and/or processes for verifying 

the location and identity of all quantities of SSNM and for verifying that all quantities are 

based on measurements, inventory cutoff and cutoff verification, and reconciliation. It is 

critical that the applicant demonstrate its ability to eliminate holdup before physical inventory 

and to measure holdup if it cannot be eliminated. 

The applicant's physical inventory program should be found acceptable if it meets the 

criteria specified in Chapter 4.5 of Revision 1 to NUREG-1280. NUREG/BR-0096, 

"Instructions and Guidance for Completing Physical Inventory Summary Reports," provides 

additional guidance on completing NRC Form 327, "SNM and SM Physical Inventory 

Summary Report." 

F. Accounting 

Section 74.59(g) requires that licensees establish auditable records sufficient to 

demonstrate that the requirements of 10 CFR 74.51, 74.53, 74.55, 74.57, and 74.59 have 

been met. The applicant's accounting programs should establish and maintain records in 

an auditable form, available for inspection, for at least 3 years, unless a longer retention 

time is required by 10 CFR Part 75. The programs should specify in what form those 

records will be kept. The programs should provide adequate safeguards against tampering 

with and loss of records. (SRP Chapter 11.8 provides additional guidance on records 

management.) 

The applicant's programs for recordkeeping should be found acceptable if they meet the 

criteria specified in Chapter 4.6 of Revision 1 to NUREG-1280. NUREG/BR-0006,

NUREG-1718 13.2-8
Safeguards

"Instructions for Completing Nuclear Material Transaction Reports and Concise Note Forms 

(Form DOE/NRC 741, 741 A, and 740M)," and NUREG/BR-0007, "Instructions for 

Completing Nuclear Balance Report and Physical Inventory Listing (Forms DOE/NRC 742 

and 742C)," provide additional guidance on the use of NRC-required forms for reporting 

transactions involving nuclear materials. 

G. Shipments and Receipts 

Section 74.59(h)(1) requires that licensees establish procedures for the measurement of 

shipments and receipts and for the review, evaluation, and investigation of shipper-receiver 

differences (SRD). The applicant should establish a program to timely and accurately 

quantify the content of SSNM and other nuclear materials in shipments and receipts. The 

program should provide: 

i. Accurate identification and measurements of the quantity shipped and received; 

ii. Clear definition of statistically significant SRD; 

iii. Review and evaluation of SRD; 

iv. Investigation and corrective actions when SRD exceed the specified limit; and 

v. Documentation of SRD evaluations, investigations, and corrective actions. 

The program should identify a reasonable time frame for completing the verification 

measurements of receipts. The documentation of shipments and receipts should be 

completed and transmitted within the time frame specified in NUREG/BR-0006. The 

applicant's program for shipper-receiver comparisons should be found acceptable if it meets 

the criteria specified in Chapter 4.7 of Revision 1 to NUREG-1 280. 

H. Scrap Control 

Section 74.59(h)(2) establishes requirements regarding the segregation of internally 

generated scrap from scrap received from other nuclear facilities and regarding the prompt 

recovery of scrap that cannot be measured to within ±5 percent. The applicant's scrap 

control program should ensure that: 

i. Internally generated scrap and scrap from other licensees or contractors are segregated 

until accountability is established; and 

ii. Any scrap measured with a standard deviation greater than 5 percent of the measured 

amount is recovered, so that the results are segregated by inventory period and 

received within 6 months of the end of the inventory period in which the scrap was 

generated, except where it can be demonstrated that the scrap measurement 

uncertainty will not cause noncompliance with 10 CFR 74.59(e)(5). 

iii. Scrap and waste will be stored only in approved locations and disposed only by 

approved methods;

13.2-9 NUREG-1718
Safeguards

iv. The facility's recovery capability is adequate to preclude buildup of excess amounts of 

scrap; 

v. Special handling procedures for waste or independent measurement verification are 

described; 

vi. Scrap generated onsite and offsite are adequately separate, and the individuals 

performing measurements on scrap materials have the authority to reject containers that 

demonstrably violate segregation practices; and 

vii. Procedures and processes for offsite scrap recovery are discussed. 

The applicant's scrap control program should be found acceptable if it meets the criteria 

specified in Chapter 4.8 of Revision 1 to NUREG-1280. 

I. Human Errors 

Section 74.59(h)(3) requires that the licensees incorporate checks and balances in the 

MC&A system to control the rate of human errors in MC&A information. The applicant's 

program should minimize human errors in the following areas: 

i. The development and management of MC&A procedures, especially procedures for 

processing MC&A data; 

ii. The use of job performance aids, such as illustrations and graphs; 

iii. The methods and technologies used to automate MC&A functions; and 

iv. The quality control system used to monitor the frequency and types of human errors. 

The applicant's human error controls should be found acceptable if they meet the format 

and criteria specified in Chapter 4.9 of Revision 1 to NUREG-1280. (SRP Chapter12 

provides additional guidance on human factors.) 

J. Independent Assessment 

Section 74.59(h)(4) of Title 10 requires that the licensees independently assess the past 

performance of MC&A program. The applicant's audit and assessment program should be 

acceptable if it: 

i. Independently assesses the effectiveness of the MC&A system at least every 12 

months; 

ii. Documents the results of the assessment; 

iii. Documents management's findings on whether the MC&A system is effective;

NUREG-1718 13.2-10
Safeguards

iv. Documents any actions taken on recommendations from prior assessments; and 

v. Assesses the measurement control program of any outside contractor laboratory 

performing MC&A measurements for the applicant. 

The selection of assessment team members should ensure and balance independence and 

knowledge in the MC&A area. An assessment by a third-party organization is not required, 

but is often an effective way to bring both knowledge and independence to the assessment 

effort (SRP Section 11.6 provides additional guidance on audits & assessments). 

The applicant's program for assessing and reviewing the MC&A program should be found 

acceptable if it meets the criteria in Chapter 4.10 of Revision 1 to NUREG-1280. 

K. SSNM Custodianship 

Section 74.59(h)(5) establishes requirements for assigning custodial responsibility for 

SSNM. The applicant's assignment of custodial responsibility should ensure that such 

responsibility is clearly defined and can be effectively executed. The applicant's SSNM 

custodial assignments should be found acceptable if they meet the criteria specified in 

Chapter 4.11 of Revision 1 to NUREG-1280. 

13.2.5 REVIEW PROCEDURES 

13.2.5.1 Acceptance Review 

The primary reviewer should perform an acceptance review to determine if the license 

application adequately addresses the items in Section 13.2.3, "Areas of Review," for either the 

construction approval review or the review for a license to possess and use SNM. 

Guidance specific to the construction approval review and the review for a license to possess 

and use SNM is provided below. 

A. Construction Approval 

Specifically, the safeguards assessment of the design basis should address Section 13.2.3 

at the level of commitments and program goals. 

B. License To Possess and Use SNM 

Specifically, the application should address Section 13.2.3 in full and should include the 

fundamental nuclear material control plan (FNMCP). The secondary and supporting 

reviewers should confirm that the FNMCP is consistent with descriptions in other sections of 

the application. Information provided in the FNMCP should be of comparable quality and

13.2-11 NUREG-1718
Safeguards

detail, and should not contradict or adversely affect information contained in other sections 

of this application. 

If the primary reviewer verifies that MC&A is adequately addressed (construction approval 

review or the license to posses and use SNM), the primary reviewer should accept the 

application for the safety evaluation in Section 13.2.5.2. If the primary reviewer identifies 

significant deficiencies in the material provided, the primary reviewer should request that the 

applicant submit additional information prior to the start of the safety evaluation. 

13.2.5.2 Safety Evaluation 

After determining that the application is acceptable for review in accordance with either 

Section 13.2.5.1(A) (construction approval) or 13.2.5.1(B) (license to possess and use SNM), 

the primary reviewer should perform a safety evaluation against the acceptance criteria 

described in Section 13.2.4. On the basis of its review, the staff may request that the applicant 

provide additional information or modify the application to meet those acceptance criteria. 

Guidance specific to the construction approval review and the review for a license to possess 

and use SNM is provided below. 

A. Construction Approval 

The primary reviewer should establish that the applicant's design basis for MC&A and 

related commitments will lead to an FNMCP that will meet or exceed the regulatory 

acceptance criteria in Section 13.2.4. 

B. License To Possess and Use SNM 

The primary reviewer should establish that the applicant's FNMCP provides reasonable 

assurance in satisfying the acceptance criteria in Section 13.2.4. The primary reviewer 

should also ensure that adequate documentation is provided. 

For an existing facility, the NRC reviewers may wish to visit the site and hold discussions with 

facility personnel to gain a better understanding of the safeguards systems. For a planned 

facility, the NRC reviewers may wish to meet with the design team to gain a better 

understanding of the process, its potential safeguards concerns, and safeguards system/design 

approaches. 

When the evaluation is complete, the primary reviewer, with assistance from other reviewers, 

should prepare input for the SER as described in Section 13.2.6 using the acceptance criteria 

from Section 13.2.4.

NUREG-1718 13.2-12
Safeguards

13.2.6 EVALUATION FINDINGS 

The primary reviewer should document the safety evaluation by preparing material suitable for 

inclusion in the SER. The primary reviewer should describe the review, explain the basis for the 

findings, and state the conclusions. 

The staff could document the safety evaluation for the application for the construction approval 

review as follows: 

The staff reviewed the license application for construction approval for [insert name of 

facility] according to Section 13.2 of NUREG-1718. The staff evaluated [Insert a summary 

statement of what was evaluated] and found that [summarize the findings]. 

The staff concluded that the applicant provided adequate commitments and goals for the 

design basis as it applies to material control and accounting and that these commitments 

and goals should result in a MC&A program and an FNMCP that will meet or exceed the 

regulatory acceptance criteria outlined in NUREG-1 718. As a result, the applicant meets 

the requirements in the area of MC&A to approve construction of the proposed facility under 

10 CFR Part 70. 

The staff could document the safety evaluation for the review for the license to possess and 

use SNM as follows: 

The staff reviewed the license application for [insert facility name] to possess and use SNM 

according to Section 13.2 of NUREG-1718. The staff evaluated [Insert a summary 

statement of what was evaluated] and found [insert a description of the findings]. Based on 

the review of the license application, the NRC staff concluded that the applicant's FNMCP 

satisfies the staff's acceptance criteria. Specifically, the applicant has satisfactorily 

addressed the applicable regulatory requirements in 10 CFR 74.51, 74.53, 74.55, 74.57, 

and 74.59. 

13.2.7 REFERENCES 

Code of Federal Regulations, Title 10, Energy, Part 74, Subpart E, "Formula Quantities of 

Strategic Special Nuclear Material." 

Jaech, John L. Statistical Methods in Nuclear Material Control. TID-26298. Washington, D.C: 

Technical Information Center, U.S. Atomic Energy Commission. 1973. 

Nuclear Regulatory Commission (U.S.) (NRC). NUREG-0228, "Calorimetric Assay of 

Plutonium." NRC: Washington, D.C. May 1977. 

* NUREG-0256, " Methods for the Accountability of Mixed Oxide." NRC: Washington, 

D.C. April 1977.

13.2-13 NUREG-1718
Safeguards

* NUREG-1280, Rev. 1, "Standard Format and Content Acceptance Criteria for the 

Material Control and Accounting (MC&A) Reform Amendment." NRC: Washington, D.C. 

April 1995. 

NUREG/BR-0006, Revision 3, "Instructions for Completing Nuclear Material 

Transaction Reports and Concise Note Forms (Forms DOE/NRC 741, 741A, and 740M)." 

NRC: Washington, D.C. January 1989. 

NUREG/BR-0007, Revision 2, "Instructions for Completing Nuclear Balance Report 

and Physical Inventory Listing (Forms DOE/NRC 742, and 742C)." NRC: Washington, D.C. 

July 1989. 

• NUREG/BR-0096, "Instructions and Guidance for Completing Physical Inventory 

Summary Reports (NRC Form 327)." NRC: Washington, D.C. October 1992. 

NUREG/CR-0602, "Active Nondestructive Assay of Nuclear Materials." NRC: 

Washington, D.C. January 1981. 

NUREG/CR-2078, "Handbook of Nuclear Safeguards Measurement Methods." NRC: 

Washington, D.C. September 1983. 

NUREG/CR-4604, "Statistical Methods for Nuclear Material Management." NRC: 

Washington, D.C. December 1988. 

NUREG/CR-5550, "Passive Nondestructive Assay of Nuclear Materials." NRC: 

Washington, D.C. March 1991. 

13.2.8 DEFINITIONS 

formula kilogram (FKG): SSNM in any combination in a quantity of 1,000 grams computed by 

the formula: grams = (grams contained U-235) + 2.5 (grams U-233 + grams plutonium). 

formula quantity: SSNM in any combination in a quantity of 5,000 grams or more computed 

by the formula: grams = (grams contained U-235) + 2.5 (grams U-233 + grams plutonium).

NUREG-1718 13.2-14
14.0 EMERGENCY MANAGEMENT 

14.1 PURPOSE OF REVIEW 

The purpose of this review is to determine if the applicant established, before the start of 

operations, adequate emergency management facilities and procedures to protect the public, 

the workers, and the environment. The applicant should also show how the emergency 

management facilities and procedures comply with NRC regulations while coexisting with the 

U.S. Department of Energy's (DOE's) emergency planning requirements, as applicable. 

An emergency plan is required when an evaluation shows that the maximum dose to a member 

of the public offsite due to a release of radioactive materials would exceed 0.01 Sv (1 rem) 

effective dose equivalent. This section applies to facilities authorized to possess enriched 

uranium (U) or plutonium (Pu) for which a criticality accident alarm system is required, uranium 

hexafluoride (UF 6) in excess of 50 kg (110 Ib) in a single container or 1,000 kg (2,200 Ib) total, 

or Pu in excess of 2 Ci in unsealed form or on foils or plated sources. 

Emergency capability is incorporated into the baseline design criteria of 10 CFR Part 70, as 

revised, and is intended to ensure control of licensed material, evacuation of personnel, and 

availability of emergency facilities. 

14.2 RESPONSIBILITY FOR REVIEW 

Primary: Emergency Preparedness Specialist 

Secondary: Project Manager 

SuDporting: Regional Emergency Preparedness Inspector 

Fuel Facility Inspection staff 

14.3 AREAS OF REVIEW 

NRC staff should review the applicant's submittal for an acceptable level of evidence of 

planning for emergency preparedness directed at situations involving real or potential 

radiological hazards. The review should address those design features, facilities, functions, 

and equipment that may affect some aspect of emergency planning or the capability of an 

applicant to cope with plant emergencies. In addition, the review should address coordination 

with offsite organizations. The staff should either review the emergency plan made in 

accordance with 10 CFR 70.22(i)(1)(ii) and with the guidance contained in the acceptance 

criteria below, or should review the applicant's evaluation that demonstrates that the maximum 

dose to a member of the public would not exceed 0.01 Sv (1 rem) effective dose equivalent in 

accordance with 10 CFR 70.22(i)(1)(i). 

The NRC staff reviewer should review the material presented, as described below.

14.0-1 NUREG-1718
Emergency Management

14.3.1 Specific Items to be Reviewed When the Applicant Submits an Evaluation 

If the applicant submits an evaluation to demonstrate that the maximum dose to a member of 

the public would not exceed 0.01 Sv (1 rem) effective dose equivalent, staff should review the 

evaluation against 10 CFR 70.22(i)(1)(i), and NUREG-1 140, "A Regulatory Analysis of 

Emergency Preparedness for Fuel Cycle and Other Radioactive Material Licensees." 

NUREG/CR-641 0, "Nuclear Fuel Cycle Facility Accident Analysis Handbook," also contains 

useful information. Areas to be evaluated should include the following: 

A. A description of the facility and proposed licensed activities; 

B. Types of materials used, including both radioactive material and hazardous chemicals; 

C. Types of accidents; 

D. Detection of accidents; 

E. Site-specific information used to support the evaluation; 

F. An evaluation of the consequences, both onsite and offsite; and 

G. One or more of the factors provided in 10 CFR 70.22(i)(2). 

14.3.2 Specific Items to be Reviewed When the Applicant Submits an Emergency 

Plan 

If the applicant submits an emergency plan, staff should evaluate the emergency plan against 

10 CFR 70.22(i)(1)(ii) and Regulatory Guide 3.67, "Standard Format and Content for 

Emergency Plans for Fuel Cycle and Materials Facilities," which provides a standard format and 

content for an emergency plan. Elements in the emergency plan should include: 

A. Facility description (including both onsite and offsite emergency facilities); 

B. Types of accidents; 

C. Classification of accidents; 

D. Detection of accidents; 

E. Mitigation of consequences (and safe shutdown); 

F. Assessment of releases (both radioactive materials and hazards chemicals); 

G. Responsibilities of applicant; 

H. Notification and coordination; 

I. Information to be communicated and parties to be contacted; 

J. Training; 

K. Safe shutdown (recovery and plant restoration); 

L. Exercises and drills; 

M. Hazardous chemicals inventories and locations; and 

N. Responsibilities for developing and maintaining the emergency program and its procedures.

NUREG-1718 14.0-2
Emergency Management

14.4 ACCEPTANCE CRITERIA 

14.4.1 Regulatory Requirements 

10 CFR Part 70.22(i)(1)(i) specifies when an emergency plan does not have to be submitted to 

the NRC and, if an emergency plan is required to be submitted, 10 CFR Part 70.22(i)(3) 

contains the information that must be included in the emergency plan. 

10 CFR Part 70.64(a)(6) requires that applicants address the control of licensed material, 

evacuation of personnel, and availability of onsite emergency facilities that facilitate the use of 

available offsite services. 

14.4.2 Regulatory Guidance 

Regulatory guidance for preparing an emergency plan includes: 

Nuclear Regulatory Commission (U.S.) (NRC). NUREG-1 140, "A Regulatory Analysis of 

Emergency Preparedness for Fuel Cycle and Other Radioactive Materials." NRC: Washington, 

D.C. 1988. 

NUREG/CR-6410, "Nuclear Fuel Cycle Facility Accident Analysis Handbook." NRC: 

Washington, D.C. 1998. 

Regulatory Guide 3.67, "Standard Format and Content for Emergency Plans for Fuel 

Cycle and Materials Facilities." NRC: Washington, D.C. 1992. 

14.4.3 Regulatory Acceptance Criteria 

If the applicant's proposed total possession limit for radioactive material exceeds the 

emergency plan threshold in 10 CFR 70.22(i)(1), the applicant may submit either a site-specific 

evaluation that demonstrates maximum public exposure is less that the limits in 70.22(i)(1)(i), or 

an emergency plan. If the applicant submits an evaluation, the regulatory acceptance criteria in 

Section 14.4.3.1 apply. If the applicant submits an emergency plan, the regulatory acceptance 

criteria in Section 14.4.3.2 apply. 

14.4.3.1 Evaluation 

The adequacy of the applicant's evaluation that the maximum dose to a member of the public 

offsite due to a release of radioactive materials would not exceed 0.01 Sv (1 rem) or an intake 

of 2 mg (7.04x10-5 ounces) of soluble uranium should be evaluated by the reviewer against the 

requirements in 10 CFR Part 70.22(i)(2) and the specific criteria given in this section of the 

SRP. The applicant's evaluation should be acceptable if the regulatory requirements and the 

following criteria are met:

14.0-3 NUREG-1718
Emergency Management

14.4.3.1.1 Facility Description 

The applicant's evaluation includes a description of the facility and site, the area near the site, 

and the licensed activities conducted at the facility sufficient to support the evaluation. The 

facility description should be acceptable if it includes: 

A. A detailed drawing of the site showing (1) onsite and near offsite (within 1.6 km [1 mile]) 

structures with building numbers and labels, (2) roads and parking lots onsite and main 

roads near the site, (3) site boundaries showing fences and gates, (4) major site features, 

(5) water bodies within approximately 1.6 km (1 mile), and (6) the location(s) of nearest 

residence(s); 

B. The stack heights, typical stack flow rates, and the efficiencies of any emission control 

devices; and 

C. A general description of the proposed licensed and other major activities conducted at the 

facility, and the type, form, solubility, and maximum quantities of radioactive and other 

hazardous material normally onsite. 

14.4.3.1.2 Types of Accidents 

The applicant's evaluation describes each type of accident identified by the Integrated Safety 

Analysis (ISA) Summary that has the maximum offsite consequences exceeding the limit of 

10 CFR 70.22(i)(1)(i). The types of accidents should be acceptable if they include: 

A. The process and physical location where each accident could occur; 

B. Complicating factors and possible onsite and offsite consequences, including 

nonradioactive hazardous material released; and 

C. The accident sequence that has the potential for the greatest radiological and toxic 

chemical impact. 

14.4.3.1.3 Detection of Accidents 

The applicant's evaluation should be acceptable if, for each type of accident identified, the 

applicant identifies: 

A. The means of detecting the accident; 

B. The means of detecting any release of radioactive or other hazardous material; 

C. The means of alerting the operating staff; and 

D. The anticipated response of the operating staff.

NUREG-1718 14.0-4
Emergency Management

14.4.3.1.4 Maximum Public Exposure 

In addition to the acceptance criteria in Sections 14.4.3.1-14.4.3.3, the applicant's evaluation 

should be acceptable if it includes a description of the following information sufficient to allow 

the primary reviewer to independently verify the calculations: 

A. Type of accident (e.g., fire, exposure, chemical release, nuclear criticality); 

B. Location of accident; 

C. Maximum source term; 

D. Solubility of material; 

E. Facility design or engineered safety features in the facility and the proposed release 

fraction; 

F. Location and distance of the nearest member of the public to the facility; 

G. Dose model used and the process used to verify the reliability of the model and the validity 

of the assumptions; 

H. Assumed worst case weather condition; and 

I. Maximum calculated dose to a member of the public at the facility boundary. 

The applicant's site-specific evaluation should include a list and a description of the factors in 

10 CFR 70.22(i)(2) that the applicant considered in evaluating maximum dose to members of 

the public. The applicant should demonstrate why the factors used in the evaluation are 

appropriate when compared with the factors in NUREG-1 140. If the factors and evaluation 

show that the maximum dose to a member of the public offsite due to a release of radioactive 

materials could not exceed 0.01 Sv (1 rem) effective dose equivalent or the intake of soluble 

uranium of 2 mg (7.04x1 0"- ounces), no emergency plan is required in accordance with 

10 CFR 70.22(i)(1)(i). If the primary reviewer finds that the maximum dose to a member of the 

public could exceed 0.01 Sv (1 rem), the applicant must either submit an emergency plan 

consistent with the requirements in Section 14.4.3.2, or decrease the total possession limit for 

radioactive material below the emergency plan threshold in 10 CFR 70.22(i)(1). 

14.4.3.2 Criteria When an Emergency Plan is Required 

The adequacy of the applicant's proposed emergency plan should be evaluated by the reviewer 

against the requirements in 10 CFR Part 70.22(i)(3) and the specific criteria given in 

Section 14.4.3.2 of the SRP. The applicant's emergency plan should be acceptable if the 

regulatory requirements and the criteria in the following sections are met.

14.0-5 NUREG-1 718
Emergency Management

14.4.3.2.1 Facility Description 

14.4.3.2.1.1 Operational Facilities 

The applicant's emergency plan includes a description of the facility and site, the area near the 

site, and the licensed activities conducted at the facility sufficient to support emergency 

management activities. The description should be acceptable if it includes: 

A. A detailed drawing of the site showing: 

i. Onsite and near offsite (within 1.6 km [1 mile]) structures with building numbers and 

labels; 

ii. Roads and parking lots onsite and main roads near the site; 

iii. Site boundaries showing fences and gates; 

iv. Major site features; and 

v. Water bodies within approximately 1.6 km (1 mile). 

B. A general area map (approximately 16 km [10 mile] radius), a United States Geological 

Survey topographical quadrangle (71/2 minute series; including the adjacent quadrangle(s) if 

the site is located less than 1.6 km [1 mile] from the edge of the quadrangle), and a map or 

aerial photograph indicating onsite structures and near-site structures (about 1.6 km [1 mile] 

radius). The general area map indicates the location of sensitive facilities near the site, 

such as hospitals, schools, nursing homes, nearest residence(s), fire departments, prisons, 

environmental sampling locations, and other structures and facilities important to 

emergency management. 

C. The stack heights, typical stack flow rates, and the efficiencies of any emission control 

devices. 

D. A general description of licensed and other major activities conducted at the facility and the 

type, form, and quantities of radioactive and other hazardous materials normally onsite by 

location (use and storage) and building, including the hazardous characteristics (exposure 

rates, pH, temperature, and other characteristics) important to emergency management. 

E. Certification that the applicant has met responsibilities under the Emergency Planning and 

Community Right To Know Act of 1986, Title III, Public Law 99-499, in accordance with 

10 CFR 70.22(i)(3)(xiii).

NUREG-1718 14.0-6
Emergency Management

14.4.3.2.2 Onsite and Offsite Emergency Facilities 

The applicant's emergency plan includes a list and description of onsite and offsite facilities that 

could be relied upon in the event of an emergency. The onsite and offsite emergency facilities 

should be acceptable if they include: 

A. A list and description of both onsite and offsite emergency facilities by location and purpose 

of the facility. 

B. A description of emergency monitoring equipment that is available for personnel and area 

monitoring, as well as that for assessing the release of radioactive or hazardous materials 

to the environment. 

C. A description of the onsite and offsite services that support emergency response 

operations, including: 

i. Decontamination facilities; 

ii. Medical treatment facilities; 

iii. First aid personnel; 

iv. Firefighters; 

v. Law enforcement assistance; and 

vi. Ambulance services. 

D. In addition, the applicant's emergency facilities, equipment, and resources are ready to 

support emergency response operations, including: 

i. Facilities of adequate size and appropriate location that are designated, equipped, and 

ready for emergency use; 

ii. Adequate backup facilities required by the emergency plan and supporting documents 

that are available and ready for use; 

iii. Appropriate equipment and supplies necessary to support emergency response 

activities that are accessible during accident conditions; 

iv. Emergency equipment that is inventoried, tested, and serviced on a periodic basis to 

ensure accountability and reliability; 

v. Sufficient reliable primary and backup communications channels that are available to 

accommodate emergency needs; 

vi. Offsite emergency resources and services that are identified and are ready to ensure 

their timely mobilization and use; 

vii. Operational engineering information, such as current as-built drawings and procedures, 

that are readily available in the emergency facilities;

14.0-7 NUREG-1 718
Emergency Management

viii. Sufficient equipment for personnel protection and monitoring; and 

ix. Systems in place to alert onsite and offsite personnel in the event of an emergency. 

14.4.3.2.3 Types of Accidents 

The applicant's emergency plan includes a description for each accident identified by the ISA 

Summary (e.g., by reference) for which protective actions may be needed. The types of 

accidents should be acceptable if they include: 

A. The process and physical location(s) where accidents could occur; 

B. Complicating factors and possible onsite and offsite consequences, including 

nonradioactive hazardous material releases that could affect emergency response efforts; 

C. The accident sequence that has the potential for the greatest radiological and toxic 

chemical impact; and 

D. Figure(s) projecting dose and toxic substance concentration as a function of distance and 

time for various meteorological stability classes. 

14.4.3.2.4 Classification of Accidents 

The applicant's emergency classification system for classifying events at the facility should be 

acceptable if it includes: 

A. The following two event classifications: 

i. "Alert": Events that may occur, are in progress, or have occurred that could lead to a 

release of radioactive material or hazardous chemicals incident to the process, but the 

release is not expected to require a response by an offsite response organization to 

protect persons offsite; and 

ii. "Site area emergency": Events that may occur, are in progress, or have occurred that 

could lead to a significant release of radioactive material or hazardous chemicals 

incident to the process that could require a response by offsite emergency response 

organizations to protect persons offsite. 

B. For each accident in the emergency plan, the classification (alert or site area emergency) 

that is expected for each accident is identified. 

C. The emergency plan specifies emergency action levels (EALs) at which an alert or site area 

emergency will be declared. EALs are specific conditions that require emergency response 

measures to be performed. The applicant's EALs are consistent with Appendix A of 

Regulatory Guide 3.67 and are compared with the Environmental Protection Agency's

NUREG-1718 14.0-8
Emergency Management

Protective Action Guides. Transportation accidents more than 1.6 km (1 mile) from the 

facility are not classified. 

D. The emergency plan designates the personnel positions and alternates with the 

responsibility for accident classification during normal and backshift hours. 

14.4.3.2.5 Detection of Accidents 

The emergency plan should be acceptable if it describes, for each type of accident identified: 

A. The means of detecting the accident; 

B. The means of detecting any release of radioactive or other hazardous material; 

C. The means of alerting the operating staff; and 

D. The anticipated response of the operating staff. 

14.4.3.2.6 Mitigation of Consequences 

The applicant's emergency plan should be acceptable if it adequately describes mitigation of 

consequences, including the following: 

A. The emergency plan describes for each accident identified, adequate measures and 

equipment for safe shutdown and for mitigating the consequences to workers onsite and 

offsite as well as to the public offsite. 

B. For impending danger from an accident initiator, the emergency plan describes the 

following: 

i. The criteria that will be used to determine whether a single process or the entire facility 

will be shut down; 

ii. The steps that will be taken to ensure a safe, orderly shutdown of a single process or 

the entire facility; 

iii. The approximate time required to accomplish a safe shutdown of processes; and 

iv. The compensatory measures required for safety during the shutdown period following 

an accident. 

14.4.3.2.7 Assessment of Releases 

The applicant's emergency plan should be acceptable if it describes how the applicant 

assesses any radioactive material or hazardous chemical releases, including: 

A. The applicant's procedures to promptly and effectively assess the release of radioactive 

material or hazardous chemicals associated with the processing of radioactive material, 

including:

14.0-9 NUREG-1718
Emergency Management

i. The procedures for estimating or measuring the release rate or source term; 

ii. Valid computer codes used to project doses or concentrations to the public or 

environment and associated assumptions, along with adequate justifications to show the 

validity of the assumptions; 

iii. The types, methods, frequencies, implementation time, and other details of onsite and 

offsite sampling and monitoring that will be performed to assess a release of radioactive 

material or hazardous chemicals; and 

iv. The method for assessing collateral damage to the facility, especially safety controls. 

B. The applicant's procedure for validating any code used to assess releases of radioactive 

material or hazardous chemicals. 

14.4.3.2.8 Responsibilities 

The applicant's emergency plan should be acceptable if it describes the emergency response 

organization and administration that ensures effective planning, implementation, and control of 

emergency preparedness activities and meets the following criteria: 

A. The organizational structure and chain of command are clearly defined; 

B. Staffing and resources are sufficient to accomplish assigned tasks; 

C. Responsibilities and authority for each management, supervisory, and professional position 

are clearly defined. Responsibility is assigned for the coordination of onsite and offsite 

radiation/hazardous material emergency response preparedness; 

D. Interfaces with supporting groups, both onsite and offsite, are clearly defined; 

E. Mutual cooperation agreements exist with local agencies such as fire, police, 

ambulance/rescue, and medical units; 

F. Plant management measures include audit and assessment (SRP Section 15.6) of 

emergency preparedness to ensure site readiness to handle emergencies and to identify 

and correct problems; 

G. The onsite emergency response organization as described provides reasonable assurance 

of effective command and control of the site during the assessment, mitigation, and 

recovery phase of an accident; 

H. The emergency public information staff provides advance and ongoing information to the 

media and public on subjects that would be discussed during an emergency, such as 

radiation hazards, chemical hazards, site operation, and site emergency plans; and

NUREG-1718 14.0-10
Emergency Management

i. The schedule of emergency preparedness procedure development provides for availability 

of procedures to support startup and operation of new processes/facilities onsite. 

14.4.3.2.9 Notification and Coordination 

The applicant's emergency plan should be acceptable if it adequately describes the applicant's 

notification and coordination procedures, including: 

A. Reasonable assurance that emergency notification procedures will enable the emergency 

organization to correctly classify emergencies, notify emergency response personnel, and 

initiate or recommend appropriate actions in a timely manner, based on the following: 

i. Classification of emergency events are based on the current emergency plan; 

ii. Notification procedures minimize distractions of shift operating personnel and include 

concise, preformatted messages. Appropriate followup messages to offsite authorities 

are issued in a timely manner; 

iii. Information on the nature and magnitude of the hazards is made available to 

appropriate emergency response personnel; 

iv. Radiological and chemical source term data are available to the command post, 

technical support center, emergency operations center, and appropriate State 

personnel, in cooperation with NRC; 

v. When available, offsite field monitoring data are logged, compared with source term 

data, and used in the protective action recommendation process; 

vi. Protective Action Guides are available and used by appropriate personnel in a timely 

manner; 

vii. The emergency public information program ensures timely dissemination of accurate, 

reliable, and understandable information; 

viii. Systems are in place, if required, to alert, notify, and mobilize onsite and offsite 

response personnel in the event of an emergency; 

ix. Notification of and coordination with responsible parties when some personnel, 

equipment, and facility components are not available. 

B. How and by whom the following actions will promptly and effectively be taken: 

i. Decision to declare an alert or site area emergency; 

ii. Activation of onsite emergency response organization during all shifts;

14.0-11 NUREG-1718
Emergency Management

iii. Prompt notification of offsite response authorities that an alert or site area emergency 

has been declared, including the licensee's initial recommendation for offsite protective 

actions (normally within 15 minutes); 

iv. Notification to the NRC Operations Center (as soon as possible and, in any case, no 

later than1 hour after a declared emergency); 

v. Decision on what onsite protective actions to initiate; 

vi. Decision on what offsite protective actions to recommend; 

vii. Decision to request support from offsite organizations; and 

viii. Decision to terminate the emergency or enter recovery mode. 

14.4.3.2.10 Information To Be Communicated 

The applicant's emergency plan should be acceptable if it describes the information to be 

communicated during an emergency and includes: 

A. A standard reporting checklist to facilitate timely notification; 

B. The types of information to be provided concerning facility status, radioactive or hazardous 

chemical releases, and protective action recommendations; 

C. A description of preplanned protective action recommendations to be made to each 

appropriate offsite organization; 

D. The offsite officials to be notified, as a function of the classification of the event; and 

E. The recommended actions to be implemented by offsite organizations for each accident 

treated in the emergency plan. 

14.4.3.2.11 Training 

The applicant's emergency plan includes an adequate training program for onsite and offsite 

emergency response personnel to ensure knowledge of the emergency plan, assigned duties, 

and effective response to an actual emergency. The training program should be acceptable if it 

includes: 

A. The topics and general content of training programs used for training the onsite and offsite 

emergency response personnel to satisfy the objectives described above;

NUREG-1 718 14.0-12
Emergency Management

B. The administration of the training program, including responsibility for training, the positions 

to be trained, the schedules for training, the frequency of retraining, the use of team 

training, and the estimated number of hours of initial training and retraining; 

C. The training to be provided on the use of protective equipment such as respirators, 

protective clothing, monitoring devices, and other equipment used in emergency response; 

D. The training program for onsite personnel who are not members of the emergency 

response staff; and 

E. The instructions and tours that will be offered to fire, police, medical, and other emergency 

personnel to the extent necessary commensurate with the results of the ISA. 

14.4.3.2.12 Safe Shutdown (Recovery and Plant Restoration) 

The applicant's emergency plan describes the plans for adequately restoring the facility to a 

safe status after an accident and recovery after an emergency. The safe shutdown should be 

acceptable if it includes: 

A. Appropriate methods and responsibilities for assessing the damage to and the status of the 

facility's capabilities to safely control radioactive material or hazardous chemicals 

associated with the process; 

B. Procedures for promptly determining the actions necessary to reduce any ongoing releases 

of radioactive or other hazardous chemicals and to prevent further incidents; 

C. Provisions for promptly and effectively accomplishing required restoration action; and 

D. Descriptions of the key positions in the recovery organization. 

14.4.3.2.13 Exercises and Drills 

The applicant's emergency plan commits to conducting exercises and drills in a manner that 

demonstrates the capability of the organization to plan and perform an effective response to an 

emergency. The commitment should be acceptable if it demonstrates that: 

A. Task-related knowledge is demonstrated through periodic participation by all qualified 

individuals for each position in the emergency response organization; 

B. Drill performance is assessed against specific scenario objectives using postulated 

accidents that adequately test personnel, equipment, and resources, including previously 

identified weaknesses; 

C. Effective player, controller, evaluator, and observer predrill briefings are conducted;

14.0-13 NUREG-1718
Emergency Management

D. Scenario data and exercise messages provided by the controllers effectively maintain the 

timeline and do not interfere with the emergency organization's response to exercise 

scenario events, except where safety considerations are involved; 

E. Trained evaluators are used to identify and record participant performance, scenario 

strengths and deficiencies, and equipment problems; 

F. Prestaging of equipment and personnel is minimized to realistically test the activation and 

staffing of emergency facilities; 

G. Critiques are conducted in a timely manner and include a followup plan for correcting 

identified weaknesses and improving training effectiveness; 

H. Emergency drills demonstrate that resources are effectively used to control the site, to 

mitigate further damage, to control radiological/chemical releases, to perform required 

onsite activities under simulated radiation/airborne and other emergency conditions, to 

provide accurate assessments and status during an accident, and to initiate recovery; 

I. Emergency drills demonstrate personnel protection measures, including controlling and 

minimizing hazards to individuals during events such as fires, medical emergencies, 

mitigation activities, search and rescue, and other similar events; 

J. The emergency drill demonstrates that onsite communications effectively support 

emergency response activities; 

K. The emergency drill demonstrates that the emergency public information organization 

disseminates accurate, reliable, timely, and understandable information; 

L. Provisions are made for conducting quarterly communications checks with offsite response 

organizations; and 

M. Offsite organizations are invited to participate in the biennial onsite exercise that tests the 

major elements of the emergency plan and response organizations. 

14.4.3.2.14 Responsibilities for Developing and Maintaining the Emergency 

Program and Its Procedures 

The applicant's emergency plan describes the responsibilities for developing and maintaining 

the emergency program and its procedures. The responsibilities should be acceptable if they 

include: 

A. The means for ensuring that the revisions to the emergency plan and the procedures that 

implement the emergency plan are adequately prepared, kept up to date normally (within 

30 days of any changes), and distributed to all affected parties, including the NRC.

NUREG-1718 14.0-14
Emergency Management

B. The provisions for approving the implementing emergency procedures, making and 

distributing changes to the procedures, and ensuring that each person responsible for an 

emergency response function has immediate access to a current copy of emergency 

procedures. 

C. The provisions for approval of changes to the emergency plan and procedures and a list of 

those individuals authorized to make these changes; 

D. Procedures for allowing offsite response organizations 60 days to comment on the 

emergency plan (except those changes allowed by 10 CFR 70.69(o)) before submitting it to 

the NRC, and for providing NRC any comments received within 60 days along with the plan; 

and 

E. Procedures for modifying the emergency plan in accordance with 10 CFR 70.32(i). 

14.5 REVIEW PROCEDURES 

14.5.1 Acceptance Review 

The primary reviewer should perform an acceptance review to determine if the application 

(construction or license) adequately addresses the items in Section 14.1.3, "Areas of Review." 

Guidance specific to the construction approval review and the review for the license to possess 

and use special nuclear material (SNM) is provided below. 

A. Construction Approval 

The applicant is not expected to submit either an emergency plan as described in 

Section 14.3.2 or an evaluation as described in Section 14.3.1 with the portion of the license 

application submitted for the construction approval review. However, the primary reviewer 

should evaluate the safety assessment of the design basis to ensure that the commitments 

and program goals are appropriate for emergency protection at the design stage. 

B. License To Possess and Use SNM 

Specifically, the license application should either contain an evaluation as described in 

Section 14.3.1 or an emergency plan as described in Section 14.3.2. 

If the primary reviewer verifies that emergency protection is adequately addressed in the 

application for construction approval or the license application, the primary reviewer should 

accept the application for the safety evaluation in Section 13.1.5.2. If the primary reviewer 

identifies significant deficiencies in the material provided, the primary reviewer should request 

that the applicant submit additional information prior to the start of the safety evaluation.

14.0-15 NUREG-1718
Emergency Management

14.5.2 Safety Evaluation 

After determining that the application is acceptable for review in accordance with either 

Section 14.5.1 (A) (construction approval) or 14.5.1 (B) (license to possess and use SNM), the 

primary reviewer should perform a safety evaluation against the acceptance criteria described 

in Section 14.4. On the basis of its review, the staff may request that the applicant provide 

additional information or modify the application to meet those acceptance criteria. 

Guidance specific to the construction approval review and the review for a license to possess 

and use SNM is provided below. 

A. Construction Approval 

The primary reviewer should ensure that the design basis includes appropriate 

commitments for emergency protection at the design stage. For example, if the safety 

assessment of the design basis shows a dose to a member of the public that exceeds the 

limits in 10 CFR 70.22(i)(1)(i), the applicant should commit to providing an emergency plan 

with the license application. 

B. License To Possess and Use SNM 

i. No Emergency Plan 

The primary reviewer should verify that the applicant's evaluation is consistent with the 

potential accident sequences described in the ISA Summary. The ISA reviewer and the 

primary reviewer should coordinate to ensure the resolution of any issues concerning 

the evaluation relative to ISA information. The final step for the primary reviewer should 

be to prepare a Safety Evaluation Report (SER) in accordance with Section 14.6 which 

either agrees with the applicant's conclusion that no emergency plan is required or 

indicates that the staff does not accept the applicant's evaluation and recommends that 

an emergency plan be required by the applicant. 

ii. Emergency Plan 

After it is determined that an acceptable application containing an emergency plan has 

been received from the applicant, the primary reviewer should conduct a complete 

review of the emergency plan and determine its acceptability in accordance with 

Section 14.4.3.2. The reviewer should verify that emergency planning is consistent with 

the potential accident sequences described in the ISA Summary. The ISA reviewer and 

emergency plan reviewer should coordinate to ensure the resolution of any issues 

concerning the emergency plan relative to ISA information. This information may be 

supplemented by a personal visit to the site by the primary reviewer and meetings with 

the applicant. The final step for the primary reviewer should be to prepare an SER in 

accordance with Section 14.6, "Evaluation Findings."

NUREG-1718 14.0-16
Emergency Management

14.6 EVALUATION FINDINGS 

The primary reviewer should document the safety evaluation by preparing material suitable for 

inclusion in the SER. The primary reviewer should describe the review, explain the basis for the 

findings, and state the conclusions. 

The staff could document the safety evaluation for the construction approval review as follows: 

The staff evaluated the portion of the license application submitted for construction approval 

for [insert facility name] in accordance with Chapter 14.0 of NUREG-1718. The staff 

evaluated [insert a summary statement of what was evaluated] and found that [insert a 

summary statement of the findings]. The NRC staff determined that the applicant's 

commitments, including the commitment to provide an emergency plan with the license 

application [if the applicant's design basis safety assessment shows it is required], are 

adequate to meet the requirements for construction approval in accordance with 

10 CFR Part 70. 

The staff could document the safety evaluation for the review for the license to possess and 

use SNM, where the applicant submits an emergency plan, as follows: 

The staff evaluated the emergency plan submitted as part of the license application for 

[insert facility name] to possess and use SNM in accordance with Chapter 14.0 of 

NUREG-1718. The staff evaluated [insert a summary statement of what was evaluated] 

and found that [insert a summary statement of the findings]. In accordance with 

10 CFR 70.22(i), the licensee commits to maintaining and executing an emergency plan for 

responding to the radiological hazards resulting from a release of radioactive material and to 

any associated chemical process hazards. NRC staff determined that the applicant's 

emergency plan is adequate to demonstrate compliance with 10 CFR 70.22(i), including: 

(1) the plant is properly configured to limit releases of radioactive materials in the event of 

an accident, (2) a capability exists for measuring and assessing the significance of 

accidental releases of radioactive materials, (3) appropriate emergency equipment and 

procedures are provided onsite to protect workers against radiation and other chemical 

hazards that might be encountered following an accident, (4) a notification system has been 

established for notifying Federal, State, and local government agencies and recommending 

appropriate protective actions to protect members of the public, and (5) necessary recovery 

actions are established for returning the plant to a safe condition following an accident. The 

requirements of the emergency plan are implemented through approved written procedures. 

Changes that decrease the effectiveness of the emergency plan may not be made without 

NRC approval. The NRC will be notified of other changes that do not decrease the 

effectiveness of the emergency plan within 6 months of the changes. 

The NRC staff concluded that the applicant's emergency plan meets the requirements of 

10 CFR 70.22(i).

14.0-17 NUREG-1718
Emergency Management

14.7 REFERENCES 

Environmental Protection Agency (U.S.) (EPA). EPA 400-R-92-001, "Manual of Protective 

Action Guides and Protective Actions for Nuclear Incidents." EPA: Wahsington, D.C.1992. 

Nuclear Regulatory Commission (U.S.), Washington, D.C. "Part 30 Statements of 

Consideration and Emergency Preparedness for Fuel Cycle and Other Radioactive Material 

Licensees," Federal Register: Vol. 54, p. 14051. 1989. 

Nuclear Regulatory Commission (U.S.) (NRC). NUREG-1 140, "A Regulatory Analysis of 

Emergency Preparedness for Fuel Cycle and Other Radioactive Materials." NRC: Washington, 

D.C. 1988. 

NUREG/BR-0150, Vol. 1, Rev. 4, RTM-96, "Response Technical Manual." NRC: 

Washington, D.C. 1996. 

NUREG/CR-641 0, "Nuclear Fuel Cycle Accident Analysis Handbook." NRC: 

Washington, D.C. 1998. 

Regulatory Guide 3.67, "Standard Format and Content for Emergency Plans for Fuel 

Cycle and Materials Facilities." NRC: Washington, D.C. 1992.

NUREG-1718 14.0-18
15.0 MANAGEMENT MEASURES 

15.1 QUALITY ASSURANCE 

15.1.1 PURPOSE OF REVIEW 

The purpose of this review is to establish that the applicant has a quality assurance (QA) 

program that will provide reasonable assurance against natural phenomena and the 

consequences of potential accidents through the QA program's application to the design, 

fabrication, construction, testing, and operation of the applicant's structures, systems, and 

components1 (SSCs); the applicant is required to describe the QA program for construction 

approval under 10 CFR 70.22(f). This review also establishes that the applicant has a QA 

program that will provide reasonable assurance that all items relied on for safety2 (IROFS) will 

be available and reliable to perform their designated safety functions when needed, which the 

applicant is required to describe as part of its license application under 10 CFR Part 70. 

15.1.2 RESPONSIBILITY FOR REVIEW 

Primar: QA Engineer/Specialist 

Secondary: Project Manager 

Supportinc: Fuel Cycle Facility Inspector 

Primary Reviewers of applicable SRP Chapters 5.0 through 15.0 

15.1.3 AREAS OF REVIEW 

The applicant is required to submit a description of the QA program for construction approval 

and should update the QA program when it applies for a license to possess and use special 

nuclear material (SNM). The areas of review should include: 

1

The QA program should be appropriately applied to all SSCs pending design process 

identification of the applicable QA or other controls and designation of principal SSCs. 

"Principal structures, systems, and components" are, by definition, IROFS (see Footnote 2). 

For the purposes of the review guidance provided under this section, references to IROFS are 

intended to include the principal SSCs identified in the application for construction approval (see 

10 CFR 70.4 or the glossary to this SRP). 

2"Items relied on for safety" is defined in 10 CFR Part 70, as revised, as "structures, systems, 

equipment, components, and activities of personnel that are relied on to prevent potential 

accidents at the facility that could exceed the performance requirements specified in § 70.61 or 

to mitigate their potential consequences."

15.1-1 NUREG-1718
Management Measures

A. Organization. 

B. QA Function3. 

C. Design control. 

D. Procurement document control. 

E. Instructions, procedures4 , and drawings. 

F. Document control. 

G. Control of purchased items. 

H. Identification and control of items. 

I. Control of special processes. 

J. Inspection. 

K. Test control. 

L. Control of measuring and test equipment. 

M. Handling, storage, and shipping. 

N. Inspection, test, and operating status. 

0. Nonconformances. 

P. Corrective action. 

Q. QA records. 

R. Audits and assessments5 . 

S. Applicant's provisions for continuing QA. 

15.1.4 ACCEPTANCE -CRITERIA 

15.1.4.1 Regulatory Requirements 

The regulation, 10 CFR Part 70, requires that the applicant establish an appropriate QA 

program to ensure that all IROFS perform their designated safety functions and are continually 

available and reliable. The regulatory requirements for QA are addressed in the following: 

Nuclear Regulatory Commission (U.S.), Washington, D.C. "Domestic Licensing of 

Special Nuclear Material (10 CFR Part 70)." Federal Register: Vol. 64, No. 146. 

pp. 41338-41357. July 30, 1999. 

In addition, an applicant to possess and use special nuclear material in a plutonium processing 

and fuel fabrication facility such as the mixed oxide (MOX) fuel fabrication facility is required, 

3SRP Section 15.4 addresses training and qualification of plant personnel. Section G2 of SRP 

Appendix G on QA addresses training and qualification of other personnel. 

4SRP Section 15.5 addresses plant procedures. Section G5 of SRP Appendix G on QA 

addresses other procedures. 

5Guidance for audits and assessments is given in SRP Section 15.6 as referenced in SRP 

Appendix G on QA.

NUREG-1718 15.1-2
Management Measures

pursuant to § 70.22(f), to describe the QA program to be applied to the design, fabrication, 

construction, testing, and operation of the structures, systems, and components of the facility. 

The footnote of § 70.22(f) states that the description of the QA program should include a 

discussion of how the criteria in Appendix B of 10 CFR Part 50 will be met. 

The footnote to § 70.23(b) states that the criteria in Appendix B of Part 50 of this chapter will be 

used by the Commission in determining the adequacy of the QA program. 

Additional pertinent regulatory requirements for identifying, controlling, and reporting of defects 

with a facility, activity, or basic component supplied to a facility licensed under 10 CFR Part 70 

are addressed in the following: 

Code of Federal Regulations, Title 10, Energy, Part 21, "Reporting of Defects and 

Noncompliances." 

15.1.4.2 Regulatory Guidance 

Guidance for QA is addressed in the following: 

American Society of Mechanical Engineers (ASME). ASME NQA-1-1994, "Quality 

Assurance Requirements for Nuclear Facility Applications, as revised by the ASME 

NQA-1 a-1 995 Addenda." ASME: New York. 1994/1995. 

Nuclear Regulatory Commission (U.S.) (NRC). Regulatory Guide 1.28, Revision 3, "Quality 

Assurance Program Requirements (Design and Construction)," NRC: Washington, D.C. 

August 1985. 

Note that while the above guidance has separate sections for "requirements" and "guidance," 

NRC's regulatory QA requirements exist only in the applicable Commission regulations. 

15.1.4.3 Regulatory Acceptance Criteria 

The NRC reviewers should find that the applicant's QA program adequately addresses and 

satisfies the regulatory acceptance criteria below. The applicant may reference material in 

other sections of the license application, or incorporate material by reference, provided these 

references are clear and specific. 

The applicant should identify the principal SSCs (construction approval review) or IROFS 

(review for a license to possess and use SNM) and the degree of their importance. The graded 

approach for the application of QA should be described unless the applicant chooses to apply 

the highest level of QA and quality control to all principal SSCs and IROFS. 

For principal SSCs (construction approval review) or IROFS (license to possess and use SNM), 

the applicant should apply either Option A or Option B (whichever the applicant chooses for the 

construction approval review) as described below.

15.1-3 NUREG-1718
Management Measures

Option A. Address the regulatory acceptance criteria given in this section and 

provide a commitment to implement and maintain the QA program in conformance 

with the applicable "requirements" of Parts I and II of ASME-NQA-I-19946, as revised 

by the ASME NQA-1 a-1 995 Addenda or equivalent. 

OR 

Option B. Address the checklist provided in SRP Appendix G on QA. 

Depending on the option chosen, the applicant should address the criteria specified below. 

That is, if Option A is used, the applicant should (a) include a commitment that it will implement 

and maintain its QA program to comply with the applicable "requirements" of ASME 

NQA-1 -1994, as revised by the ASME NQA-1 a-1 995 Addenda (that is, the basic and 

supplemental "requirements" of Parts I and II, including Subpart 2.7), or equivalent and should 

(b) be responsive to the five regulatory acceptance criteria given below. 

A. Organization 

The applicant should describe the organizational structure and functional responsibilities 

and provide charts of the lines of responsibilities, interrelationships, and areas of 

responsibility and authority for all organizations performing activities relied on for safety, 

including the applicant's organization and, if applicable, the organization of the applicant's 

principal contractors (architect/engineer, constructor, construction manager, and/or 

operator). Persons or organizations responsible for ensuring that appropriate QA has been 

established and verifying that activities affecting quality/safety have been correctly 

performed should have sufficient authority, access to work areas, and organizational 

independence to carry out their responsibilities. 

B. QA Function 

QA should be well documented, planned, implemented, and maintained to ensure the 

availability and reliability of controls relied on for safety. It should be implemented during all 

phases of the facility's life. It should be functional prior to performing the Integrated Safety 

Analysis required by 10 CFR Part 70, as revised. 

C. Applicant's Provisions for Continuing QA 

The applicant's provisions for continuing QA should address review and updates based on 

reorganizations, revised activities, lessons learned, changes to applicable regulations, and 

other QA changes. 

6

This SRP section refers to regulatory QA requirements and ASME-NQA-1 "requirements." 

Regulatory QA requirements are given in the Part 70, as revised. ASME-NQA-1 "requirements" are the Basic and Supplementary Requirements given in Parts I and II of ASME 

NQA-1 -1994, including Subpart 2.7, as revised by the ASME NQA-1 a-1 995 Addenda.

NUREG-1718 15.1-4
Management Measures

D. Management Measures 

The applicant's QA program should describe how the applicable QA criteria contained in 

Sections 15.2, 15.3, 15.4, 15.5, 15.6, 15.7, and 15.8 of this review plan will be met. 

E. Regulatory Guide 1.28 

The applicant should address the appropriate guidance and NRC positions for determining 

if its QA program adequately meets the requirements of Appendix B to 10 CFR Part 50 as 

described in RG 1.28, "Quality Assurance Program Requirements (Design and 

Construction)." 

If Option B , as appropriate, is used, the application should address the checklist items in 

SRP Appendix G on QA, and the additional reviewer guidance and NRC positions for 

determining if the applicant's QA program adequately meets the requirements of Appendix 

B to 10 CFR Part 50 as described in RG 1.28, "Quality Assurance Program Requirements 

(Design and Construction)." 

"If the applicant chooses to apply graded QA to principal SSCs, its QA program needs to 

describe the following four essential elements of the graded QA process (this applies whether 

the applicant selects Option A or Option B): 

A. Categorization of SSCs: A process that determines the safety significance of SSCs in a 

reasonable and consistent manner, including the use of both traditional engineering 

(deterministic methods) and probabilistic evaluations. 

B. Identification of QA Controls: The identification and implementation of appropriate QA 

controls for principal SSCs, or groups of SSCs, according to the safety function and safety 

significance. These controls need to maintain reasonable confidence in equipment 

performance and to support the graded QA corrective action feedback process. 

C. Feedback Mechanisms: Provisions for a feedback process to adjust graded QA controls 

need to be described as well as provisions for reassessing the QA controls when new 

information becomes available through adverse trends or nonconformance reporting. 

Provisions for an effective root cause analysis and corrective action process as a result of 

the feedback process should be described. Provisions should also be described for 

evaluating common cause/mode failures. The licensee corrective action efforts should 

determine, as a minimum, the apparent cause of repetitive failures of SSCs under the 

graded QA controls so that it can be decided whether graded QA controls should be 

adjusted. In some instances, a failure may result in an unanticipated event and may cause 

the categorization of the SSC to be changed.

15.1-5 NUREG-1718
Management Measures

D. Reassessing Safety Significance: A means for reassessing SSC safety significance and 

QA controls when new information becomes available through construction and operating 

experience, or from changes in plant design. 

The applicant should also commit to update the QA program to reflect any changes between 

the construction approval and the license application to possess and use SNM. 

In either case, the requirements of 10 CFR Part 21 should be addressed by the applicant. 

Exceptions or alternatives to the QA acceptance criteria and positions contained in this review 

plan section may be adopted by the applicant or licensee provided adequate justification is 

given. 

15.1.5 REVIEW PROCEDURES 

15.1.5.1 Acceptance Review 

The primary reviewer should perform an acceptance review to determine if the application 

adequately addresses the items in Section 15.1.3, "Areas of Review," for either the construction 

approval review or the review for a license to possess and use SNM. 

Guidance specific to the construction approval review and the review for a license to possess 

and use SNM is provided below. 

E. Construction Approval 

Specifically, the application for construction approval should address Section 15.1.3 in full 

and should identify whether Option A or Option B of Section 15.1.4.3 has been chosen. 

F. License To Possess and Use SNM 

The areas of review for the updated material in the license application should include items 

A through S identified in Section 15.1.3., with special attention on the identification of any 

new or changed aspects of the QA program. 

Note that the applicant's commitment to implement and maintain its QA in conformance with the 

applicable basic and supplemental "requirements" of Parts I and II of ASME-NQA-1-1994 or 

equivalent should satisfy the acceptance review criteria in item A or B of this section. 

If the primary reviewer verifies that QA is adequately addressed in either the construction 

approval review or the review for a license to possess and use SNM, the primary reviewer 

should accept the application for the safety evaluation in Section 15.1.5.2. If the primary 

reviewer identifies significant deficiencies in the material provided, the primary reviewer should 

request that the applicant submit additional information prior to the start of the safety 

evaluation.

NUREG-1718 15.1-6
Management Measures

15.1.5.2 Safety Evaluation 

After determining that the application is acceptable for review in accordance with either 

Section 15.1.5.1 (A) (construction approval) or 15.1.5.1 (B) (license to possess and use SNM), 

the primary reviewer should perform a safety evaluation against the acceptance criteria 

described in Section 15.1.4. On the basis of its review, the staff may request that the applicant 

provide additional information or modify the application to meet those acceptance criteria. 

Guidance specific to the construction approval review and the review for a license to possess 

and use SNM is provided below. 

A. Construction Approval 

The primary reviewer should review material submitted for the construction approval review 

to determine whether the applicant has met either Option A or Option B as defined in 

Section 15.1.4.3. 

In either case, the applicant should also (a) describe how the QA will be graded for items of 

lesser or no effect on consequences of concern (unless the applicant chooses to apply the 

highest level of QA and quality control to all principal SSCs) and (b) list the principal SSCs 

as determined in the safety assessment of the design basis. The primary reviewer should 

determine whether the applicant and its principal contractors have adequately planned for 

QA to be accomplished. Some of the information may be referenced to other sections of 

the application, or incorporated by reference, provided these references are clear and 

specific. 

The secondary reviewer should confirm that the applicant's and the applicant's principal 

contractors' QA commitments are consistent with other sections of the application. 

The other supporting reviewers should determine, within their areas of review, whether 

principal SSCs have been specified with the appropriate level of QA. 

The review should result in a determination that there is reasonable assurance that the 

applicant's and the applicant's principal contractors' QA programs will provide reasonable 

assurance against natural phenomena and the consequences of potential accidents through 

the QA program's application to the design, fabrication, construction, testing and operation 

of the applicant's SSCs. 

B. License To Possess and Use SNM 

When the applicant updates the QA program for the license application, new or changed 

material should include any IROFS identified since the NRC approved construction of the 

principal SSCs. The primary reviewer should focus the review on any new or changed 

material and determine Whether the necessary QA policies, procedures, and instructions will 

be in place and applied to IROFS before personnel begin activities relied on for safety. The

15.1-7 NUREG-1718
Management Measures

primary reviewer should also confirm that the material presented remains consistent with 

the material provided in the license application in support of other chapters of this SRP. 

The supporting reviewer (Fuel Cycle Facility Inspector) should become familiar with the 

applicant's and principal contractors' QA commitments and determine whether ongoing 

activities are in agreement with them. 

The review should result in a determination that there is reasonable assurance that the 

applicant's and the applicant's principal contractors' QA programs will provide reasonable 

assurance that IROFS will be available and reliable to perform their safety functions in a 

satisfactory manner when needed. 

When the safety evaluation is complete, the primary reviewer, with assistance from the other 

reviewers, should prepare the QA input for the Safety Evaluation Report (SER) as described in 

Section 15.1.6 using the acceptance criteria from Section 15.1.4. The secondary reviewer 

should coordinate the QA input with the balance of the reviews and the SER. 

15.1.6 EVALUATION FINDINGS 

The primary reviewer should document the safety evaluation by preparing material suitable for 

inclusion in the SER. The primary reviewer should describe the review, explain the basis for the 

findings, and state the conclusions. 

The staff could document a safety evaluation for the application for construction approval 

review as follows: 

The staff reviewed the quality assurance (QA) program for construction approval for 

[insert facility name] according to Chapter 15.1 of NUREG-1 718. [Here the primary 

reviewer provides a summary statement of what was evaluated and why the reviewer 

finds the application acceptable.] Based on its review of the application, the NRC staff 

concluded that the applicant has adequately described its QA program and the 

applicant's QA program meets the regulatory requirements of 10 CFR Part 70, as 

applied to structures, systems, and components, will provide reasonable assurance of 

protection against natural phenomena and the consequences of potential accidents. 

The staff could document a safety evaluation for the review for a license to possess and use 

SNM as follows: 

The staff reviewed the quality assurance (QA) program for a license for [insert facility name] 

to possess and use special nuclear material according to Chapter 15.1 of NUREG- 1718. 

[Here the primary reviewer provides a summary statement of what was evaluated and why 

the reviewer finds the application acceptable.] Based on its review of the license 

application, focusing on new or updated material when compared to the safety evaluation 

for the construction approval review, the NRC staff concludes that (A) the applicant has 

adequately described its updated QA program and (B) the applicant's updated QA program

NUREG-1718 15.1-8
Management Measures

meets the regulatory requirements of 10 CFR Part 70 and thus provides reasonable 

assurance that all items relied on for safety will be available and reliable to perform their 

designated safety functions when needed. 

15.1.7 REFERENCES 

American Society of Mechanical Engineers (ASME). ASME-NQA-1-1994, as revisied by the 

ASME NQA-1 a-1 995 Addenda, "Quality Assurance Requirements for Nuclear Facility 

Applications." ASME: New York, NY.1994/1995. 

Code of Federal Regulations, Title 10, Energy, Part 21, "Reporting of Defects and 

Noncompliances." 

Code of Federal Regulations, Title 10, Energy, Part 70, "Domestic Licensing of Special Nuclear 

Material." 

Nuclear Regulatory Commission (U.S.), Washington, D.C. "Domestic Licensing of Special 

Nuclear Material (10 CFR Part 70)," Federal Register: Vol. 64, No. 146. pp. 41338-41357. 

July 30, 1999. 

Nuclear Regulatory Commission (U.S.) (NRC). Regulatory Guide 1.28, Revision 3, "Quality 

Assurance Program Requirements (Design and Construction)," NRC: Washington, D.C. 

August 1985.

15.1-9 NUREG-1718
15.0 MANAGEMENT MEASURES 

15.2 CONFIGURATION MANAGEMENT 

15.2.1 PURPOSE OF REVIEW 

The purpose of this review is to establish with reasonable assurance that the applicant has a 

plan for or has implemented an acceptable configuration management (CM) system. The 

review should result in a determination that the applicant has described and committed to a CM 

system during design and construction (as described in the portion of the license application 

submitted for construction approval) and operations (as updated and described in the 

application for a license to possess and use special nuclear material [SNM]) that provides 

reasonable assurance that the applicant will maintain in a consistent and up-to-date manner 

design information, safety information, and modifications (both temporary and permanent for 

design and operations) information, that might affect the ability of the principal structures, 

systems, or components' (SSCs) or items relied on for safety (IROFS) to perform their function 

when needed. The review should also result in a determination that the applicant's CM system 

captures formal documentation governing the design and continued maintenance of the SSCs 

and IROFS and supporting management measures, as identified and described in the 

integrated safety analysis (ISA) programmatic commitments and ISA Summary (see 

Chapter 5.0). The review should ensure that the CM system is adequately coordinated and 

integrated with the other management measures, such as maintenance, quality assurance, 

training and qualifications, procedures, and audits and assessments. 

15.2.2 RESPONSIBILITY FOR REVIEW 

Primary: Project Manager 

Secondary: Primary ISA Reviewer, Quality Assurance Reviewer, Records 

Management Reviewer, Organization and Administration Reviewer 

Supporting: Fuel Cycle Facility Inspector 

15.2.3 AREAS OF REVIEW 

The applicant should submit a description of the CM system with the portion of the license 

application that it submits for construction approval and should submit updated information with 

the application for a license to possess and use SNM. The applicant's descriptions and 

commitments for CM should be reviewed with an emphasis on the processes for documenting 

an established baseline configuration and controlling changes to it to preclude inadvertent 

degradation of safety. An examination should be conducted of the descriptions of the 

organizational structure responsible for CM activities and the process, procedures, and 

documentation required by the applicant for modifying SSCs, principal SSCs, and IROFS and 

"1 "Principal structures, systems, and components" are, by definition, items relied on for safety (see 

10 CFR 70.4 or the glossary to this SRP).

15.2-1 NUREG-1 718
Management Measures

the supporting management measures. The review should focus on the applicant's 

management level controls that ensure (a) the disciplined documentation of engineering, 

installation, and operation of modifications; (b) the training and qualification of affected staff; 

(c) revision and distribution of operating, test, calibration, surveillance, and maintenance 

procedures and drawings; (d) post-modification testing; and (e) operational readiness review. 

The following topics should be reviewed: 

A. CM Policy 

The review should cover the applicant's description of overall CM systems, including at least 

the following topics: (a) the scope of the SSCs and IROFS to be included in the CM 

system, (b) objectives of each CM activity, (c) a description of each CM activity, and (d) the 

organizational structure and staffing interfaces. 

The review should examine the applicant's establishment of a baseline CM policy applicable 

to all design and construction (construction approval review) and operations (review for a 

license to possess and use SNM). The review should also examine any reduced level of 

CM that the applicant may propose for certain principal SSCs or IROFS based on the safety 

assessment of the design bases or ISA results, respectively. 

Specifically, the primary reviewer should review the CM plan that provides management 

commitments and policy directives and defines key responsibilities, terminology, and 

equipment scope. The method for initiating prompt corrective actions should be examined. 

The secondary reviewers should examine the safety assessment of the design bases 

(construction approval review) or the ISA Summary (review for a license to possess and use 

SNM) for the identification of dependence on CM of principal SSCs or IROFS. Appropriate 

interfaces both within the CM system and with other facility organizations and functions 

should be examined. In particular, the quality assurance (QA) reviewer should assist in 

examining the functional interfaces with QA, maintenance, and training (including 

qualification). The reviewers should look for the applicant's identification of required data 

bases and the rules for their maintenance. The reviewers should examine implementing 

procedures for the CM system. 

B. Design Requirements 

The review should cover the applicant's demonstration that design requirements and 

associated design bases have been established and are maintained by an appropriate 

organizational unit. The applicant's CM controls on the design requirements and the safety 

assessment of the design bases (construction approval review) or the ISA (review for a 

license to possess and use SNM) should be evaluated. The review should be coordinated 

with the primary reviewer of Chapter 5.0. 

C. Document Control

NUREG-1718 15.2-2
Management Measures

The review should include the applicant's methods used to establish and control documents 

within the CM system. 

D. Change Control 

The review should examine the applicant's commitments to provide reasonable assurance 

that the CM system maintains strict consistency among the design requirements, the 

construction or physical configuration, and the facility documentation. An important 

component of this review is the applicant's process, within the CM system, for ensuring that 

the safety assessment of the design bases (construction approval review) or the ISA (review 

for a license to possess and use SNM) will be systematically reviewed and modified to 

reflect design or operational changes from an established safety basis, and that all other 

documents that are affected by safety basis changes will be properly modified, 

authoritatively approved, and made available to personnel. 

E. Assessments 

The review should examine the applicant's commitments to conduct initial and periodic 

assessments of the CM system to determine the system's effectiveness and to correct 

deficiencies, consistent with the acceptance criteria in SRP Section 15.6, "Audits and 

Assessments." 

15.2.4 ACCEPTANCE CRITERIA 

15.2.4.1 Regulatory Requirements 

The staff's requirements applicable to CM are the following: 

A. 10 CFR 70.62(d), relating to the requirement that the applicant or licensee is to establish 

management measures to provide continuing assurance of compliance with the 

performance requirements. 

B. 10 CFR 70.64(a)(1), relating to the requirement that the design of new facilities or the 

design of new processes at existing facilities be developed and implemented in accordance 

with management measures. 

C. 10 CFR 70.65(a), relating to the requirement that the application include a description of the 

management measures. 

D. 10 CFR 70.72(a), relating to the requirement that the licensee establish a CM system.

15.2-3 NUREG-1718
Management Measures

15.2.4.2 Regulatory Guidance 

None. 

15.2.4.3 Regulatory Acceptance Criteria 

The reviewers should determine that an applicant's CM system is acceptable if it satisfies the 

following criteria: 

A. CM Policy 

The applicant's description of its overall CM system describes at least the following topics: 

(a) the scope of the SSCs (construction approval review) or IROFS (review for a license to 

possess and use SNM) and supporting management measures to be included in the CM 

system (coordinate with the reviewer of Chapter 5.0), (b) a description of each CM system 

activity, (c) the objectives of each CM system activity, (d) any reduced level of CM that the 

applicant may propose for certain SSCs (construction approval review) or IROFS (review for 

a license to possess and use SNM) based on the safety assessment of the design bases or 

the ISA results, respectively, and (e) the organizational structure and staffing interfaces. 

The scope of SSCs (construction approval review) or IROFS (review for a license to 

possess and use SNM) includes all those SSCs or IROFS as defined by the safety 

assessment of the design bases or the ISA, respectively; furthermore, those items are 

included in the QA, maintenance, and training and qualifications programs. The functional 

interfaces with QA, maintenance, and training and qualifications are of particular importance 

and should be addressed individually. 

B. Design Requirements 

The applicant demonstrates that design requirements and associated design bases have 

been established and are maintained by an appropriate organizational unit. The applicant 

demonstrates that the CM system provides for keeping design requirements and the safety 

assessment of the design bases (construction approval review) or the ISA (review for a 

license to possess and use SNM) current and that suitable hazard/accident analysis 

methods, including controlled computer codes, if applicable, are available to evaluate safety 

margins of proposed changes. Technical management review and approval procedures are 

described. 

The design process leading to drawings and other statements of requirements proceeds 

logically from the design bases. Specific personnel are assigned the responsibility for 

maintaining the design bases and requirements. These may be the same personnel that 

maintain the safety assessment of the design bases (construction approval review) or the 

ISA (review for a license to possess and use SNM) and controlled computer codes. SSCs 

(construction approval review) or IROFS (review for a license to possess and use SNM) to 

be listed under CM are clearly defined in the requirements documents, along with the 

assignment of any grades or quality levels. The grades or quality levels, if specified, are

NUREG-1718 15.2-4
Management Measures

based on the qualitative risk associated with postulated accident sequences in which the 

SSCs (construction approval review) or IROFS (review for a license to possess and use 

SNM) are required to function. The applicant should have indicated in the safety 

assessment of the design bases (construction approval review) or the ISA (review for a 

license to possess and use SNM) the level of CM attributes that are applied to a particular 

item. However, in the safety assessment of the design bases or ISA, this indication may 

consist of only an index or category designation. The definition of the multiple CM levels, if 

used, should be in the CM chapter of the application. 

C. Document Control 

The applicant describes an acceptable method to establish and control documents within 

the CM system, including cataloging the document data base, the information content of the 

document data base, maintenance and distribution of documents, document retention 

policies, and document retrieval policies. A list of the types of documents controlled is 

established and includes key documents, such as drawings, procurement specifications, 

engineering analyses, operating procedures, training/qualification records, and maintenance 

procedures. 

The applicant's material shows that the CM system will capture documents that are relevant 

and important to safety. This includes design requirements; the safety assessment of the 

design bases (application for construction) or the ISA (review for a license to possess and 

use SNM); as-built drawings; specifications; all safety-important operating procedures; 

procedures involving training, QA, maintenance, audits, and assessments; emergency 

operating procedures; emergency response plans; system modification documents; 

assessment reports; and others, as necessary, that the applicant may deem part of the CM 

system. A controlled document data base is used to control documents and track document 

change status. Rules of storage for originals or master copies of documents within the CM 

system should follow the guidance of "Records Management" discussed in SRP 

Section 15.8. 

D. Change Control 

The applicant demonstrates that the CM system will maintain strict consistency among the 

design requirements, the physical configuration, and the facility documentation. The 

applicant commits to an acceptable process for identifying and authorizing proposed 

changes; performing appropriate technical, management, and safety reviews of proposed 

changes in configurations of SSCs (construction approval review) or IROFS (review for a 

license to possess and use SNM); approving changes; tracking and implementing changes; 

and documenting changes (including placement of documentation in a document control 

center and dissemination to affected functions such as training, engineering, operations, 

maintenance, and QA). The appllicant describes an acceptable process within the CM 

system for ensuring that the safety assessment of the design bases (construction approval 

review) or the ISA (review for a license to possess and use SNM) is systematically reviewed 

and modified to reflect design or operational changes from an established safety basis, and 

that all documents outside the safety assessment of the design bases (construction

15.2-5 NUREG-1718
Management Measures

approval review) or the ISA (review for a license to possess and use SNM) that are affected 

by safety basis changes are properly modified, authoritatively approved, and made available 

to personnel. 

Post-modification testing of items (or procedure drills or walkthroughs) may be performed in 

conjunction with periodic item performance monitoring and normal maintenance functions. 

E. Assessments 

The applicant confirms that assessments, including initial and periodic examinations of the 

CM system, will be conducted to determine the system's effectiveness and to correct 

deficiencies. The applicant indicates that such assessments will be systematically planned 

and conducted in accordance with an overall facility audit and assessment program as 

described by the applicant and reviewed by the NRC in accordance with Section 15.6 of this 

SRP. 

Both document assessments and physical assessments (system walkdowns) will be 

conducted periodically to check the adequacy of the CM system. All assessments and 

follow-ups are documented. These reports can provide a supporting basis for future 

changes. Assessments will include reviews of safety systems from design requirements 

through implementation. 

The applicant should also commit to updating the CM system to reflect any changes between 

the construction approval review and application for a license. 

15.2.5 REVIEW PROCEDURES 

15.2.5.1 Acceptance Review 

The primary reviewer should perform an acceptance review to determine if the application for 

construction approval or license application adequately addresses the items in Section 15.2.3, 

"Areas of Review." 

Guidance specific to the construction approval review and the review for a license to possess 

and use SNM is provided below. 

A. Construction Approval 

The construction approval review should address each item in Section 15.2.3 with an 

emphasis on the CM for managing the design bases during design and construction. This 

should include a reviewer determination that the applicant committed to a formal CM system 

for establishing the design bases and reviewing proposed changes to SSCs. 

B. License To Possess and Use SNM

NUREG-1718 15.2-6
Management Measures

The review for a license to possess and use SNM should address each item in 

Section 15.2.3 with an emphasis on the CM for operation (e.g., procedures, maintenance, 

and training) and any new or changed material in the CM program that will arise as a part of 

the transition from design and construction (design bases) to operations (integrated safety 

analysis). This should include a reviewer determination that the applicant committed to a 

formal CM system for establishing and managing the ISA and reviewing proposed changes 

to IROFS or items, procedures, and processes that may affect IROFS. 

If the primary reviewer verifies that CM is adequately addressed in the construction approval 

review or the review for a license to possess and use SNM, the primary reviewer should accept 

the application for the safety evaluation in Section 15.2.5.2. If the primary reviewer identifies 

significant deficiencies in the material provided, the primary reviewer should request that the 

applicant submit additional information prior to the start of the safety evaluation. 

15.2.5.2 Safety Evaluation 

After determining that the application is acceptable for review in accordance with either 

Section 15.2.5.1 (A) (construction approval review) or 15.2.5.1 (B) (review for a license to 

possess and use SNM), the primary reviewer should perform a safety evaluation against the 

acceptance criteria described in Section 15.2.4. On the basis of its review, the staff may 

request that the applicant provide additional information or modify the application to meet those 

acceptance criteria. 

Guidance specific to the construction approval review and the review for a license to possess 

and use SNM is provided below. 

A. Construction Approval 

The primary reviewer should determine whether the applicant has adequately planned for 

CM to be accomplished during design and construction and whether necessary policies, 

personnel, procedures, and instructions will be in place to begin CM early, that is, during the 

safety assessment of the design bases and the design and construction of the SSCs. The 

secondary reviewers should confirm that the applicant's CM commitments are consistent 

with other sections of the application. 

B. License To Possess and Use SNM 

When the applicant updates the CM system for the review for a license to possess and use 

SNM, the primary reviewer should focus the review on any new or changed material. 

Particularly, the primary reviewer should ensure that the applicant has adequately planned 

for CM to be accomplished during operations and whether necessary policies, personnel, 

procedures, and instructions will be in place to transition from CM during design and 

construction to CM during operations, that is, from the safety assessment of the design 

bases and the design and construction of the SSCs to the ISA and the IROFS.

15.2-7 NUREG-1718
Management Measures

The primary reviewer should also confirm that the material presented remains consistent 

with the material provided in the license application in support of other chapters of this SRP. 

The supporting reviewer (Fuel Cycle Facility Inspector) should become familiar with the 

applicant's CM commitments and determine whether ongoing activities are in agreement 

with them. 

The review for a license to possess and use SNM should result in a determination that there 

is reasonable assurance that the CM system will provide additional assurance that IROFS 

will perform satisfactorily in service and that activities relied on for safety will be performed 

satisfactorily. 

When the safety evaluation is complete, the primary staff reviewer, with assistance from the 

other reviewers, should prepare the CM input for the safety evaluation report (SER) as 

described in Section 15.2.6 using the acceptance criteria from Section 15.2.4. 

15.2.6 EVALUATION FINDINGS 

The primary reviewer should document the safety evaluation by preparing material suitable for 

inclusion in the SER. The primary reviewer should describe the review, explain the basis for the 

findings, and state the conclusions. 

The staff could document a safety evaluation for the construction approval review as follows: 

The staff reviewed the configuration management (CM) system for (name of facility) 

according to Section 15.2 of NUREG-1718. [Insert a summary statement of what was 

evaluated and why the reviewer finds the submittal acceptable.] 

Based on its review of the material submitted for construction approval, the NRC staff 

concluded that the applicant suitably and acceptably described its commitment to a 

proposed CM system, including the method for managing changes in procedures, facilities, 

activities, and equipment for SSCs identified in the safety assessment for the design bases. 

Management-level policies and procedures, including an analysis and independent safety 

review of any proposed activity involving SSCs, are described that will ensure that the 

relationship between design requirements, construction, and facility documentation is 

maintained as part of a new design or change in an existing design. The administrative 

control will ensure that the organizational structure, procedures, and responsibilities 

necessary to implement CM are in place or committed to; that the design requirements and 

bases are documented and supported by analyses and the documentation is maintained 

current; that documents, including drawings, are appropriately stored and accessible; that 

drawings and related documents adequately describe SSCs; that procedures adequately 

describe how the applicant will achieve and maintain strict consistency among the design 

requirements, facility construction, and facility documentation; and that methods are in place 

for suitable analysis, review, approval, and implementation of identified changes to SSCs.

NUREG-1718 15.2-8
Management Measures

In situations where the applicant proposes a graded CM system based on risk significance, the 

following can be added: 

The applicant described its approach to applying at least two levels of CM attributes to 

SSCs and identified which SSCs involve lower risk and may receive the reduced level of 

CM requirements. The applicant's proposed reduced CM features are found adequate to 

contribute to the reliability and availability of the lesser risk items relied on for safety 

identified in the application. 

The staff could document a safety evaluation for the review for a license to possess and use 

SNM using paragraphs that are similar to those used for the construction approval but that 

encompass the new or updated material when compared to the safety evaluation for the 

construction approval and address CM as applied to IROFS during operations, including 

controls to assure configuration verification, correct functional tests, accurate documentation 

for equipment and procedures, adequate methods or plans for initial and periodic examination 

of the CM system's effectiveness, and thorough assessments and follow-up reports of 

corrective actions. 

15.2.7 REFERENCES 

Code of Federal Regulations, Title 10, Energy, Part 70, "Domestic Licensing of Special Nuclear 

Material." 

Department of Energy (U.S.) (DOE). DOE-STD-1073-93-Pt.1 and -Pt.2, "DOE Standard 

Guide for Operational Configuration Management Program." DOE: Washington, D.C. 1993. 

Nuclear Regulatory Commission (U.S.), Washington, D.C. "Domestic Licensing of Special 

Nuclear Material (10 CFR Part 70)." Federal Register: Vol. 64, No.146. pp. 41338-41357. 

July 30, 1999.

15.2-9 NUREG-1718
15.0 MANAGEMENT MEASURES 

15.3 MAINTENANCE 

15.3.1 PURPOSE OF REVIEW 

The purpose of this review is to establish reasonable assurance that the facility will have an 

adequate maintenance program for items relied on for safety (IROFS)-with the exception of 

personnel activities-to ensure their availability and reliability to perform their intended safety 

functions when needed. The maintenance performed to meet the availability and reliability 

requirements for the IROFS should be commensurate with risk-levels identified in the integrated 

safety analysis (ISA) Summary. 

15.3.2 RESPONSIBILITY FOR REVIEW 

Primary: Project Manager 

Secondary: Quality Assurance, Criticality, Chemical, Fire, Radiation Protection, 

and Environmental Reviewers 

Supportinq: Fuel Cycle Facility Inspector 

15.3.3 AREAS OF REVIEW 

The applicant's description of its maintenance program should be reviewed during the license 

application with emphasis on demonstrating that I ROFS with the exception of personnel 

activities (safety controls) are inspected, calibrated, tested, and maintained so as to ensure 

their ability to perform their safety functions when needed. The safety controls should be 

identified by the ISA Summary (discussed in Chapter 5.0 of this SRP). Individual components 

and support systems for the safety controls may have to be individually maintained to ensure 

the availability and reliability of the control function. The reviewers should review the applicant's 

description of how each of the following essential components is implemented within the site 

organization: 

A. Surveillance/monitoring; 

B. Corrective maintenance; 

C. Preventive maintenance; and 

D. Functional testing.

15.3-1 NUREG-1718
Management Measures

15.3.4 ACCEPTANCE CRITERIA 

15.3.4.1 Regulatory Requirements 

The requirement for maintenance is addressed in the following: 

Nuclear Regulatory Commission (U.S.), Washington, D.C. "Domestic Licensing of 

Special Nuclear Material (10 CFR Part 70)." Federal Register: Vol. 64, No. 146. pp. 

41338-41357. July 30, 1999. 

Specific references are as follows: 

A. In § 70.4, "Definitions," the term "management measures" is defined. Maintenance is 

included as a management measure. 

B. In § 70.62(d), the applicant is required to establish management measures to provide 

continuing assurance of compliance with the performance requirements. 

C. In § 70.64(a)(1), the design of new facilities or new processes at existing facilities is 

required to be developed and implemented in accordance with management measures. 

D. In § 70.64(a)(8), inspection, testing, and maintenance are required to be addressed as one 

of the Baseline Design Criteria to provide reasonable assurance that IROFS will be 

designed to allow them to be adequately inspected, tested, and maintained to ensure their 

availability and reliability to perform their function when needed. 

E. In § 70.65(a), the application is required to include a description of the management 

measures. 

15.3.4.2 Regulatory Guidance 

None. 

15.3.4.3 Regulatory Acceptance Criteria 

For the construction approval, the applicant should commit to establishing a maintenance 

program that meets or exceeds the acceptance criteria in Section 15.3.4. 

The applicant's maintenance program should be considered acceptable (application for the 

license to possess and use special nuclear material [SNM]) if it adequately addresses the 

following: 

A. Safety Controls Identified in the ISA 

An assessment of whether components and support systems need to be individually 

maintained to ensure the availability and reliability of specific safety controls. The reliability

NUREG-1718 15.3-2
and availability of a particular item should be commensurate with the risk levels identified in 

the ISA. 

B. Essential Components 

i. Surveillance/monitoring: The surveillance/monitoring function, its responsible 

organization, and the conduct of surveillance/monitoring at specified frequencies to 

measure the degree to which safety functions or safety controls meet performance 

specifications. This activity is used in setting preventive maintenance frequencies for 

safety controls and the determination of performance trends for safety controls. How 

results from incident investigations (described in Section 15.7 of this SRP) and identified 

root causes are used to modify the affected maintenance function and eliminate or 

minimize the root cause from recurring should be addressed. For surveillance tests that 

can be done only while equipment is out of service, proper compensatory measures 

should be prescribed. 

ii. Corrective maintenance: The documented approach used to perform corrective actions 

or repairs on safety controls. The maintenance function should provide a planned, 

systematic, integrated, and controlled approach for the repair and replacement activities 

associated with identified failures of safety controls. 

iii. Preventive maintenance: A description of the preventive maintenance function that 

contains a commitment to conduct preplanned and scheduled periodic refurbishing or 

partial or complete overhaul for the purpose of providing reasonable assurance that the 

reliability and availability goals for the IROFS will continue to be met even with 

unplanned outages. This activity includes using the results of the 

surveillance/monitoring component of maintenance. Instrumentation calibration and 

testing should be addressed as part of this component. 

iv. Functional testing: A description of the functional testing function that contains a 

commitment to the functional testing, as warranted, of safety controls after corrective or 

preventive maintenance or calibration. Functional testing should be conducted using 

approved procedures that include compensatory measures while the test is being 

conducted. 

C. Work Control Methods 

A list of maintenance-related work control methods. 

D. Relationship of the Maintenance Elements to Other Management Control Sections 

Discussed in SRP Chapter 15.0 

A discussion of how the maintenance function uses, interfaces with, or is linked to these 

elements.

15.3-3 NUREG-1718
Management Measures

15.3.5 REVIEW PROCEDURES 

15.3.5.1 Acceptance Review 

The primary reviewer should perform an acceptance review to determine if the application 

adequately addresses the items in Section 15.3.3, "Areas of Review," for either the construction 

approval review or the review for a license to possess and use SNM. If the primary reviewer 

verifies that maintenance is adequately addressed for the appropriate review, the primary 

reviewer should accept the application for the safety evaluation in Section 15.3.5.2. If the 

primary reviewer identifies significant deficiencies in the material provided, the primary reviewer 

should request that the applicant submit additional information prior to the start of the safety 

evaluation. 

15.3.5.2 Safety Evaluation 

For the construction approval review, the reviewer should determine that the applicant has 

committed to a maintenance program that will meet or exceed the acceptance criteria in 

Section 15:3.4. 

For the review for a license to possess and use SNM, and after determining that the application 

is acceptable for review in accordance with Section 15.3.5.1, the primary reviewer should 

perform a safety evaluation against the acceptance criteria described in Section 15.3.4. On the 

basis of its review, the staff may request that the applicant provide additional information or 

modify the application to meet those acceptance criteria. 

The primary reviewer should establish that the applicant's maintenance program meets or 

exceeds the acceptance criteria. The primary reviewer should determine if the applicant has 

adequately planned the work to be accomplished and whether necessary policies, procedures, 

and instructions either are in place or will be in place before work starts. The primary reviewer 

should also determine that there is reasonable assurance that the applicant's quality assurance, 

configuration management, and maintenance programs, as described in SRP Sections 15.1 

through 15.3, are coordinated. 

When an applicant's maintenance program references other sections of the application, the 

primary reviewer should confirm that these sections of the application are consistent with the 

applicant's selection of acceptance criteria and the proposed method for implementation. 

The primary reviewer should coordinate with secondary staff reviewers to ensure there is no 

contradiction between maintenance and other areas of the application. The secondary staff 

reviewers should ensure that the scope of the applicant's maintenance program includes the 

IROFS that are in their primary review areas of the application. The supporting staff reviewer 

(Fuel Cycle Facility Inspector) should become familiar with the applicant's maintenance 

program and determine whether ongoing activities are in agreement with it.

NUREG-1718 15.3-4
15.3.6 EVALUATION FINDINGS

The primary reviewer should document the safety evaluation by preparing material suitable for 

inclusion in the Safety Evaluation Report (SER). The primary reviewer should describe the 

review, explain the basis for the findings, and state the conclusions. 

The staff could document the safety evaluation for the construction approval review by stating 

that the applicant has committed to establishing a maintenance program that meets or exceeds 

the acceptance criteria contained in Section 15.3.4 of NUREG-1718. 

The staff could document the safety evaluation for the review for a license to possess and use 

SNM as follows: 

The staff reviewed the license application for [insert facility name] according to Section 15.3 

of NUREG-1 718. Based on the review of the license application, the staff concluded that 

the applicant committed to maintenance of items relied on for safety with the exception of 

personnel activities (safety controls). [Insert a summary statement of what was evaluated 

and why the reviewer finds the submittal acceptable.] The applicant's maintenance 

commitments contain the basic elements to ensure availability and reliability: 

surveillance/monitoring, corrective maintenance, preventive maintenance, and functional 

testing. The applicant's maintenance function is proactive, using surveillance/monitoring 

and maintenance records to analyze equipment performance and identify the root causes of 

repetitive failures. 

In addition, the surveillance/monitoring activities described in this section of the application 

provide assurance of the validity of the ISA by examination and calibration and testing of 

equipment that monitors process safety parameters and acts to prevent or mitigate accident 

consequences. 

The maintenance function: (1) is based on approved procedures; (2) employs work control 

methods that properly consider personnel safety, awareness of facility operating groups, 

quality assurance, and the rules of configuration management; (3) links items relied on for 

safety requiring maintenance to the ISA; (4) justifies the preventive maintenance intervals in 

the terms of equipment reliability goals; (5) provides for training that emphasizes importance 

of ISA identified controls, regulations, codes, and personal safety; and (6) creates 

documentation that includes detailed records of all surveillances, inspections, equipment 

failures, repairs, and replacements. 

The staff concludes that the applicant's maintenance function meets the requirements of 

10 CFR Part 70 and provides reasonable assurance that the environment and the health 

and safety of the public are protected.

15.3-5 NUREG-1718
Management Measures 

15.3.7 REFERENCES 

Code of Federal Regulations, Title 10, Energy, Part 70, "Domestic Licensing of Special Nuclear 

Material." 

.Title 10, Energy, Section 50.65, "Requirements for Monitoring the Effectiveness of 

Maintenance at Nuclear Power Plants." 

.Title 29, Labor, Section 1910.119, "Process Safety Management of Highly Hazardous 

Chemicals." 

. Title 40, Protection of Environment, Part 68, "Risk Management Program for Chemical 

Accidental Release Prevention." 

Nuclear Regulatory Commission (U.S.), Washington, D.C. "Guidance on Management 

Controls/Quality Assurance, Requirements for Operation, Chemical Safety, and Fire Protection 

for Fuel Cycle Facilities." Federal Register: Vol. 54, No. 53. pp. 11590-11598. March 21, 

1989. 

-. '"Domestic Licensing of Special Nuclear Material (10 CFR Part 70)," Federal Register: 

Vol. 64, No. 146. pp. 41338-41357. July 30, 1999. 

Nuclear Regulatory Commission (U.S.) (NRC). Inspection Manual, Procedure 88025, 

"Maintenance and Surveillance Testing." NRC: Washington, D.C. May 23, 1984. 

Inspection Manual, Procedure 88062, "Maintenance and Inspection." NRC: 

Washington, D.C. January 1996. 

Regulatory Guide 1.160, Rev. 2, "Monitoring the Effectiveness of Maintenance at 

Nuclear Power Plants." NRC: Washington, D.C. March 1997.

NUREG-1718 15.3-6
15.0 MANAGEMENT MEASURES 

15.4 TRAINING AND QUALIFICATION OF PLANT PERSONNEL 

15.4.1 PURPOSE OF REVIEW 

The purpose of this review is to establish that there is reasonable assurance that personnel 

who perform activities relied on for safety at the plant1 will understand, recognize the 

importance of, and be qualified to perform these activities as required by 10 CFR Part 70 in a 

manner that adequately protects the public and worker health and safety and the environment. 

15.4.2 RESPONSIBILITY FOR REVIEW 

Primary: Training, Quality Assurance, or Human Factors Engineer/Specialist 

Secondary: Licensing Project Manager 

SuDporting: Fuel Cycle Facility Inspector 

15.4.3 AREAS OF REVIEW 

Personnel who perform activities relied on for safety are required by 10 CFR Part 70 to be 

trained and qualified as necessary. The applicant should train, test, and qualify personnel to 

provide adequate assurance that they understand, recognize the importance of, and are 

qualified to perform these activities in a manner that adequately protects the public and worker 

health and safety and the environment. The training and qualification should be commensurate 

with the assigned functional responsibility, authority, and activities relied on for safety of the 

respective personnel. The application of training and qualification may be graded and should 

be adequate to fulfill the objectives as identified by the licensee, especially when human factors 

are relied on for safety (see Chapter 12.0 of this SRP). Personnel at the facility should have 

the knowledge and skills necessary to start-up, operate, maintain, modify, and decommission 

the facility in a safe manner. The applicant should address the training and qualification of 

plant personnel for the construction approval review and should submit updated information for 

the application for a license to possess and use special nuclear material (SNM). 

The training, testing, and qualification of these personnel as described for the construction 

approval should be reviewed. This should include the training, testing, retesting, and 

qualification of managers, supervisors, designers, technical staff, construction personnel, plant 

operators, technicians, maintenance personnel, and other personnel whose level of knowledge 

is relied on for safety. 

1 This SRP section provides guidance for the review of information on the training and qualification of 

plant personnel who perform activities relied on for safety. Section G2 of SRP Appendix G on quality 

assurance and Supplement 2S-4 of ASME-NQA-I-1994 provide review guidance on the subject of 

training and qualification of other personnel (for example, construction personnel) who perform 

activities relied on for safety.

15.4-1 NUREG-1718
Management Measures

The following areas should be reviewed: 

A. Organization and management of training; 

B. Analysis and identification of functional areas requiring training; 

C. Position training requirements; 

D. Development of the basis for training, including objectives; 

E. Organization of instruction using lesson plans and other training guides; 

F. Evaluation of trainee learning; 

G. Conduct of on-the-job training; 

H. Evaluation of training effectiveness; 

I. Personnel qualification; and 

J. Applicant's provisions for continuing assurance, including the needs for retraining or 

reevaluation of qualification. 

15.4.4 ACCEPTANCE CRITERIA 

15.4.4.1 Regulatory Requirements 

The requirement for training and qualification is addressed in the following: 

Nuclear Regulatory Commission (U.S.), Washington, D.C. "Domestic Licensing of 

Special Nuclear Material (10 CFR Part 70)." Federal Register: Vol. 64, No. 146. pp. 

41338-41357. July 30, 1999. 

Specific references are as follows: 

A. In § 70.4, "Definitions," the term "management measures" is defined. Training and 

qualification are included as management measures. 

B. In § 70.62(d), the applicant is required to establish management measures to provide 

continuing assurance of compliance with the performance requirements. 

C. In § 70.64(a)(1), the design of new facilities or new processes at existing facilities is 

required to be developed and implemented in accordance with management measures. 

D. In § 70.65(a), the application is required to include a description of the management 

measures. 

An additional requirement for training and qualification is addressed in the following: 

Code of Federal Regulations, Title 10, Energy, Part 19, "Notices, Instructions and 

Reports to Workers: Inspection and Investigations." (The specific reference is to § 9.12, 

"Instructions to Workers.")

NUREG-1718 15.4-2
Management Measures

15.4.4.2 Regulatory Guidance 

NRC guidance applicable to training and qualification of personnel that provide guidance for 

implementing and satisfying the regulatory requirements and acceptance criteria is: 

Nuclear Regulatory Commission (U.S.) (NRC). NUREG-1220, Rev.1, "Traininig Review 

Criteria and Procedures." NRC: Washington, D.C. January 1993. 

15.4.4.3 Regulatory Acceptance Criteria 

For the construction approval, the applicant should commit to meet or exceed the acceptance 

criteria in Section 15.4.4 and to update the training and qualification of plant personnel 

descriptions to reflect any changes between the construction approval review and the review for 

a license to possess and use special nuclear material (SNM). 

The NRC reviewers should find that the applicant's submittal regarding training and qualification 

of plant personnel provides reasonable assurance that the regulatory acceptance criteria below 

are adequately addressed and satisfied. 

In addition to the regulatory acceptance criteria given below, SRP Sections 9.2.4.4 and 9.2.4.6 

provide criteria for training and qualification of plant personnel for radiation safety functions. 

A. Organization and Management of Training 

The organization and management of training of plant personnel should be acceptable if the 

training functions are organized, staffed, and managed to facilitate planning, directing, 

evaluating, and controlling a training process that fulfills the objectives for the training as 

identified by the licensee, especially where human factors are relied on for safety for start

up, operation, maintenance, and modification of the facility. The training and qualification 

should be commensurate with the assigned functional responsibility, authority, and activities 

relied on for safety of the respective personnel. The application of training and qualification 

may be graded and should be adequate to fulfill the objectives as identified by the licensee, 

especially when human factors are relied on for safety. Formal training should be provided 

for each position or activity for which the required performance is relied on for safety. The 

application should state what training will be conducted and which personnel will be 

provided this training. Training should include retraining of previously trained and qualified 

personnel based on specified criteria. 

The following commitments should be in the application regarding organization and 

management of training. 

i. Line management should be responsible for the content and effective conduct of the 

training. 

ii. The job function, responsibility, authority, and accountability of personnel involved in 

managing, supervising, and implementing training should be clearly defined.

15.4-3 NUREG-1718
Management Measures

iii. Performance-based training should be used as the primary management tool for 

analyzing, designing, developing, conducting, and evaluating training. 

iv. Training procedures should be documented and implemented to provide reasonable 

assurance that all phases of training are conducted reliably and consistently. 

v. Training documents should be linked to the configuration management system to 

provide reasonable assurance that design changes and plant modifications are 

accounted for in the training. 

vi. Exceptions from training may be granted to trainees and incumbents when justified, 

documented, and approved by management. 

vii. Auditable training records should be maintained. Training records, both programmatic 

and individual, should support management information needs and provide required 

data on each individual's training, job performance, and qualifications. (Refer to 

Section 15.8 and Appendix I for detailed guidance on records management.) 

B. Analysis and Identification of Functional Areas Requiring Training 

Analysis and identification of areas requiring training should be acceptable if the areas 

required for competent and safe job performance are identified, documented, and 

addressed by the training. 

Operations personnel, training staff, and other subject matter experts, as appropriate, 

should have conducted or should conduct a needs/job analysis to develop a valid task list 

for specific jobs. The jobs treated in this manner should include, as a minimum, those 

responsible for managing, supervising, performing, and verifying the activities relied on for 

safety and those specified in the Integrated Safety Analysis Summary (ISA; see SRP 

Chapter 5.0) that prevent or mitigate accidents. Each task selected for training (initial or 

continuing) from the facility-specific task list should be matrixed to supporting procedures 

and training materials. The facility-specific list of tasks selected for training and the 

comparison to training materials should be reviewed on an established schedule and 

updated as necessitated by changes in procedures, facility systems/equipment, or job 

scope. 

C. Position Training Requirements 

Position training requirements are acceptable if minimum requirements for positions are 

specified for candidates whose activities are relied on for safety or who perform actions that 

prevent/mitigate accident sequences described in the ISA Summary. Trainees should meet 

entry-level criteria defined for the position, including minimum educational, technical, 

experience, and physical fitness (if necessary) requirements. 

D. Development of the Basis for Training, Including Objectives

NUREG-1 718 15.4-4
Management Measures

The development of the basis for training, including the objectives, is acceptable if the basis 

identifies training content, defines satisfactory trainee performance, and identifies objectives 

from the analysis of activities and performance requirements. Objectives should state the 

knowledge, skills, and abilities the trainee should demonstrate; the conditions under which 

required actions will take place; and the standards of performance the trainee should 

achieve on completion of the training activity. 

E. Organization of Instruction Using Lesson Plans and Other Training Guides 

The organization of instruction using lesson plans and other training guides should be 

acceptable if the plans/guides are based on the required learning objectives derived from 

specific job performance requirements and the needs/job analysis. Plans/guides should be 

used for in-class training and on-the-job training and should include standards for evaluating 

proper trainee performance. Review and approval requirements should be established for 

all plans/guides and other training materials before their issue and use. 

F. Evaluation of Trainee Learning 

The evaluation of trainee accomplishment of learning should be acceptable if trainees are 

evaluated during training, when appropriate, to determine their progress toward mastery of 

job performance requirements and at the completion of training to determine their mastery 

of job performance requirements. 

G. Conduct of On-the-Job Training 

The conduct of on-the-job training should be acceptable if on-the-job training used for 

activities identified in the ISA Summary is fully described. On-the-job training should be 

conducted using well-organized and current performance-based training materials. On-the

job training should be conducted by designated personnel who are competent in the 

program standards and methods of conducting the training. Completion of on-the-job 

training should be by actual task performance. When the actual task cannot be performed 

by the trainee and is therefore "walked-down," the conditions of task performance, 

references, tools, and equipment should reflect the actual task to the extent possible. 

H. Systematic Evaluation of Training Effectiveness 

An evaluation of training effectiveness and its relation to on-the-job performance should be 

acceptable if it provides reasonable assurance that the training program conveys the 

required skills and knowledge and is used to revise the training, where necessary, based on 

the performance of trained personnel in the job setting. A comprehensive evaluation of 

individual training programs should be conducted periodically by qualified individuals to 

identify program strengths and weaknesses. Feedback from trainee performance during 

training and from former trainees and their supervisors should be used to evaluate and 

refine the training. Change actions (for example, procedure changes, equipment changes, 

and facility modifications) should be monitored and evaluated for their impact on the

15.4-5 NUREG-1718
Management Measures

development or modification of initial and continuing training and should be incorporated in 

a timely manner. Change actions should be accomplished through the configuration 

management system (see SRP Section 15.2). Improvements and changes to initial and 

continuing training should be systematically initiated, evaluated, tracked, and incorporated 

to correct training deficiencies and performance problems. 

I. Personnel Qualification 

Commitments should be provided regarding minimum qualifications for personnel required 

to meet NRC regulations. Minimum qualifications should be commensurate with the 

assigned functional responsibility and authority of the respective personnel. The following 

commitments should be in the application regarding personnel qualification for managers, 

supervisors, designers, technical staff, construction personnel, plant operators, technicians, 

maintenance personnel, and other plant staff required to meet NRC regulations: 

i. Managers should have a minimum of a B.S./B.A. or equivalent. Each manager should 

have either management experience or technical experience in nuclear facilities or 

activities similar to the mixed oxide (MOX) facility or activities that they are to manage. 

ii. Supervisors should have at least the qualifications required of personnel being 

supervised and either 1 additional year experience supervising the technical area at a 

similar facility or completion of the supervisor training. 

iii. Technical staff identified in the ISA Summary whose activities are relied on for safety to 

satisfy the performance requirements identified in 10 CFR Part 70, should have a B.S. 

or equivalent in an appropriate technical field and experience and training appropriate 

for their activities, authority, and responsibilities. 

iv. Facility operators, technicians, maintenance personnel, and other staff whose actions 

are required to comply with NRC regulations should have completed the applicant's 

training process or have equivalent experience or training. 

v. Candidates for process operators positions should be required to meet minimum 

qualifications described in the application. Candidates for job functions other than 

process operators should also be required to meet minimum qualifications, but these 

minimum qualifications need not be described in the application. 

J. Applicant's Provisions for Continuing Assurance 

The applicant's provisions for continuing assurance of training and qualification of plant 

personnel should be acceptable if the applicant's submittal addresses periodic retesting of 

personnel as necessary to provide reasonable assurance that the personnel continue to 

understand, recognize the importance of, and have the qualifications to perform their 

activities that are relied on for safety.

NUREG-1718 15.4-6
Management Measures

15.4.5 REVIEW PROCEDURES 

15.4.5.1 Acceptance Review 

The primary reviewer should perform an acceptance review to determine if the application 

adequately addresses the items in Section 15.4.3, "Areas of Review," for either the construction 

approval review or the review for a license to possess and use SNM. 

Guidance specific to the construction approval review and the review for a license to possess 

and use SNM is provided below. 

A. Construction Approval 

Specifically, the safety assessment of the design basis should address Section 15.4.3 

consistent with the level of design. Where information is under development or not yet 

available, the applicant may include a commitment to provide the material with the license 

application in lieu of the actual material. The primary reviewer should also verify that the 

applicant has committed to meeting or exceeding the acceptance criteria of Section 15.4.4. 

B. License To Possess and Use SNM 

Specifically, the license application should address Section 15.4.3 in full. The applicant is 

expected to have developed a program for the training and qualification of plant personnel 

prior to facility licensing for operations. 

If the primary reviewer verifies that the training and qualification of plant personnel is 

adequately addressed, the primary reviewer should accept the application for the safety 

evaluation in Section 15.4.5.2. If the primary reviewer identifies significant deficiencies in the 

material provided, the primary reviewer should request that the applicant submit additional 

information prior to the start of the safety evaluation. 

15.4.5.2 Safety Evaluation 

After determining that the application is acceptable for review in accordance with 

Section 15.4.5.1(A) (construction approval) or 15.4.5.1(B) (license to possess and use SNM), 

the primary reviewer should perform a safety evaluation against the acceptance criteria 

described in Section 15.4.4. On the basis of its review, the staff may request that the applicant 

provide additional information or modify the application to meet those acceptance criteria. 

Guidance specific to the construction approval review and the review for a license to possess 

and use SNM is provided below. 

A. Construction Approval

15.4-7 NUREG-1718
Management Measures

The primary reviewer should verify that the applicant's commitments and goals as they 

relate to the training and qualification of plant personnel are adequate to meet or exceed the 

acceptance criteria in Section 15.4.3. 

B. License To Possess and Use SNM 

The primary reviewer should focus the review on any new or changed material covering the 

training and qualification of plant personnel that the applicant updated with the license 

application. The primary reviewer should also confirm that the material remains consistent 

with the material provided in the license application in support of other chapters of this SRP. 

The primary reviewer should recognize that the training objectives and methods and the 

required qualification of plant personnel may be graded to correspond to the hazard 

potential of the facility, the items relied on for safety (I ROFS), and the complexity of the 

training needed. The review should evaluate the adequacy of training and qualification on 

the basis of how well it fulfills the objectives for the training as identified by the applicant, 

especially when human factors are relied on for safety. The primary reviewer should 

determine whether the applicant has adequately planned for the training and qualification of 

plant personnel to be accomplished and whether necessary policies, procedures, and 

instructions will be in place and appropriate training and qualification will be accomplished 

before these personnel begin activities relied on for safety. Some of the information may be 

referenced to other sections of the application, or incorporated by reference, provided that 

these references are clear and specific. 

The secondary reviewer should confirm that the applicant's commitments regarding the 

training and qualification of plant personnel are consistent with other sections of the 

applicant's submittal. 

The supporting reviewer (Fuel Cycle Facility Inspector) should become familiar with the 

applicant's commitments for the training and qualification of plant personnel and determine 

whether ongoing activities are in agreement with them. 

The review should result in a determination that there is reasonable assurance that the 

applicant's training and qualification of plant personnel will ensure that only properly trained 

and qualified personnel will perform activities relied on for safety. 

15.4.6 EVALUATION FINDINGS 

The primary reviewer should document the safety evaluation by preparing material suitable for 

inclusion in the Safety Evaluation Report (SER). The primary reviewer should describe the 

review, explain the basis for the findings, and state the conclusions. 

The staff could document a safety evaluation for the application for the construction approval 

review as follows:

NUREG-1718 15.4-8
Management Measures

The staff reviewed the application for [insert facility name] according to Section 15.4 of 

NUREG-1718. [Here the primary reviewer provides a summary statement of what was 

evaluated (including the applicant's commitments) and why the reviewer finds the 

applicant's submittal acceptable.] Based on its review of the application, the NRC staff 

concludes that the applicant adequately described its training and qualification of plant 

personnel (or made commitments to meet the acceptance criteria of Section 15.4.4 of 

NUREG-1 718) and that the applicant's training and qualification of plant personnel will, 

based on commitments, meet the requirements of 10 CFR Part 70 and provide reasonable 

assurance of protection of public health and safety and of the environment. 

The staff could document a safety evaluation for the license application using a similar 

paragraph as that used for the construction approval, but encompassing any new or updated 

material (and possible fulfilled commitments) when compared with the safety evaluation for the 

construction approval review. 

15.4.7 REFERENCES 

American Society of Mechanical Engineers (ASME). ASME-NQA-1-1994, "Quality Assurance 

Requirements for Nuclear Facility Applications." ASME: New York, New York. 1994. 

Code of Federal Regulations, Title 10, Energy, Part 70, "Domestic Licensing of Special Nuclear 

Material." 

Title 10, Energy, Part 19, "Notices, Instructions and Reports to Workers: Inspection 

and Investigations." 

Nuclear Regulatory Commission (U.S.), Washington, D.C. "Domestic Licensing of Special 

Nuclear Material, (10 CFR Part 70)." Federal Register: Vol. 64, No. 146. pp. 41338-41357. 

July 30, 1999. 

Nuclear Regulatory Commission (U.S.) (NRC). NUREG-1220, Rev. 1, "Training Review Criteria 

and Procedures." NRC: Washington, D.C. January 1993.

15.4-9 NUREG-1718
15.0 MANAGEMENT MEASURES 

15.5 PLANT PROCEDURES 

15.5.1 PURPOSE OF REVIEW 

The purpose of this review is to establish that there is reasonable assurance that the applicant 

is capable of and committed to providing management control of facility operations identified as 

items relied on for safety (IROFS) through the development, review, approval, control, and 

implementation of written plant procedures1 that will protect the workers, the public, and the 

environment during testing, startup, and operation of the facility. 

15.5.2 RESPONSIBILITY FOR REVIEW 

Primary: Project Manager 

Secondary: Primary staff reviewers of other management measures 

Human Factors Engineer 

SupDorting: Fuel Cycle Facility Inspector 

15.5.3 AREAS OF REVIEW 

The staff's review of the license application should address the process the applicant has 

developed for the production, use, and management control of written plant procedures. This 

should include the basic elements of identification, development, verification, initial review, 

comment resolution, approval, validation, issuance, change control, and periodic review. There 

should be two general types of plant procedures: 

A. Plant procedures used to directly control process operations, commonly called "operating 

procedures." These are procedures for workstation operators and they should include 

directions for normal operations as well as off-normal incidents caused by human error or 

equipment failure. Procedures of this type should include required actions to ensure 

nuclear criticality safety, chemical safety, fire protection, emergency planning, and 

environmental protection. 

B. Plant procedures used to perform activities that support the process operations, commonly 

referred to as "management control procedures." These are procedures used to manage 

the conduct of activities such as configuration management, radiation safety, 

maintenance, human-systems interface, quality assurance, design control, test control, 

1 This SRP section provides guidance for the review of information on plant procedures identified as 

IROFS. Section G5 of SRP Appendix G on quality assurance and Basic Requirement 5 of 

ASME-NQA-1-1994 provide review guidance for other procedures (for example, construction 

procedures) relied on for safety.

15.5-1 NUREG-1718
Management Measures

startup, plant personnel training and qualification, audits and assessments, incident 

investigations, recordkeeping, and reporting. 

15.5.4 ACCEPTANCE CRITERIA 

15.5.4.1 Regulatory Requirements 

The requirement for plant procedures is addressed in the following: 

Nuclear Regulatory Commission (U.S.), Washington, D.C. "Domestic Licensing of 

Special Nuclear Material (10 CFR Part 70)." Federal Register: Vol. 64, No. 145. pp. 

41338-41357. July 30,1999. 

Specific references are as follows: 

A. In § 70.4, "Definitions," the term "management measures" is defined. Procedures are 

included as a management measure. 

B. In § 70.22(a)(8), the application is required to include proposed procedures to protect 

health and minimize danger to life or property. 

C. In § 70.62(d), the applicant or licensee is required to establish management measures to 

provide continuing assurance of compliance with the performance requirements. 

D. In § 70.64(a)(1), the design of new facilities or new processes at existing facilities is 

required to be developed and implemented in accordance with management measures. 

E. In § 70.65(a), the application is required to include a description of the management 

measures. 

15.5.4.2 Regulatory Guidance 

None. 

15.5.4.3 Regulatory Acceptance Criteria 

For the construction approval, the applicant should commit to establish a process for the 

production, use, and management control of written plant procedures that meets or exceeds the 

acceptance criteria in Section 15.5.4. 

The reviewers should determine that the applicant's process for developing and implementing 

plant procedures is acceptable (for license approval) if the process satisfies the following:

NUREG-1718 15.5-2
Management Measures

A. Plant procedures should be written or planned for the conduct of all operations involving 

controls identified in the Integrated Safety Analysis (ISA) as activities relied on for safety 

and for all management control systems supporting those controls. 

B. Operating procedures contain the following elements: 

i. Purpose of the activity; 

ii. Regulations, polices, and guidelines governing the procedure; 

iii. Type of procedure; 

iv. Steps for each operating process phase; 

v. Initial startup; 

vi. Normal operations; 

vii. Temporary operations; 

viii. Emergency shutdown; 

ix. Emergency operations; 

x. Normal shutdown; 

xi. Startup following an emergency or extended downtime; 

xii. Hazards and safety considerations; 

xiii. Operating limits; 

xiv. Precautions necessary to prevent exposure of hazardous chemicals or licensed 

special nuclear material (SNM); 

xv. Measures to be taken if contact or exposure occurs; 

xvi. Safety controls associated with the process and their functions; and 

xvii. Specified time period or other limitations, if applicable, on the validity of the 

procedure. 

C. Management control procedures reflect the important elements of the functions described 

in the applicable chapters of this SRP. Management control procedures should exist for 

the following activities: 

i. Configuration management; 

ii. Radiation safety; 

iii. Maintenance; 

iv. Human-systems interface; 

v. Quality assurance; 

vi. Training and qualification; 

vii. Audits and assessments; 

viii. Incident investigations; 

ix. Records management; 

x. Nuclear criticality safety; 

xi. Fire safety; 

xii. Chemical process safety; 

xiii. Design control; 

xiv. Test control; 

xv. Startup; and 

xvi. Reporting requirements.

15.5-3 NUREG-1718
Management Measures

D. The applicant's method for identifying plant procedures includes using ISA results to 

identify needed procedures. Operating procedures should provide specific direction 

regarding administrative controls to ensure operational safety. 

E. The applicant's method for identifying, developing, approving, implementing, and 

controlling plant procedures should include, as a minimum: 

i. Operating limits and controls are specified in the procedure; 

ii. Procedures include required actions for off-normal conditions of operation as well as 

normal operations; 

iii. If needed, safety checkpoints are identified at appropriate steps in the procedure; 

iv. Procedures are validated through field tests or other methods where appropriate; 

v. Procedures are approved by management personnel responsible and accountable 

for the operation; 

vi. A mechanism is specified for revising and reissuing procedures in a controlled 

manner; 

vii. The quality assurance and configuration management programs at the facility provide 

reasonable assurance that current procedures are available at all work locations or, 

where not feasible for specific work locations, are readily accessible by all personnel 

and are used for all work. 

viii. The facility training program ensures that the required persons are trained in the use 

of the latest procedures. 

F. The application should include the following statement regarding adherence to plant 

procedures: "Activities involving special licensed nuclear material will be conducted in 

accordance with approved procedures." 

G. The applicant should discuss plant procedure categories used at the facility. An 

acceptable discussion should clearly state areas for which a plant procedure is required. 

The applicant should provide a list of the types of activities that are covered by the plant 

procedures. This list should include the topics of administrative plant procedures; system 

plant procedures that address startup, operation, and shutdown; abnormal operation/alarm 

response; maintenance activities that address system repair, calibration, inspection, and 

testing; and emergency procedures. Appendix H to this SRP provides an acceptable list 

of the items to be included under each topic. 

H. The applicant should indicate that following an incident-such as an accident, unexpected 

transient, significant operator error, or equipment malfunction--or following any

NUREG-1718 15.5-4
Management Measures

modification to a system, an appropriate review of all applicable written plant procedures 

will take place. 

The applicant should indicate how technical accuracy of plant procedures will be ensured 

as written. The discussion should identify who is responsible for verification. 

J. The applicant should indicate how documents will be distributed in accordance with current 

distribution lists. A process limiting the use of outdated plant procedures should be 

addressed. 

K. The applicant should describe how formal requirements governing temporary changes to 

plant procedures will be developed and implemented. 

L. Formal requirements for design control of items relied on for safety should be provided 

and should identify who is responsible for design inputs, processes, outputs, changes, 

interfaces, and records. 

M. A description of the test control program should be provided and should indicate that an 

effective program has been established for tests, including commissioning and 

preoperational tests. Acceptable plant procedures for test control should provide criteria 

for determining when a test is required or how and when testing activities are performed. 

i. Tests should be performed under conditions that simulate the most adverse design 

conditions, as determined by analysis. 

ii. Test results should be documented and evaluated, and their acceptability should be 

determined by a responsible individual or group. 

N. Plant procedures for maintenance involving safety controls should commit to the topics 

listed below for corrective and preventive maintenance, functional testing after 

maintenance, and surveillance/monitoring of maintenance activities: 

i. Premaintenance activity involving reviews of the work to be performed, including 

appropriate reviews of facility procedures for maintenance for accuracy and 

completeness. 

ii. Steps that require notification of all affected parties (operators and supervisors) prior 

to performing work and upon completion of maintenance work. 

iii. Control of work by comprehensive facility procedures to be followed by maintenance 

technicians. 

0. The applicant should commit to conducting periodic reviews of plant procedures to ensure 

their continued accuracy and usefulness. The applicant should establish the time frame 

for these reviews. At minimum, all procedures should be reviewed every 5 years. 

Emergency procedures should be reviewed every year initially and, if experience warrants,

15.5-5 NUREG-1718
Management Measures

subsequently reviewed at least every 2 years. All applicable procedures should be 

reviewed for major facility or process modifications. 

P. The applicant should describe the use and. control of procedures. 

Q. A preoperational testing (startup) program should be described. Information pertaining to 

how, and to what extent, the facility operating, emergency, and surveillance procedures 

will be user-tested during this test program should be provided. 

15.5.5 REVIEW PROCEDURES 

15.5.5.1 Acceptance Review 

The primary reviewer should perform an acceptance review to determine if the application 

adequately addresses the items in Section 15.5.3, "Areas of Review," for either the construction 

approval review or the review for a license to possess and use SNM. If the primary reviewer 

verifies that procedures are adequately addressed, the primary reviewer should accept the 

application for the safety evaluation in Section 15.5.5.2. If the primary reviewer identifies 

significant deficiencies in the material provided, the primary reviewer should request that the 

applicant submit additional information prior to the start of the safety evaluation. 

15.5.5.2 Safety Evaluation 

For the construction approval review, the reviewer should determine that the applicant has 

committed to management control of written plant procedures that will meet or exceed the 

acceptance criteria in Section 15.5.4. 

After determining that the application for license approval is acceptable for review in 

accordance with Section 15.5.5.1, the primary reviewer should perform a safety evaluation 

against the acceptance criteria described in Section 15.5.4. On the basis of its review, the staff 

may request that the applicant provide additional information or modify the application to meet 

those acceptance criteria. 

The safety evaluation forms the basis for staff findings and supports the reviewers' conclusions 

that the applicant has committed to: 

A. Controls that are identified in the ISA for facility safety procedures (i.e., procedures that 

constitute administrative controls for safety). 

B. Appropriate independent verification and validation of procedures for IROFS before use. 

C. The independent review and approval, and, where appropriate, review and approval by an 

independent multidisciplinary safety review, and control by the configuration management 

function of any change to facility procedures.

NUREG-1718 15.5-6
Management Measures

D. Following approved procedures while processing licensed SNM. 

E. Having procedures for the notification of operations personnel before and after 

maintenance is performed on safety controls. 

Secondary staff reviewers should ensure that the applicant's facility procedures do not conflict 

with their primary review areas. 

The supporting staff reviewer (Fuel Cycle Facility Inspector) should become familiar with the 

applicant's written plant procedures and determine whether ongoing activities are in agreement 

with them. 

15.5.6 EVALUATION FINDINGS 

The primary reviewer should document the safety evaluation by preparing material suitable for 

inclusion in the Safety Evaluation Report (SER). The primary reviewer should describe the 

review, explain the basis for the findings, and state the conclusions. 

The staff could document the safety evaluation for the construction approval review by stating 

that the applicant has committed to establish a process for the production, use, and 

management control of written plant procedures that meets or exceeds the acceptance criteria 

in Section 15.5.4. 

The staff could document the safety evaluation for the review for the license to possess and 

use SNM as follows: 

The staff reviewed the license application for [insert facility name] according to 

Section 15.5 of NUREG-1 718. The applicant has described suitably detailed processes 

for the development, review, approval, control, and implementation of procedures. [Insert 

a summary statement of what was evaluated and why the reviewer finds the submittal 

acceptable.] Special attention has been paid to items relied on for safety, as well as to 

systems important to the health of workers and the public and to the protection of the 

environment during testing, startup, and operation of the facility. 

15.5.7 REFERENCES 

American Society of Mechanical Engineers (ASME). NQA-1-1994, "Quality Assurance 

Requirements for Nuclear Facility Applications." ASME: New York, New York. 1994. 

Code of Federal Regulations, Title 10, Energy, Part 70, "Domestic Licensing of Special Nuclear 

Material."

15.5-7 NUREG-1718
Management Measures

Nuclear Regulatory Commission (U.S.), Washington, D.C. "Domestic Licensing of Special 

Nuclear Material (10 CFR Part 70)." Federal Register: Vol. 64, No. 146. pp. 41338- 1357. 

July 30, 1999. 

"Guidance on Management Controls/Quality Assurance, Requirements for Operation, 

Chemical Safety, and Fire Protection for Fuel Cycle Facilities." Federal Register: Vol. 54, No. 

53. pp. 11590-11598. March 21, 1989. 

Nuclear Regulatory Commission (U.S.) (NRC). Regulatory Guide 1.33, Rev. 2, "Quality 

Assurance Program Requirements (Operation)." NRC: Washington, D.C. February 1978.

NUREG-1 718 15.5-8
15.0 MANAGEMENT MEASURES 

15.6 AUDITS AND ASSESSMENTS 

15.6.1 PURPOSE OF REVIEW 

The purpose of this review is to establish that the applicant developed and adequately 

described a system of audits and assessments that provides reasonable assurance that the 

items relied on for safety (IROFS) will be available and reliable to perform their safety function 

when needed, as required by 10 CFR Part 70. 

15.6.2 RESPONSIBILITY FOR REVIEW 

Primary: Quality Assurance (QA) Engineer/Specialist 

Secondary: Project Manager 

SuDDortina: Fuel Cycle Facility Inspector 

15.6.3 AREAS OF REVIEW 

The applicant should submit a description of the system of audits and assessments for the 

construction approval review and should submit updated information with the application for a 

license to possess and use special nuclear material (SNM). The applicant's system of audits 

and assessments should consist of two distinct levels of activities: 

A. An independent internal or external audit is a QA organization activity to evaluate the scope, 

status, adequacy, programmatic compliance, and implementation effectiveness of QA and 

other management measures that provide reasonable assurance of continued availability 

and reliability of IROFS. 

B. An internal or external assessment is a management activity to evaluate the scope, status, 

adequacy, programmatic compliance, and implementation effectiveness of QA and other 

management measures that provide reasonable assurance of continued availability and 

reliability of IROFS. 

The following areas should be reviewed (construction approval): 

A. Audits and assessments-general; 

B. Audits; 

C. Assessments; and 

D. The applicant's provisions for continuing assurance.

15.6-1 NUREG-1718
Management Measures

15.6.4 ACCEPTANCE CRITERIA 

15.6.4.1 Regulatory Requirements 

The requirement for audits and assessments is addressed in the following: 

Nuclear Regulatory Commission (U.S.), Washington, DC. "Domestic Licensing of 

Special Nuclear Material (10 CFR Part 70)." Federal Register: Vol. 64, No. 146. 

pp. 41338-41357. July 30, 1999. 

Specific references are as follows: 

A. In § 70.4, "Definitions," the term "management measures" is defined. Audits and 

assessments are included as a management measure. 

B. In § 70.62(d), the applicant or licensee is required to establish management measures to 

provide continuing assurance of compliance with the performance requirements. 

C. In § 70.64(a)(1), the design of new facilities or new processes at existing facilities is 

required to be developed and implemented in accordance with management measures. 

D. In § 70.65(a), each application is required to include a description of the management 

measures. 

15.6.4.2 Regulatory Guidance 

None. 

15.6.4.3 Regulatory Acceptance Criteria 

The NRC reviewers should find that the applicant's audits and assessments provide reasonable 

assurance that the regulatory acceptance criteria below are adequately addressed and 

satisfied. 

A. Audits and Assessments 

General: Audits and assessments should be acceptable if: 

i. Internal audits, external audits, and assessments are to be conducted with a graded 

approach based on the results of the integrated safety analysis (ISA; see SRP 

Chapter 5.0). The stated objective of the audits and assessments should be to 

objectively evaluate the effectiveness and proper implementation of QA and other 

management measures for IROFS and to address the technical adequacy of the items 

being audited/assessed.

NUREG-1718 15.6-2
Management Measures

ii. The applicant describes, commits to, and provides justification for a frequency and 

scope of audits and assessments of IROFS. A commitment to perform audits and 

assessments in all areas where the requirements for QA and other management 

measures are applicable should be provided. Audits and assessments will be regularly 

scheduled on the basis of the status and the safety significance of the items being 

audited/assessed and will be initiated early enough to ensure the implementation of 

effective QA and other management measures. 

iii. Policy directives are established for audits and assessments. Policy directives cover 

schedules, guidance for conducting the audits and assessments, assigned 

responsibilities, and procedures for recording the audit/assessment results and ensuring 

that identified deficiencies are corrected in a timely and effective manner for each 

activity audited/assessed. 

iv. The applicant identifies the position title, qualifications, and responsibilities of the 

manager responsible for the overall success of the audits and assessments. Other 

organizational responsibilities for audits and assessments should be identified in the 

application. 

v. The applicant describes the training and qualification requirements for audit and 

assessment personnel. (SRP Section 15.4 addresses training and qualification 

requirements in detail.) 

vi. The applicant describes the authority each audit and assessment team has to 

investigate any aspect of the audited/assessed items with access to all relevant 

information. 

vii. Performance indicators are established so that audit and assessment personnel can 

determine the degree to which IROFS are meeting performance requirements. 

viii. Audits and assessments are conducted according to written procedures/checklists. 

(SRP Section 15.5 provides procedure guidance.) 

ix. Audits and assessments include detailed walkdowns of plant areas, including out-of-the

way and limited-access areas, with provisions for accurate, documented descriptions of 

any deficiencies. 

x. The applicant describes provisions for on-the-spot corrective actions with appropriate 

documentation. 

xi. Audit and assessment results are reviewed by management having responsibility in the 

area audited/assessed. 

xii. Audit and assessment findings and recommendations are documented and distributed 

to appropriate management for review and response. As described in SRP

15.6-3 NUREG-1718
Management Measures

Section 15.1, a corrective action program is administered to ensure timely and- effective 

corrective action. 

xiii. Audit and assessment deficiency data are analyzed and trended and resultant reports, 

which indicate quality trends and the effectiveness of management measures, are 

given to appropriate management for review, response, corrective action, and follow

up. 

B. Audits 

Audits should be acceptable if, in addition to addressing the acceptance criteria in 

Section 15.6.4.3.1 above: 

i. Audit personnel have no direct responsibility for the items they audit. 

ii. Audits are led by appropriately qualified and certified audit personnel from the QA 

organization. 

iii. Audit team membership may include personnel (not necessarily from the QA 

organization) who have technical expertise in the areas being audited. 

iv. Technical and programmatic audits are performed internally (that is, within the 

applicant's organization) and externally (that is, within the organization of suppliers, 

contractors, and subcontractors) and these audits provide a comprehensive 

independent verification and evaluation of procedures and activities for IROFS. 

v. Auditing organizations schedule and conduct appropriate follow-up to ensure timely and 

effective corrective action. 

vi. Audit reports are issued to appropriate management on a timely basis. 

vii. Reports on the status of corrective actions for audit findings are issued periodically to 

appropriate management. 

viii. Internal audits address compliance with selected operating limits during facility 

operation. 

C. Assessments 

Assessments should be acceptable if, in addition to addressing the acceptance criteria in 

Section 15.6.4.3.1 above, the application indicates that responsible management personnel 

(or that qualified, but not necessarily certified, personnel with no direct responsibility for the 

items being assessed who are designated by the responsible management) perform the 

assessments. 

D. Applicant's Provisions for Continuing Assurance

NUREG-1718 15.6-4
Management Measures

The applicant's provisions for continuing audits and assessments should be acceptable if 

changes to the program of audits and assessments due to reorganizations, revised 

activities, lessons learned, changes to applicable regulations, and other changes are 

reviewed and reflected in the program description. 

The applicant should also commit to update the system of audits and assessments to reflect 

any changes in the license application between the construction approval review and the review 

for a license to possess and use SNM. 

15.6.5 REVIEW PROCEDURES 

15.6.5.1 Acceptance Review 

The primary reviewer should perform an acceptance review to determine if the application 

adequately addresses the items in Section 15.6.3, "Areas of Review," for either the construction 

approval review or the review for a license to possess and use SNM. If the primary reviewer 

verifies that audits and assessments are adequately addressed, the primary reviewer should 

accept the application for the safety evaluation in Section 15.6.5.2. If the primary reviewer 

identifies significant deficiencies in the material provided, the primary reviewer should request 

that the applicant submit additional information prior to the start of the safety evaluation. 

15.6.5.2 Safety Evaluation 

After determining that the application is acceptable for review in accordance with 

Section 15.6.5.1, the primary reviewer should perform a safety evaluation against the 

acceptance criteria described in Section 15.6.4. On the basis of its review, the staff may 

request that the applicant provide additional information or modify the application to meet those 

acceptance criteria. 

The primary reviewer should determine whether the applicant has adequately planned for audits 

and assessments to be accomplished and whether necessary policies, personnel, procedures, 

and instructions will be in place to begin audits and assessments early, that is, during the ISA 

process and the design of IROFS. 

The secondary reviewer should confirm that the applicant's audit and assessment commitments 

are consistent with other sections of the submittal. 

The supporting reviewer should become familiar with the applicant's audit and assessment 

commitments and determine whether the ongoing audits and assessments of the applicant and 

the applicant's suppliers, contractors, and subcontractors are in agreement with them. 

The review should result in a determination that the audits and assessments will provide 

additional assurance that IROFS will perform satisfactorily in service and that activities relied on 

for safety will be performed satisfactorily.

15.6-5 NUREG-1718
Management Measures

When the applicant updates the system of audits and assessments, the primary reviewer 

should focus the review on any new or changed material. The primary reviewer should also 

confirm that the material presented remains consistent with the material provided in the license 

application for operations in support of other chapters of this SRP. 

15.6.6 EVALUATION FINDINGS 

The primary reviewer should document the safety evaluation by preparing material suitable for 

inclusion in the Safety Evaluation Report (SER). The primary reviewer should describe the 

review, explain the basis for the findings, and state the conclusions. 

The staff could document a safety evaluation of the application for the construction approval 

review as follows: 

The staff reviewed the license application for construction approval for [insert facility name] 

according to Section 15.6 of NUREG-1 718. [Here the primary reviewer provides a 

summary statement of what was evaluated and why the reviewer finds the submittal 

acceptable.] Based on its review of the application, the NRC staff concludes that the 

applicant has adequately described its system of audits and assessments, and the 

applicant's system of audits and assessments meets the requirements of 10 CFR Part 70 

and provides reasonable assurance of protection of public health and safety and of the 

environment. 

The staff could document a safety evaluation for the license to possess and use SNM as 

follows: 

[Here the primary reviewer provides a summary statement of what was evaluated and why 

the reviewer finds the submittal acceptable.] Based on its review of the application for a 

license to possess and use special nuclear material, focusing on new or updated material 

when compared with the safety evaluation for construction approval, the NRC staff 

concludes that the applicant has adequately described its updated system of audits and 

assessments, and the applicant's updated system of audits and assessments meets the 

requirements of 10 CFR Part 70 and thus provides reasonable assurance of protection of 

public health and safety and of the environment. 

15.6.7 REFERENCES 

Code of Federal Regulations, Title 10, Energy, Part 70, "Domestic Licensing of Special Nuclear 

Material." 

Nuclear Regulatory Commission (U.S.), Washington, D.C. "Domestic Licensing of Special 

Nuclear Material (10 CFR Part 70)." Federal Register: Vol. 64, No. 146. pp. 41338-41357. 

July 30, 1999.

NUREG-1718 15.6-6
15.0 MANAGEMENT MEASURES 

15.7 INCIDENT INVESTIGATIONS 

15.7.1 PURPOSE OF REVIEW 

The purpose of this review is to establish, with reasonable assurance, that the applicant will 

have a system in place for the systematic investigation 6f incidents 1 , assignment and 

acceptance of corrective actions, and follow-up to ensure completion of the actions. The review 

should confirm that incidents will be investigated and corrective action taken to prevent (or 

minimize) their recurrence or their leading to more serious consequences. Furthermore, the 

review should find that the results of incident investigations will be compared against the 

Integrated Safety Analysis (ISA) Summary (see SRP Chapter 5.0) to provide assurance that 

there is continued compliance with the performance requirements contained in 10 CFR Part 70. 

15.7.2 RESPONSIBILITY FOR REVIEW 

Primary: Project Manager 

Secondary: Quality Assurance (QA) Engineer/Specialist and ISA Reviewers 

Supporting: Fuel Cycle Facility Inspector 

15.7.3 AREAS OF REVIEW 

The staff's review of the license application should encompass the following areas: 

A. The description of the functions, qualifications, and responsibilities of the management 

person who would lead the investigation and those of team members2 , the scope of the 

investigation person or team's authority and responsibilities, the process for determining the 

need for an investigation team rather than an individual investigator, and assurance of 

cooperation of management. 

B. The investigation person's or team's ability to obtain all the information considered 

necessary and independence from responsibility for or to the functional area involved in the 

incident under investigation. 

C. The maintenance of documentation consistent with SRP Section 15.8, "Records 

Management." 

Incidents are unplanned events such as accidents, unexpected transients, equipment malfunctions, 

operator error, and unacceptable performance deficiencies. 

2 Individual members of the team may have responsibility for the functional area provided that they had 

no involvement in the incident being investigated. The team leader or individual investigator is 

independent of the functional area involved in the incident.

15.7-1 NUREG-1718
Management Measures

D. Guidance for the person or team conducting the investigation on how to apply a reasonable, 

systematic, structured approach to determine the root cause(s) of the incident and any 

generic implications. 

E. The system for comparing the results of the investigation against the ISA. 

F. The system for monitoring to ensure completion of any corrective measures specified, 

including revisions to the ISA. 

15.7.4 ACCEPTANCE CRITERIA 

15.7.4.1 Regulatory Requirements 

Nuclear Regulatory Commission (U.S.), Washington, D.C. "Domestic Licensing of 

Special Nuclear Material (10 CFR Part 70)." Federal Register: Vol. 64, No. 146. 

pp. 41338-41357. July 30, 1999. 

Specific references are as follows: 

A. In § 70.4, "Definitions," the term "management measures" is defined. Incident 

investigations are included as a management measure. 

B. In § 70.62(d), the applicant or licensee is required to establish management measures to 

provide continuing assurance of compliance with the performance requirements. 

C. In § 70.64(a)(1), the design of new facilities or the design of new processes at existing 

facilities is required to be developed and implemented in accordance with management 

measures. 

D. In § 70.65(a), the application is required to include a description of the management 

measures. 

15.7.4.2 Regulatory Guidance 

See the references listed in Section 15.7.7 for useful background information on specific 

aspects of incident management, such as corrective action and root cause analysis. 

15.7.4.3 Regulatory Acceptance Criteria 

For the construction approval review, the applicant should commit to establishing a system for 

the systematic investigation of incidents, assignment and acceptance of corrective actions, and 

follow-up to ensure completion of the actions that meets or exceeds the acceptance criteria in 

Section 15.7.4.

NUREG-1718 15.7-2
Management Measures

The NRC reviewers should find the license application for operations acceptable if the 

applicant's system of incident investigations provides reasonable assurance that the regulatory 

acceptance criteria below are adequately addressed and satisfied. Some of the information 

may be referenced to other sections of the SRP, or incorporated by reference, provided that 

these references are clear and specific. 

A. Acceptability should be based on commitments for the prompt investigation of incidents that 

include the following elements: 

i. The establishment of teams to investigate incidents that may occur during operation of 

the facility, to determine the root cause(s) of the incident and any generic implications, 

and to recommend corrective actions. 

ii. The monitoring and documenting of corrective actions (including effectiveness) through 

completion. 

iii. The maintenance of documentation so that "lessons learned" may be applied to future 

operations of the facility. Details of the incident sequence should be compared with 

incident sequences already considered in the ISA, and actions should be taken to 

ensure that the ISA includes the evaluation of the risk associated with incidents of the 

type actually experienced. 

B. Acceptability should be based on the adequacy of the applicant's commitments to establish 

and use a plan for the investigation of incidents. Acceptability should also be based on the 

following acceptance criteria: 

i. The licensee has described the overall plan and method for investigating incidents. The 

plan is separate from any required emergency plan. 

ii. The functions, responsibilities, and scope of authority of investigation teams are 

documented in the plan. 

iii. Qualified internal or external investigators are appointed to serve on investigation teams. 

Each team should include at least one process expert and one team member trained in 

root cause analysis. 

iv. The investigation process and investigation person or team members are independent 

of the line function(s) involved with the incident under investigation2 , and participants are 

assured of protection from retribution for participating in investigations. 

v. A reasonable, systematic, structured approach is used to determine the root cause(s) of 

incidents. The level of investigation should be based on a graded approach relative to 

the severity of the incident. 

vi. Auditable records and documentation related to incidents, investigations, and root cause 

analysis are maintained.

15.7-3 NUREG-1718
Management Measures

vii. For each incident, an incident report is prepared that includes a description of the 

incident, contributing factors, root cause analysis, findings, and recommendations. 

Relevant findings should be reviewed with all affected personnel, and the reports should 

be made available to the NRC on request. 

viii. Documented corrective actions are taken within a reasonable period to resolve findings 

from incident investigations. 

15.7.5 REVIEW PROCEDURES 

15.7.5.1 Acceptance Review 

The primary reviewer should perform an acceptance review to determine if the application 

adequately addresses the items in Section 15.7.3, "Areas of Review," for either the construction 

approval review or the review for a license to possess and use special nuclear material (SNM). 

If the primary reviewer verifies that incident investigations are adequately addressed, the 

primary reviewer should accept the application for the safety evaluation in Section 15.7.5.2. If 

the primary reviewer identifies significant deficiencies in the material provided, the primary 

reviewer should request that the applicant submit additional information prior to the start of the 

safety evaluation. 

15.7.5.2 Safety Evaluation 

For construction approval, the reviewer should determine that the applicant has committed to a 

system for the systematic investigation of incidents, assignment and acceptance of corrective 

actions, and follow-up to ensure completion of the actions that will meet or exceed the 

acceptance criteria in Section 15.7.4. 

For a license application for operations, after determining that the application is acceptable for 

review in accordance with Section 15.7.5.1, the primary reviewer should perform a safety 

evaluation against the acceptance criteria described in Section 15.7.4. On the basis of its 

review, the staff may request that the applicant provide additional information or modify the 

application to meet those acceptance criteria. 

The review should determine if the applicant has adequately planned for incident investigations 

to be conducted and resulting corrective actions to be appropriately implemented. 

The primary reviewer should confirm that the organizational structure for incident investigations 

is consistent with SRP Chapter 4.0, "Organization and Administration." 

The QA secondary reviewer should verify that methods used for determining root causes and 

any generic implications, procedures for tracking and implementing the corrective actions, and 

the process of applying the "lessons learned" to the other operations are appropriate for 

incident investigations.

NUREG-1718 15.7-4
Management Measures

The ISA reviewers should verify that the applicant ensures that the results of the investigation 

are compared against the ISA and that the necessary follow-up actions occur. 

The supporting reviewer(s) should become familiar with pertinent procedures and determine 

whether planned future and ongoing activities are consistent with them. 

15.7.6 EVALUATION FINDINGS 

The primary reviewer should document the safety evaluation by preparing material suitable for 

inclusion in the Safety Evaluation Report (SER). The primary reviewer should describe the 

review, explain the basis for the findings, and state the conclusions. 

The staff could document the safety evaluation for the construction approval review by stating 

that the applicant has committed to establishing a system for the systematic investigation of 

incidents, the assignment and acceptance of corrective actions, and follow-up to ensure 

completion of the actions that meets or exceeds the acceptance criteria in Section 15.7.4. 

The staff could document a safety evaluation for the review for the license to possess and use 

SNM as follows: 

The staff reviewed the license application for [insert facility name] according to 

Section 15.7 of NUREG-1 718. Based on its review, [Insert a summary statement of 

what was evaluated and why the reviewer finds the submittal acceptable.] The NRC 

staff concluded that the applicant has committed to and established an organization 

responsible for investigating incidents that occur during operation of the facility, 

determining the root cause(s) and any generic implications of each incident, and taking 

corrective actions for ensuring a safe facility and safe facility operations in accordance 

with the acceptance criteria of Section 15.7.4 of the SRP; committed to review the 

results of the investigation against the ISA; committed to monitoring and documenting 

corrective actions through completion; and committed to the maintenance of related 

documentation and the application of "lessons learned" to future operations of the 

facility. 

Accordingly, the staff concludes that the applicant's description of the incident investigation 

process complies with applicable NRC regulations and is adequate. 

15.7.7 REFERENCES 

Code of Federal Regulations, Title 10, Energy, Part 70, "Domestic Licensing of Special Nuclear 

Material." 

Department of Energy (U.S.) (DOE). DOE-STD-1010-92, "Guide to Good Practices for 

Incorporating Operating Experiences." DOE: Washington, D.C. July 1992.

15.7-5 NUREG-1718
Management Measures

DOE-NE-STD-1 004-92, "Root Cause Analysis Guidance Document." 

DOE: Washington, D.C. February 1992. 

Nuclear Regulatory Commission (U.S.), Washington, D.C. "Domestic Licensing of Special 

Nuclear Material (10 CFR Part 70)." Federal Register: Vol. 64, No. 146. pp. 41338-41357. 

July 30, 1999. 

Nuclear Regulatory Commission (U.S.) (NRC). Information Notice 96-28, "Suggested 

Guidance Relating to Development and Implementation of Corrective Action." NRC: 

Washington, D.C. May 1996. 

NUREG/CR-4616, "Root Causes of Component Failures Program: Methods and 

Applications." NRC: Washington, D.C. December 1986. 

NUREG/CR-5665, "A Systematic Approach to Repetitive Failures." NRC: Washington, 

D.C. February 1991.

NUREG-1718 15.7-6
15.0 MANAGEMENT MEASURES 

15.8 RECORDS MANAGEMENT 

15.8.1 PURPOSE OF REVIEW 

The purpose of this review is to verify thatthe applicant has established a facility records 

management system that complies with NRC requirements. 

15.8.2 RESPONSIBILITY FOR REVIEW 

Primary: Project Manager 

Secondary: None 

SupDorting: Primary reviewers of SRP Sections 15.1, "Quality Assurance," and 

15.2, "Configuration Management" 

15.8.3 AREAS OF REVIEW 

The applicant should submit a description of the facility records management system for the 

construction approval review and should submit updated information with the application for a 

license to possess and use special nuclear material (SNM). 

Areas related to the handling and storing of records generated or needed in the design, 

construction, and operation phases of the facility, including the following, should be reviewed for 

construction approval. 

A. The process whereby records-such as training records, dosimetry records, effluent 

records, and records regarding the facility structures, systems, or components that are 

items relied on for safety-are specified, created, verified, categorized, indexed, 

inventoried, protected, stored, maintained, distributed, and deleted or preserved. The 

process may be linked with or be a part of the facility quality assurance and configuration 

management systems. 

B. The handling and control of various kinds of records and the methods of recording media 

that comprise the records, including contaminated and classified records. 

C. The physical characteristics of the record storage facilities with respect to the preservation 

and protection of the records for their designated lifetimes.

15.8-1 NUREG-1718
Management Measures

15.8.4 ACCEPTANCE CRITERIA 

15.8.4.1 Regulatory Requirements 

The requirements for records management are addressed in the following: 

A. Code of Federal Regulations, Title 10, Energy, Part 19, "Notices, Instructions and Reports 

to Workers: Inspection and Investigations." 

B. , Title 10, Energy, Part 20, "Standards for Protection Against Radiation." 

C. , Title 10, Energy, Part 21, "Reporting of Defects and Noncompliance." 

D. , Title 10, Energy, Part 25, "Access Authorization for Licensee Personnel." 

E. Nuclear Regulatory Commission (U.S.), Washington, D.C. "Domestic Licensing of Special 

Nuclear Material (10 CFR Part 70)." Federal Register: Vol. 64, No. 146. pp. 41338-41357. 

July 30, 1999. 

15.8.4.2 Regulatory Guidance' 

Regulatory guidance applicable to the area of records management is as follows: 

U.S. Nuclear Regulatory Commission (U.S.) (NRC). NUREG-1460, Rev. 1, "Guide to 

NRC Reporting and Recordkeeping Requirements." NRC: Washington, D.C. 

July 1994. 

15.8.4.3 Regulatory Acceptance Criteria 

The reviewer should find the applicant's records management system acceptable if it satisfies 

the following criteria: 

A. Records are specified, prepared, verified, characterized, and maintained. 

B. Records are legible, identifiable, and retrievable for their designated lifetimes. 

C. Records are protected against tampering, theft, loss, unauthorized access, damage, or 

deterioration for the time they are in storage. 

1 Additional guidance for records is given in SRP Appendix G on quality assurance (Section G17) and in 

ASME-NQA-1-1994 (Basic Requirement 17 and Supplement 17S-1) as referenced in SRP Section 

15.1, "Quality Assurance." 

NUREG-1718 15.8-2
Management Measures

D. Procedures are established and documented specifying the requirements and 

responsibilities for record selection, verification, protection, transmittal, distribution, 

retention, maintenance, and disposition. 

E. The organization and procedures are in place to promptly detect and correct any 

deficiencies in the records management system or its implementation. 

Examples of the types of records that could be included in the system and that contribute to 

providing reasonable assurance of protection of worker and public health and safety and of the 

environment are listed in Appendix I to this SRP. Records should be categorized by relative 

safety importance to identify record protection and storage needs and to designate the retention 

period for individual kinds of records. The procedures should assign responsibilities for records 

management; specify the authority needed for records retention or disposal; specify which 

records must have controlled access and provide the controls needed; provide for the 

protection of records from loss, damage, tampering, or theft during an emergency; and specify 

procedures for ensuring that the records management system remains effective. 

For records consisting of computer codes/computerized data relied on for safety, the 

application should establish and describe procedure(s) for maintaining readability and usability 

of older codes/data as computing technology changes. 

The applicant should also commit to update the facility records management system to reflect 

any changes in the license application between the construction approval review and the review 

for a license to possess and use SNM. 

15.8.5 REVIEW PROCEDURES 

15.8.5.1 Acceptance Review 

The primary reviewer should perform an acceptance review to determine if the application 

adequately addresses the items in Section 15.8.3, "Areas of Review," for either the construction 

approval review or the review for a license to possess and use SNM. If the primary reviewer 

verifies that the records management system is adequately addressed, the primary reviewer 

should accept the application for the safety evaluation in Section 15.8.5.2. If the primary 

reviewer identifies significant deficiencies in the material provided, the primary reviewer should 

request that the applicant submit additional information prior to the start of the safety 

evaluation. 

15.8.5.2 Safety Evaluation 

After determining that the application is acceptable for review in accordance with 

Section 15.8.5.1, the primary reviewer should perform a safety evaluation against the 

acceptance criteria described in Section 15.8.4. On the basis of its review, the staff may 

request that the applicant provide additional information or modify the application to meet those

15.8-3 NUREG-1718
Management Measures

acceptance criteria. The primary reviewer should coordinate this review with the primary 

reviewers of SRP Sections 15.1, "Quality Assurance," and 15.2, "Configuration Management." 

When the applicant updates the facility records management system in the application for a 

license to possess and use SNM, the primary reviewer should focus the review on any new or 

changed material. The primary reviewer should also confirm that the material presented 

remains consistent with the material provided in the license application for operations in support 

of other chapters of this SRP. 

15.8.6 EVALUATION FINDINGS 

The primary reviewer should document the safety evaluation by preparing material suitable for 

inclusion in the Safety Evaluation Report (SER). The primary reviewer should describe the 

review, explain the basis for the findings, and state the conclusions. 

The staff could document a safety evaluation for the construction approval review as follows: 

The staff reviewed the application for a construction approval for [insert facility name] 

according to Section 15.8 of NUREG-1 718. The staff reviewed the applicant's records 

management system [insert a summary statement of what was evaluated] and 

concluded that there is reasonable assurance that the system will (1) be effective in 

collecting, verifying, protecting, and storing information about the health and safety 

aspects of the facility and its operations and will be able to retrieve the information in 

readable form for the designated lifetimes of the records; (2) provide record storage 

facilities capable of protecting and preserving records that are stored there during the 

mandated periods, including protecting the stored records against loss, theft, tampering, 

or damage during and after emergencies; and (3) ensure that any deficiencies in the 

records management system or its implementation will be detected and corrected in a 

timely manner. The staff concludes that the applicant's facility records management 

system meets the requirements of 10 CFR Part 70 and is acceptable. 

The staff could document a safety evaluation for review for a license to possess and use SNM 

using a paragraph similar to the one used for the construction approval review, but 

encompassing the new or changed material when compared with the safety evaluation for the 

construction approval. 

15.8.7 REFERENCES 

American Society of Mechanical Engineers (ASME). NQA-1-1994, "Quality Assurance 

Requirements for Nuclear Facility Applications." ASME: New York, New York. 1994. 

Code of Federal Regulations, Title 10, Energy, Part 19, "Notices, Instructions and Reports to 

Workers: Inspection and Investigations."

NUREG-1718 15.8-4
Management Measures

Title 10, Energy, Part 20, "Standards for Protection Against Radiation." 

Title 10, Energy, Part 21, "Reporting of Defects and Noncompliance." 

Title 10, Energy, Part 25, "Access Authorization for Licensee Personnel." 

Title 10, Energy, Part 70, "Domestic Licensing of Special Nuclear Material." 

Nuclear Regulatory Commission (U.S.), Washington, D.C. "Domestic Licensing of Special 

Nuclear Material, (10 CFR Part 70)." Federal Register: Vol. 64, No. 146. pp. 41338-31357. 

July 30, 1999. 

Nuclear Regulatory Commission (U.S.) (NRC). NUREG-1460, Rev. 1, "Guide to NRC 

Reporting and Recordkeeping Requirements." NRC: Washington, D.C. July 1994.

15.8--5 NUREG-1718
APPENDIX A 

EXAMPLE PROCEDURE FOR RISK EVALUATION 

NRC requires that an applicant evaluate its compliance with the performance requirements 

defined in 10 CFR 70.61, which stipulate that certain consequences must be sufficiently 

unlikely. In addition, 10 CFR 70.62(c) requires that the applicant perform an Integrated Safety 

Analysis (ISA) to identify all potential accident sequences and to assess their consequences. 

These two requirements are related. The consequences result from accident sequences 

identified in the ISA. Thus, to show that the likelihood of occurrence of the consequences is 

sufficiently low, the applicant must show that for each of the accident sequences identified in 

the ISA, the resulting consequences are sufficiently unlikely. 

As defined in 10 CFR 70.61, the required likelihood is graded according to the severity of the 

consequences of the accident. Accidents in the intermediate consequence category of 

§ 70.61(c) must be "unlikely," while those in the high consequence category of § 70.61 (b) must 

be "highly unlikely." The procedure described in this appendix is one way by which the 

applicant may use the ISA results to demonstrate that the requirements of 10 CFR 70.61 have 

been met. If the applicant evaluates accidents using a different method, the method should 

produce similar results in terms of how accidents are categorized. This method should be 

regarded as a screening method, not as a definitive method of proving the adequacy or 

inadequacy of the controls for any particular accident. The method requires the applicant to 

identify and evaluate the characteristics of controls used to limit accident sequences in a 

consistent manner. This will permit identification of accident sequences with defects in the 

combination of controls used. Such controls can then be further evaluated or improved to 

establish adequacy. The procedure also ensures the consistent evaluation of similar controls 

by different ISA teams. Sequences or controls that have risk significance and are evaluated as 

marginally acceptable are good candidates for more detailed evaluation by the applicant and 

the reviewer. 

The tabular accident summary resulting from the ISA should identify, for each sequence, what 

safety controls must fail to exceed the consequence defined in the performance requirements 

of 10 CFR 70.61. Chapter 5.0 specifies acceptance criteria for these safety controls, such that 

the performance requirements of § 70.61 are met. These criteria require that safety controls be 

sufficiently unlikely to fail. However, the criteria of Chapter 5.0 do not provide for a method for 

assessing likelihood. This appendix describes an acceptable procedure for this required 

assessment of likelihood. 

Al. DETERMINING COMPLIANCE WITH GRADED PROTECTION REQUIREMENTS 

The regulations in 10 CFR Part 70.61 describe requirements for a graded system of protection 

sufficient to bound the risk of identified accidents by making accidents with a higher potential for 

consequences have a proportionately lower likelihood of occurrence. The performance 

requirements designate two categories of consequences into which an accident may fall. The 

first category is referred to in § 70.61 as "high consequences," the second as "intermediate 

consequences." Implicitly there is a third category, namely, those accidents that produce 

consequences less than "intermediate." These will be referred to as "low consequence" 

accidents. Since the primary purpose of process hazard analysis is to identify all accidents

A-1 NUREG-1718
Appendix A

having consequences that do not comply with the performance requirements, it will, in some 

cases, be necessary to identify accidents that produce radioactive or chemical exposures, then 

subsequently determine that some of these exceed the threshold values of the regulation. For 

this reason, the list of accidents resulting from such analysis will include such low consequence 

accidents to show that they have been considered. Otherwise the analysis will not have 

demonstrated its completeness. 

The limits defining the three accident consequence categories are given in Table A-I. The 

categories are numbered in ascending order of the magnitude of their consequences. The 

usefulness of this numbering will be evident later. The symbols CHEM3, CHEM2, and CHEM1 

refer to quantitative standards selected by the applicant in accordance with 

10 CFR 70.61 (b)(4)(ii) or 70.61 (c)(4)(ii) - e.g., AEGL or ERPG, as appropriate. 

Consequence Category 3-High Consequences 

An accident resulting in any consequence specified in § 70.61(b); that is: an acute worker 

exposure of 1 Sv (100 rem)1 or greater TEDE2, or a chemical exposure that could endanger the 

life of a worker (as defined by the applicant); or acute exposure of a member of the public 

outside the controlled area to a radiation dose (D) of 0.25 Sv (25 rem) or greater TEDE, a 30 

mg soluble uranium intake, or a chemical exposure that could lead to irreversible or other 

serious long-lasting health effects, as defined by the applicant (represented herein as CHEM3). 

Consequence Category 2-Intermediate Consequences 

An accident resulting in any consequence specified in § 70.61 (c). That is, acute exposure of a 

worker to a radiation dose of 0.25 Sv (25 rem) or greater but less than 1 Sv (100 rem) TEDE, or 

chemical exposure that could lead to irreversible or other serious long-lasting health effects, as 

defined by the applicant (represented herein as CHEM2); or acute exposure of a member of the 

public outside the controlled area to a radiation dose of 0.05 (5 rem) or greater but less than 

0.25 Sv (25 rem) TEDE, or a chemical exposure that could cause mild transient health effects, 

as defined by the applicant (represented herein as CHEMI1); or prompt release of radiation 

outside the restricted area that would, if averaged over a 24-hour period, exceed 5,000 times 

the values specified in Table 2 of Appendix B to 10 CFR Part 20. 

Consequence Category 1-Low Consequences 

Any accident with potential adverse radiological or chemical consequences but at exposures 

less than consequence Categories 3 and 2 above. 

1 A nuclear criticality would normally be considered a high consequence event because of the potential 

for producing a high radiation dose to a worker. 

2 TEDE is Total Effective Dose Equivalent (see 10 CFR Part 20), represented by 'D'.

NUREG-1718 A-2
Appendix A

TABLE A-i: Consequence Severity Categories Based on 10 CFR 70.61

Workers Offsite Public Environment 

Consequence D2 > 1 Sv (100 rem) D 2 0.25 Sv (25 rem) 

Category 3: > CHEM3 30 mg sol U intake 

High Ž CHEM2 

Consequence 0.25 Sv _ D < 1 Sv 0.05 SvD < 0.25 Sv radioactive release 

Category 2: > CHEM2 _CHEM1 > 5,000 x 

Intermediate but but Table 2 App. B 

< CHEM3 < CHEM2 10 CFR 20 

Consequence Accidents of lesser Accidents of lesser Radioactive releases 

Category 1: radiological and radiological and producing effects 

Low chemical exposures chemical exposures less than those 

to workers than to the public than specified above in 

those above in this those above in this this column 

column column 

Corresponding to the two consequence categories of the rule (Categories 2 and 3 above), 

§ 70.61 requires corresponding levels of graded protection, that is, engineered or administrative 

controls (or a combination thereof), sufficient to ensure that the likelihood of these adverse 

events is correspondingly low. The two categories of likelihood thus prescribed are: 

Likelihood Category 1: Consequence Category 3 accidents must be "highly unlikely." 

Likelihood Category 2: Consequence Category 2 accidents must be "unlikely." 

Implicitly there is a third category into which an accident could fall, that is, it could fail to be 

"unlikely." This category will be referred to in this document as: 

Likelihood Category 3: "Not unlikely." Although this likelihood category includes 

unintended events that might actually be expected to happen, others might be less 

frequent. For this reason, the term "likely" was not used for these events. 

Per 10 CFR 70.61, the applicant must use the ISA is to document its compliance with the 

performance requirements. This evaluation should be done using a tabular summary of 

identified accident sequences. One acceptable way of doing so is for the applicant to assign 

two category numbers to each accident sequence, one based on its consequences and one for 

likelihood. The product of these two category numbers is then used as a risk index. Listing this 

calculated risk index in the tabular summary provides a simple method for showing that the 

graded protection requirements have been met for each accident sequence. A risk index value 

less than or equal to "4" means the sequence is acceptable. If the applicant provides this risk

A-3 NUREG-1718
Appendix A

index in one column of the tabular summary, the reviewer can quickly scan this column to 

confirm that each accident conforms to the safety performance requirements of 10 CFR 70.61. 

This system is equivalent to assigning each accident to a cell in a three by three matrix. This 

conceptual matrix is shown in Table A-2. The values in the risk matrix cells are the risk index 

numbers. 

TABLE A-2: Risk Matrix 

I[_Likelihood Category 1: Likelihood Category 2: Likelihood Category 3: 

Highly Unlikely Unlikely Not Unlikely.I 

Consequence Category 3 acceptable 6 unacceptable 9 unacceptable 

3: 

High _. _ _ _:_ _ _ ".__"_ _ _ _ 

Consequence Category 2 acceptable 4 acceptable 6 unacceptable 

2: 

Intermediate 

Consequence Category 1 acceptable 2 acceptable 3 acceptable 

1: 

Low 

To demonstrate compliance with the system described above, the applicant needs to assign 

consequence categories to each identified accident to determine which likelihood requirement 

applies. Those accident sequences identified as high or intermediate consequences must then 

be assigned to a likelihood category. To be acceptable, these assigned consequences and 

likelihoods must have a valid basis, and the applicant must demonstrate this basis in the 

documentation submitted in the application. The following sections describe an acceptable 

method for making these assignments. 

A2. CONSEQUENCE CATEGORY ASSIGNMENT 

The assignment of consequence categories is based on estimated consequences of prototype 

accidents. Criteria for the presentation of these estimates by the applicant is described in SRP 

Section 5.4.3.2(B)(iv). Although consequences of accidents can be determined by actual 

calculations, it is not necessary that such a calculation be performed for each individual 

accident sequence listed. Accident consequences may be estimated by comparison to similar 

events for which reasonably bounding conservative calculations have been made. The 

applicant should document the bases for bounding calculations of the consequence assignment 

in the submittal. NUREG/CR-6410, "Nuclear Fuel Cycle Facility Accident Analysis Handbook," 

describes valid methods and data to be used by the applicant and may be used for confirmatory 

evaluations by the reviewer.

NUREG-1718 A-4
Appendix A

A3. LIKELIHOOD CATEGORY ASSIGNMENT 

An assignment of an accident sequence to a likelihood category is acceptable if it is based on 

the record of failures at the facility or other methods that have objective validity. Failure data 

from other facilities may also be used, but care should be taken to ensure its applicability. 

Because the sequences leading to accidents often involve multiple failures, a combination of 

failure frequency and probability values determines the likelihood of the whole accident 

sequence. These values include the frequencies of initiating events and failure likelihoods of 

safety controls. As described below, the applicant may estimate an approximate likelihood 

category for an accident sequence by considering all the events involved. This method uses 

the number, type, independence, and observed failure history of safety controls. However, 

correctly evaluating the appropriate likelihood of accidents using such a qualitative approach 

depends on the informed judgement of the analyst. Safety controls, even those of the same 

type, have a wide range of reliability. The ultimate criterion for acceptability is that the 

frequencies of initiating events and the likelihood of failure of safety controls involved is 

sufficiently low so that the entire accident sequence is "highly unlikely" or "unlikely" as required 

by 10 CFR 70.61. The virtue of the approach is that it requires explicit consideration of some of 

the underlying events and factors that affect the likelihood of the accident. Another virtue is that 

the more explicit the criteria for assignment are, the more consistent are the results. 

Underlying any evaluation of an accident sequence as "unlikely" or "highly unlikely" is an 

implied assessment of its "likelihood" or frequency of occurrence. The structured procedure 

described below will indicate which likelihood category may be appropriate for an event. To 

maintain internal consistency in evaluating different control systems and accidents, it was 

necessary to derive this structured procedure based on the underlying frequencies of events. 

The following numerical guidelines are used for the purposes of this example. The underlying 

frequencies are based on definite assumptions about the numbers of intermediate and high 

consequence events, as discussed in SRP Section 5.4.3.2. 

Likelihood Category 1: Highly unlikely, a frequency of less than 10"s per year per accident. 

Likelihood Category 2: Unlikely3, a frequency of less than 4x10-4 per year per accident (but 

more frequent than 101). 

Likelihood Category 3: Not unlikely, more frequent than 4x1 04 per year per accident. 

In assigning specific numerical values to these likelihood categories, we are making definitive 

assumptions about the number of accident sequences. The Commission's strategic goals are 

stated in terms of total industry risk, so that the per accident probabilities must be expressed as 

the cumulative likelihood divided by the total number of accident sequences. For the purposes 

3 A distinction must be drawn between the concept of "unlikely" in regard to intermediate consequence 

events and "unlikely" in regard to the double contingency principle. The above definition of unlikely 

does not apply to a nuclear criticality (which should be regarded as a high consequence event in 

unshielded facilities in most instances). In meeting double contingency, "unlikely" typically means 

•_10.2.

A-5 NUREG-1718
Appendix A

of this example, it will be assumed throughout the remainder of this appendix that there are 100 

intermediate consequence accidents and 1,000 high consequence accidents throughout the 

industry (these numbers provide results that are consistent with those stated in SRP 

Section 5.4.3.2 (p. 5.0-28)). 

With this assumption, each individual accident sequence in this likelihood category should have 

a frequency no greater than 1 05 per year (i.e., one accident of this type every 100,000 years). 

This number can be multiplied by the total number of accident sequences to give the cumulative 

likelihood of all accident sequences in a given category at the facility, in units of yr'. 

In assessing the adequacy of safety controls, individual accident frequencies greater than 10"5 

per year may not be assigned a likelihood Category 1, that is, "highly unlikely." The NRC has a 

strategic safety performance measure of no inadvertent nuclear criticalities. For this reason, 

the acceptability of any given frequency depends on the total number of accidents that may be 

identified. Since the total number and consequences of all potential accidents at a facility is not 

accurately known until its ISA is completed, it is difficult to establish a definitive acceptable 

frequency. Individual accidents may need to be limited to lower frequencies to meet the 

performance requirements. On the other hand, the fact that a particular accident sequence is 

below this value does not automatically mean that it is clearly acceptable. The frequencies 

should be used as a guideline in developing more consistent and objective standards for safety 

goals. These likelihoods may be derived by considering the Commission goal that there should 

be no accidental criticalities at any regulated facility. 

As an example, the value of 10-5 per year per accident in a facility with 100 potential accident 

sequences (Consequence Category 3) would yield a cumulative frequency for Consequence 

Category 3 accidents of: 

100 accidents x 10-5 per year per accident = 10-3 per year. (Eq. A-0) 

These Category 3 accidents generally result in fatalities. The average statistic for all 

manufacturing industries is that a facility with 250 manufacturing workers would expect 10 2 on

the-job deaths per year (see Census Bureau (U.S.), "Statistical Abstract of the United States," 

Table No. 716, 1999). The number of 10.3 per year is consistent with the Commission goal that 

there should be no accidental criticalities at regulated facilities. With approximately 10 

regulated facilities in the United States, this should ensure that the likelihood of an accidental 

criticality anywhere in the country is no greater than 102. A recurrence period of 100 years is 

sufficient to provide reasonable assurance that a criticality accident will not occur during the 

lifetime of any regulated facility. 

Similarly, accident sequences having frequencies more than 4x104 per year per accident are 

considered "not unlikely" (assuming on the order of 100 accident sequences of this type in the 

industry). Again, this value should not be taken as a definitive criterion for acceptability. It is a 

guideline value to ensure consistency. It may need to be adjusted based on the numbers and 

severity of accidents. The rationale for the value 4x1 04 is that accidents of the corresponding 

severity, Consequence Category 2, are not common and should remain so. This is based on a 

Commission strategic goal, that there should be no increase in reportable radiation releases, as

NUREG-1718 A-6
Appendix A

discussed in SRP Section 5.4.3.2(B)(ix). To achieve this, the product of this frequency per 

accident per year with the assessed number of potential accidents should provide adequate 

confidence that such accidents will not occur. Note again that these values of 10-5 and 4x1 04 

are per year per accident. 

The accident evaluation method described below does not preclude the need to comply with the 

double contingency principle for sequences leading to criticality. Although exceptions are 

permitted with compensatory measures, double contingency, should be applied. The reason 

double contingency is needed is the fact that there is usually insufficient firm data as to the 

reliability of the control equipment and administrative control procedures used in criticality 

safety. If only one item were relied on to prevent a criticality and it proved to be less reliable 

than expected, then the first time it failed, a criticality accident could result. For this reason, it is 

prudent to require two independent controls. Inadequate controls can then be determined by 

observing their failure, without also suffering the consequence of a criticality. Even with double 

contingency, it is essential that each of the items relied on for safety (IROFS) be sufficiently 

unlikely to fail, so if one of the two items that establish double contingency is actually 

ineffective, criticality will still not be likely. 

A4. RISK INDEX EVALUATION SUMMARY 

As mentioned in Section A3, an acceptable way for the applicant to present the results of the 

ISA is a tabular summary of the identified accident sequences. Table A-9 is an acceptable 

format for such a table. This table lists several example accident sequences for a powder 

blender at a mixed oxide (MOX) facility. Table A-9 summarizes two sets of information: (1) the 

accident sequences identified in the ISA and (2) a risk index calculated for each sequence to 

show compliance with the regulation. 

A fault tree is another acceptable method of presenting the results. As shown by the example, 

for the purposes of documenting compliance with the double contingency principle, a fault tree 

provides a fuller description of the control systems, and the logical progression of the accident, 

than a tabular format can, and is thus considered the preferred method. Both of these methods 

will be presented in the tables that follow. 

Accident sequences result from initiating events, followed by failure of one or more controls. 

Thus, there are columns in Table A-9 for the initiating event and for controls that may be 

mitigative or preventive. In most cases, the initiating event will be the failure of one of the 

preventive controls. There may also be accident sequences resulting from external events 

such as fires or earthquakes. 

With redundant safety controls, and in certain other cases, there are sequences where an 

initiating event occurs that places the system in a vulnerable state. While the system is in this 

vulnerable state, a safety control must fail for the accident to result. Thus, the frequency of the 

accident depends on the frequency of the first event, the duration of vulnerability, and the 

frequency of the (second) control failure. For this reason, it is necessary to consider the 

duration of the vulnerable state and to assign it a duration index. The values of all index 

numbers for a sequence are added to obtain a total likelihood index, T. Sequences are then

A-7 NUREG-1718
Appendix A

assigned to one of the three likelihood categories of the Risk Matrix depending on the value of 

this index in accordance with Table A-3. 

Table A-3: Determination of Likelihood Category 

LIKELIHOOD CATEGORY LIKELIHOOD INDEX T (= sum of index numbers) 

1 T• -5 

2 -5 < T _ -4 

3 -4<T 

The likelihood category in Table A-3 applies to the accident sequence of a whole and is used to 

assess the overall likelihood of the sequence, not the likelihood of individual controls used in 

meeting double contingency. 

The values of index numbers in sequences are assigned considering the criteria in Tables A-4 

through A-6. Each table applies to a different type of event. Table A-4 applies to events that 

have frequencies of occurrence, such as initiating events and certain control failures. When 

failure probabilities are required for the event, Table A-5 provides the index values. Table A-6 

provides index numbers for durations of failure. These are used in certain accident sequences 

where two controls must simultaneously be in a failed state. In this case, one of the two 

controlled parameters will fail first. It is then necessary to consider the duration that the system 

remains susceptible to failure of the second. The reverse sequence, where the second control 

fails first, should also be considered as a separate accident sequence. (Since the example 

chosen mainly concerns criticality safety, the failure of each control relied on to meet the double 

contingency principle must be considered as the initiating event of an accident sequence.) This 

is necessary because the duration of failure of the second control will usually differ from that of 

the first. The values of these duration indices are not merely judgmental. They are directly 

related to the time interval of surveillance monitoring for failures. That is, the duration of a 

failure is the time until it is detected plus the time to restore the system to a state in which it is 

not vulnerable to the second failure. 

If the probability of failure for the first preventive control is P, (in units of events per year), its 

duration of failure is d, (in years), and the probability and duration of failure of the second 

control is P2 and d2, then P, P2 is the probability that both controls will fail within the year. The 

probability that both controls will be in a failed state simultaneously is P1P2(d1+d). The two 

terms P1P2dl and P1P2d2 correspond to the direct and reverse accident sequences (that is, 

where Control 1 fails first followed by Control 2, and vice versa). Given that the first event has 

occurred (with probability P1), the probability that the second control will fail while the first 

control is in a failed state is P2d1 , assuming the two events are truly independent. Although this 

is a simplistic model, using an approximate constant failure and repair period (P1-1 and dl), 

rather than a realistic distribution, this gives a reasonable estimate of the long-term failure 

probability of the combined control system. The applicant may also use more sophisticated and

NUREG-1718 A-8
Appendix A

rigorous reliability models to determine these quantities. Thus, we see that taking the duration 

index into account can produce a substantial reduction in the overall likelihood of the accident 

sequence. 

For all these index numbers, the more negative the number is, the less likely the failure. 

Accident sequences may consist of varying numbers of events, starting with an initiating event. 

The total likelihood index is the sum of the indices for all the events in the sequence, including 

those for duration. 

Consequences are assigned to one of the three consequence categories of the Risk Matrix 

based on calculations or estimates of the actual consequences of the accident sequence (see 

Table A-i). Multiple types of consequences can result from the same event. The 

consequence category for an event is chosen for the most severe consequence. 

As shown in the first row of Table A-9, the failure duration index can make a large contribution 

to the total likelihood index. Therefore, the reviewer should verify that there is adequate 

justification that the failure will be corrected in the time ascribed to the duration index. In 

general, duration indices with values less than minus one (-1), corresponding to 36 days (about 

1 month), to be acceptable, should be based on the intentional monitoring frequency of the 

process. The failure duration for an unmonitored process will typically be more difficult to 

determine; the duration should be estimated with sufficient conservatism to fully account for the 

uncertainty in the duration (the time needed to notice and correct the event). 

Table A-7 provides a more detailed description of the accident sequences used in the example 

of Table A-9. The reviewer needs the information in Table A-7 to understand the nature of the 

accident sequences listed in Table A-9. Table A-9 lacks sufficient room to explain any but the 

simplest failure events. 

Table A-8 is used to explain the safety controls and external initiating events that appear in the 

accident sequences in Table A-9. The reviewer needs the information in Table A-8 to 

understand why the initiating events and safety controls listed in Table A-9 have the low 

likelihood indices assigned. Thus, Table A-8 needs to address such information as the 

margins to safety limits, the redundancy of a control, and the measures taken to assure 

adequate reliability of a control. Table A-8 must also justify why those external events, which 

are not obviously extremely unlikely, have the low likelihoods that are being relied on for safety. 

The applicant should provide separate tables to list the controls for criticality, chemical, fire, 

radiological, and environmental accidents.

A-9 NUREG-1718
Appendix A

Table A-4: Failure Frequency Index Numbers

FREQUENCY BASED ON EVIDENCE BASED ON TYPE OF CONTROL** COMMENTS 

INDEX NUMBER 

-6* External event with frequency < 10.1 per yr If initiating event, no controls needed 

-4* No failures in 30 years for hundreds of similar Exceptionally robust passive engineered Rarely justified by evidence, since few 

controls in industry control (PEC), or an inherently safe process, or systems are found in such large 

two independent active engineered controls numbers. Further, most types of 

(AECs), PEC, or enhanced administrative single control have been observed to 

controls fail. 

-3* No failures in 30 years for tens of similar A single control with redundant parts, each a 

controls in industry PEC or AEC 

-2' No failure of this type in this facility in 30 years A single PEC 

-1 A few failures may occur during facility lifetime A single AEC, an enhanced administrative 

control, an administrative control with large 

margin, or a redundant administrative control 

0 Failures occur every 1-3 years A single administrative control 

1 Several occurrences per year A frequent event Not for safety controls, just initiating 

events 

2 Occurs every week or more often Frequent event, an inadequate control Not for safety controls, just initiating 

events 

* Numbers less than (more negative than) -1 should not be assigned to controls unless the configuration management, auditing, and 

other management measures are of high quality, because without these measures, the controls may be changed or not maintained. 

** The failure frequency index assigned to a control of a given type in column three may be one value higher or lower than the value 

given in column one, since the reliability of different types of controls can vary widely. Criteria justifying assignment of the lower 

(more negative) failure frequency index should be given in the narrative describing ISA methods. Exceptions should be individually 

justified.

NUREG-1718 A-1 0
Appendix A

Table A-5: Failure Probability Index Numbers

I PROBABILITY INDEX PROBABILITY OF FAILURE BASED ON TYPE OF CONTROL I COMMENTS 

NUMBER ON DEMAND I 

-6* 10*6 If initiating event, no controls needed 

-4 or -5* 10,4-10" Exceptionally robust PEC or an inherently safe Rarely can be justified by evidence, 

process, or two redundant controls better than simple since few systems are found in such 

administrative controls (AEC, PEC, or enhanced large numbers. Further, most types 

administrative) of single controls have been observed 

to fail. 

-3 or -4* 1 0-104 A single PEC or an AEC with high availability 

-2 or -3* 102_10-3 A single AEC, an enhanced administrative control, or 

an administrative control for routine planned operations 

-1 or -2 10*1-10.2 An administrative control that must be performed in 

response to a rare, unplanned demand 

* Probability index numbers less than (more negative than) -1 should not be assigned to controls unless the configuration 

management, auditing, and other management measures are of high quality because without these measures, the controls may be 

changed or not maintained.

A-1 1 NUREG-1718
Appendix A

Figure A-2 presents the same information as a set of fault tree diagrams. A discussion and 

comparison of the two methods follows the example. 

Definitions and explanations of the terms used in the following tables and figures will follow the 

example. 

As an understanding of the example systems is important, process descriptions for hypothetical 

MOX processes follow. These hypothetical systems were chosen because of their relatively 

high degree of importance for nuclear criticality and because they represent the extremes in 

terms of operational and control complexity. The first example, the solvent extraction system, is 

a complex chemical operation that is most amenable to a fault tree presentation of the results of 

the ISA Summary (though to compare the strengths and weaknesses of the two methods, both 

fault trees and a tabular format are presented). The second example, MOX blending, is much 

more straightforward in terms of controls and the results of the ISA can be summarized more 

effectively in terms of a table of accident sequences. 

These examples should only be considered typical of the degree of information required and 

the ways in which it may be displayed. It is anticipated that the applicant's ISA Summary and 

process description may differ markedly from the example. These examples should not be 

construed to preclude other methods of presenting the ISA Summary results. 

A5. OVERALL PROCESS DESCRIPTION 

The purpose of the front-end plutonium purification process (P3) is to remove impurities such as 

gallium and americium from the plutonium oxide feed, producing a more suitable plutonium 

feed stream for the oxide blending process. This process description is for illustrative purposes 

only and should not be expected to conform to any particular applicant's process. The actual 

license application would require a more detailed process description than that presented 

below, but the following brief summary is presented to aid in understanding the example: 

Raw plutonium oxide (PuO 2) powder is received from the shipper and batched into a glovebox 

at the front end of the Aqueous Polishing (AP) processing line. The containers of PuO2 are fed 

into an electrically-heated dissolver unit in the glovebox, consisting of a favorable geometry 

recirculation loop with electrodes at either end. The PuO2 is digested by the addition of nitric 

acid in the presence of Ag÷* ions, resulting in the formation of an impure plutonium nitrate 

(Pu(N0 3)4) solution at a concentration of -250 gPu/l. Plutonium can exist in several oxidation 

states in nitric solutions simultaneously, which complicates the process chemistry considerably. 

Although the plutonium in PuO 2 is tetravalent (Pu(IV)), it undergoes disproportionation, or self

oxidation and reduction, to both Pu(lll) and Pu(VI) through the reaction 

3Pu4

` (IV) + 2H20 - 2Pu3. (111) + PuO 2

2, (VI) + 4H' (Eq. A-i)

NUREG-1718 A-1 2
Appendix A

Table A-6: Failure Duration Index Numbers

NUREG-1718

DURATION INDEX NUMBER AVG. FAILURE DURATION DURATION IN YEARS COMMENTS 

-5 5 minutes 10-5 

-4 1 hour 10.4 

-3 8 hours 0.001 

-2 A few days 0.01 

-1 1 month 0.1 Formal monitoring to justify 

indices less than -1 

0 1 year 1 

1 More than 3 years 10

A-1 3
Appendix A

The plutonium must be adjusted to the tetravalent state to ensure effective extraction. This is 

done as a two-step process. First, Ag++ is generated at the cathode and acts as an oxidation 

agent to drive both Pu(lll) and Pu(IV) to Pu(VI). Tetravalent plutonium is oxidized through the 

reaction Pu(IV) + Ag' - Pu(VI) + Ag. 

After the operators determine that complete PuO 2 dissolution is achieved by means of 

independent dual sampling, the Pu(N0 3)4 is fed through a favorable geometry inline filter into 

the solvent extraction feed preparation slab tank. (N.B. Plutonium in Pu(N0 3)4 is actually in the 

tetravalent state; the chemical form after oxidation is more accurately characterized as a 

mixture of Pu(VI) cations in a N0 3 -rich solution.) The free Pu(VI), or PuO 2÷

2 plutonyl ions, 

must be adjusted from the hexavalent to the tetravalent state Pu(IV) by the addition of excess 

HNO 3 and H20 2 (a reducing agent) at a low plutonium concentration. This is done in the 

favorable geometry preparation tank. The entire aqueous polishing process is conducted on a 

batch basis, with approximately 14 kg (30.8 Ib) Pu processed through dissolution, solvent 

extraction, precipitation, and calcination in each batch. The powder is then mixed together with 

natural uranium oxide to form the master blend. 

A6. SOLVENT EXTRACTION PROCESS (PLUTONIUM PURIFICATION) 

The solvent extraction, scrub, and strip columns consist of identical long (-20 feet [6.1 m]), 5

inch (12.7 cm) diameter Pyrex columns containing a series of stationary perforated plates. For 

solvent extraction, the aqueous Pu(N0 3) 4 solution is added at the top of the column and a 

mixture of TBP, or tributyl phosphate (chemical formula (C4H9)3P0 4), and a diluent (30 percent 

hydrogenated tetrapropylene, or HTP) is added at the bottom of the column. The difference 

between the relative specific gravities of the two streams causes the aqueous solution to sink to 

the bottom and the organic mixture to rise to the top of the column. The immiscible fluids are 

pulsed in the columns by means of positive-displacement pumps. This pulsing breaks up the 

interface between the fluids and increases the surface area, resulting in intimate mixing to 

increase the efficiency of extraction. The tetravalent plutonium ion Pu4+ becomes complexed to 

the organic through the reaction. 

Pu' (aq) + 4NO3(aq) + 2TBP(o) - Pu(N0 3)4-2TBP(o) (Eq. A-2) 

The existence of a salting agent such as HNO 3 or AI(NO 3)3 increases the acid molarity of the 

excess nitric ion and causes the above reaction to be shifted to the right. 

In the scrub column, the fissile-bearing organic stream from the top of the solvent extraction 

column is fed into the bottom of the scrub column. Additional nitric acid is added to the top of 

the scrub column and the same countercurrent operation repeated to remove additional 

impurities from the organic into the aqueous phase. The plutonium remains in the organic 

phase at the end of the scrub operation. The aqueous raffinate stream-which should now 

contain low levels of plutonium but concentrated fission products-is transferred to raffinate 

storage while the fissile-bearing organic stream is fed into the bottom of the strip column. 

In the strip column, de-ionized water is added to reduce the acid molarity, causing the left-hand 

side of equation A-2 to be favored. The aqueous product stream containing purified plutonium

NUREG-1718 A-1 4
Appendix A

nitrate is then transferred to the first-pass evaporator; the spent organic phase must be 

reconditioned for reuse in the first-pass solvent extraction. The evaporator consists of a tube

and-shell heat exchanger in which the concentration of the Pu(N0 3)4 increases from 40 gPu/I to 

approximately 250 gPu/I. The second- and third-pass solvent extraction lines are nearly 

identical to the first, except that the Pu(NO3 )4 is at a higher concentration. 

Raffinate and solvent conditioning streams attach to the process at various points. The first-, 

second-, and third-pass raffinate streams are transferred to a bank of favorable geometry 

columns (from the extraction and scrub columns, solvent regeneration, and evaporator 

condensate), where they are sampled (by dual independent sampling) for Pu content. Only 

after they meet the release criteria of 0.015 gPu/I are the contents discharged (through an inline 

monitor that is interlocked to the waste tank isolation valves) to a set of unfavorable geometry 

wastewater tanks for wastewater treatment and eventual discharge from the site. In addition, 

organic solvent from the strip columns must be regenerated because it contains a buildup of 

metallic impurities (primarily gallium and americium), nitric acid (acquired through the reaction 

H+(aq) + N0 3 (aq) + 2TBP(o) - HNO 3-TBP(o)), and various radiolytic decomposition products of 

TBP and kerosene, such as dibutyl phosphate (DBP). The solvent is washed with Na 2CO3, 

NaOH, and HNO 3 in a series of favorable geometry mixer/settlers (M/Ss) to remove impurities, 

filtered, and recycled to the solvent extraction columns. Gallium and americium are further 

removed by electrolytic deposition on charged plates in the M/S units. Makeup solvent from 

bulk chemical tanks is added as needed to maintain the solvent inventory. The M/Ss consist of 

a safe geometry box partitioned by a short wall into a mixing chamber and a settling chamber. 

The mixing chamber contains a rotary impeller that draws the heavier liquid (aqueous wash 

solution) from the bottom of the mixing chamber and emulsifies it into the lighter liquid (organic 

solvent) in the top of the mixing chamber. Following this intimate mixing (which operates under 

the same principle as the pulsed extraction columns), the solution gravity drains into the settling 

chamber, where it separates into two distinct layers. The organic is drawn off to the next wash 

stage or to the fresh solvent column, while the aqueous is discharged to the raffinate storage 

columns. 

Following third-pass solvent extraction, the purified Pu(NO3 )4 must be reconverted to PuO2 for 

blending with U0 2. This is accomplished by the addition of oxalic acid (H2C20 4) to cause the 

precipitation of plutonium as plutonium oxalate (Pu(C 20 4)2). Hydrogen peroxide (H20 2) is added 

to the plutonium nitrate solution to ensure that it is in the proper oxidation state. After sampling, 

the solution is transferred to the precipitation column, a short 4-inch (10.2 cm) diameter glass 

column contained within a glovebox, in which the plutonium oxalate is prepared. Precipitation 

proceeds through the reaction. 

Pu(IV)(NO3 )4 + 2H2C20 4 = Pu(C 20 4)2 + 4HN0 3 (Eq. A-3) 

The resulting precipitate is prepared through the slow addition of oxalic acid to the column, and 

is thixotropic in nature. The nitric acid content must also be adjusted to obtain the desired level 

of consistency. The resultant plutonium oxalate slurry collects at the bottom of the column. 

The residual nitric-water solution contains only low levels of plutonium nitrate and is sampled for 

discharge. Solutions that contain greater than the release criteria of 0.015 gPu/I are recycled to 

solvent extraction for re-extraction. This dilute nitric solution is decanted and filtered before

A-1 5 NUREG-1718
Appendix A

transfer to acid recovery, and the material at the bottom of the bowl is drained out before being 

air-dried in the glovebox. When the material is dried, it forms a cake containing plutonium 

oxalate hydrates (such as Pu(C20 4)2-6H20). The material is gravity drained from the bottom of 

the precipitator, where it is automatically dropped through a chute into an inclined, rotary-kiln 

calciner in a continuous process. The slurry is then calcined in an electrically heated oxidation 

furnace at 300 °C and then converted to PuO2 at 900 °C in an oxygen-rich atmosphere in the 

same furnace. The PuO 2 is collected into a moderation-controlled hopper, which is connected 

to a favorable geometry tumbling mixer to achieve proper homogenization. The mixer consists 

of two rotating drums with a spiral blade in the intervening space with a cadmium shaft for 

neutron poison. After homogenization, the material is transferred to a glovebox where it is 

sampled, bottled, and transferred to the mixed oxide blending operation of the MOX process 

(MP) line. 

The overall process flow is shown in Figure A-i. 

Controlled parameters in the solvent extraction process are geometry, concentration, spacing, 

interstitial moderation, and process variables (material form). The solvent extraction columns 

were modeled using an optimal plutonium nitrate concentration of 140 GPU/I, without taking 

credit for the presence of gallium-a mild neutron poison-or excess nitric acid. The solution 

was modeled to the outer diameter of the columns, and thus took credit for the diameter but not 

the column thickness. Credit was not taken for the plutonium isotonic (-4 wt% 240Pu), as the 

models assume the feed consists solely of 239Pu. Concentration was not controlled for the 

extraction columns, but was credited for keeping the waste tanks subclinical upon solution 

transfer from the refined storage columns. 

Because the design relies primarily on passive engineered features (i.e., fixed geometry and 

spacing), the potential for nuclear criticality in the solvent extraction system itself is extremely 

unlikely. The main accidents of concern are transfer of concentrated solution to unfavorable 

geometry process equipment. As shown in Figure A-i, the unfavorable geometry systems that 

are connected to the process consist of (i) steam supply for the evaporators; (ii) deionize water, 

nitric acid, and solvent regeneration bulk chemical supply tanks; (iii) wastewater system tanks; 

and (iv) the floor. 

The example shown in the following tables is for the second-pass solvent extraction (2SX) in 

the p 3 process node. The list of accident sequences and controls is for illustrative purposes 

only and is not meant to be exhaustive. 

1. Process Criticality Flow Diagram 

Figure A-1 is an example of one method of describing the process flow. A good 

understanding of the process flow and the criticality control systems that exist at each 

node in the process is essential to evaluating the results contained in the ISA 

Summary. The information contained in this process criticality flow diagram (PCFD) is 

a more condensed form of the information that would be expected in the process 

description, process flow diagrams (PFDs), and criticality safety evaluations. 

Presenting the information in this way is advantageous to the applicant, as it is a more

NUREG-1718 A-1 6
Appendix A

efficient means of providing needed process knowledge to the ISA or safety discipline 

reviewer. Basically, the PCFD is a PFD modified to contain the features relied on for 

criticality safety. Note several useful features of this diagram: 

The different process steps are divided into two categories by shape, those relying on 

favorable geometry and those that are unfavorable geometry. Distinguishing between 

these two types of systems may be done by several other means. Geometry control is 

typically ranked as the most preferable control due to its inherent stability and 

robustness, and it is the primary control relied on in most of this particular system. In 

certain other systems, it may be somewhat more advantageous to draw a distinction 

between process steps that are moderation controlled and uncontrolled areas, or 

between concentration-controlled and uncontrolled areas. By reviewing this diagram, it 

is immediately apparent where the transition from favorable to unfavorable geometry 

takes place. 

Another feature of this diagram is that the engineered features relied on for criticality 

safety are clearly identified by shading. There is a simple graphic representation of the 

barriers that exist between favorable and unfavorable geometry equipment, which are 

drawn as bars across the flow path between these systems. This makes readily 

apparent the features that prevent the backflow of concentrated fissile solution to 

unfavorable geometry equipment, among other scenarios. Adding the labels that 

correspond to each of the IROFS (as in Table A-8) would provide a ready cross

reference, but may result in too much added complexity for such a system. 

The use of different line patterns to distinguish between the various 

streams-particularly with respect to different fissile compositions-facilitates 

understanding of the process flow. Another useful feature is the division of the entire 

diagram into different zones corresponding to various process nodes. This provides a 

clearer boundary definition and allows the reviewer to see how the system functions 

together as an integrated whole, including how perturbations in criticality controls in 

one process node or piece of equipment flow down into the next. The engineered 

controls tabulated in the ISA Summary (such as Table A-8) should include all features 

relied on for safety within the boundary of that process node. Finally, this diagram 

displays the actual controlled parameters at each process step; to the degree possible, 

this should be extended to display the actual controlled values of those parameters. 

This diagram should be consulted in reviewing the sample tables.

A-1 7 NUREG-1718
Appendix A

Figure A-I: Criticality Flow Diagram for 

PuO2 GVR 

PuO2 Dissolution Ag++ Pu02 

Nod.: WSS 

LEGEND Shapes

NUREG-1718 A-1 8
Appendix A

Table A-7: Accident Sequence Descriptions

NUREG-1718

Accident Sequence (see Table A-9) DESCRIPTION 

Loss of Geometry 

2SX-001 The initiating event is the leakage of concentrated plutonium nitrate solution through the safe geometry chamber of the column pulsing pumps into the oil reservoir, 

where it becomes moderated by oil. This cannot result in a criticality without failure of the oil reservoir level switch. 

2SX-002 The initiating event is the leakage of concentrated plutonium nitrate solution through the shell side of the heat exchanger in the second-pass evaporator. 

Concentrated solution could ultimately reach the unfavorable geometry steam supply after multiple failures, leading to criticality. 

2SX-003 The initiating event Is the backflow of concentrated plutonium nitrate solution to the unfavorable geometry nitric acid supply, which could lead to criticality after 

multiple failures. The backflow could result from a leak of the heat exchanger through the tube side of the second-pass evaporator, from over-addition of de-lonized 

water or other reagents, from double batching of plutonium nitrate into the columns, or from vigorous chemical reactions within the columns. 

2SX-004 The Initiating event is the backflow of concentrated plutonium nitrate solution to the unfavorable geometry de-lonized water supply, which could lead to criticality after 

multiple failures. (See 2SX-003 for a discussion of possible backllow mechanisms.) 

2SX-005 The Initiating event is the failure to strip the plutonium from the spent organic solvent, resulting in the presence of concentrated plutonium In the solvent regeneration. 

Although this equipment Is analyzed safe at optimal concentration, the initial upset, followed by backflow to the fresh solvent bulk chemical supply, could lead to 

criticality. 

2SX-006 The Initiating event Is the failure of the solvent extraction columns, leading to buildup of concentrated plutonium nitrate solution on the process floor. This would 

have to be followed by collection in an unfavorable geometry container to lead to criticality. 

Loss of Concentration 

2SX-007 The Initiating event is the failure to extract the plutonium from the aqueous raflinate stream. Although the raffinate storage columns are analyzed safe at optimal 

concentration, the Initial upset, followed by failure of the concentration controls, could lead to criticality due to transfer of concentrated solution to the wastewater 

tanks. 

Loss of Spacing 

2SX-008 The initiating event is an earthquake, fire, or other catastrophic event causing two or more columns to come together through structural failure. This would have to 

be followed by complete flooding to achieve criticality. 

2SX-009 The initiating event Is that columns are Installed closer together than that analyzed as safe. This would have to be followed by complete flooding to achieve 

criticality.

A-1 9
Appendix A

NUREG-1718

Accident Sequence (see Table A-9) DESCRIPTION 

2SX-1 0 The initiating event Is the placement of a container with plutonium nitrate or oxide placed next to a column or pump filled with plutonium nitrate solution. This would 

have to be followed by complete flooding to achieve criticality. 

Loss of Interstitial Moderation/Reflection 

2SX-1 1 The initiating event Is the Increase In reflection above nominal values. Optimal interstitial moderation is shown to be safe under normal conditions. However, certain 

upset conditions require limited interstitial moderation to demonstrate subcriticality. 

Loss of Material Form 

2SX-12 The initiating event is the introduction of more reactive forms of plutonium in the solvent extraction process, through the intrusion of oil from the pulsed column 

pumps' oil reservoirs. As a single failure this would not lead to criticality. However, It could result In nuclear criticality if followed by loss of interstitial moderation 

control. 

2SX-13 The initiating event Is the introduction of more reactive forms of plutonium in the solvent extraction process through the addition of precipitating agents (such as 

oxalic acid) into the plutonium nitrate feed. 

2SX-14 The initiating event is the introduction of more reactive forms of plutonium in the solvent extraction process through the backflow of precipitating agents from the 

precipitation columns.

A-20
Appendix A

2. Determination of Likelihood Category in Table A-3 

The likelihood category is determined by calculating the likelihood index, T, then using 

this table. The term "T" is calculated as the sum of the indices for the events in the 

accident sequence. 

3. Determination of Failure Frequency Index Numbers in Table A-4 

Table A-4 is used to assign frequency index numbers to facility initiating events and 

control system failures as found in the columns of Table A-9. The term "failure" is 

understood to mean not merely failure of the control device or procedure, but also the 

violation of the safety limit by the process. Since the controls are designed to maintain 

the system within its safety limits, any violation of the safety limits would also involve 

violation of the associated controls. The converse, however, is not necessarily true. In 

the example in Table A-9, accident sequence 2SX-001 involves loss of volume control 

due to pump failure. If criticality is the concern, failure leading to a criticality does not 

occur unless an unsafe volume of uranium-oil mixture actually collects in the oil 

reservoir before the leak is stopped. Upon violation of the control, there is some finite 

likelihood that the safety limit on the associated controlled parameter will be exceeded, 

and this should be considered when estimating the overall probability that an accident 

sequence will actually lead to criticality. (In practice, there are often cases for which 

the probability distribution of parameter values upon failure of the control is not well 

known, and the likelihood of the failure is then often conservatively assumed to be 

equivalent to the likelihood of the control failure.) This credit should be based on hard 

process data. In assessing the frequency index, this factor should be considered 

because many control failures do not cause safety limits to be exceeded. For 

radiological consequences, any amount leaked may cause exposure. 

Table A-4 provides two columns with two sets of criteria for assigning an index value, 

one based on type of control, the other based directly on observed failure frequencies. 

The types of controls are administrative, active engineered, passive engineered, etc. 

Since controls of a given type have a wide range of failure frequencies, assignment of 

index values based on this table should be done with caution. Due consideration 

should be given to whether the control will actually achieve the corresponding failure 

frequency in the next column. Based on operational experience, more refined criteria 

for judging failure frequencies may be developed by an individual applicant. In the 

column labeled "Based on Type of Control," references to redundancy allow for 

controls that may themselves have internal redundancy to achieve a necessary level of 

reliability. 

Another objective basis for assignment of an index value is actual observations of 

failure events. These actual events may have occurred in a comparable process 

elsewhere or in the licensed facility. Justification for specific assignments may be 

noted in the Comments column of Table A-9.

A-21 NUREG-1 718
Appendix A

As previously noted, the definition of failure of a safety control to be used in assigning 

indices is, for nonredundant controls, a failure severe enough to cause an accident 

with consequences. For redundant controls, it is a failure such that, if no credit is 

taken for functionality of the other control, an accident with consequences would result. 

If most control malfunctions would qualify as such failures, then the index assignments 

of this table are appropriate. If true failure is substantially less frequent, then credit 

should be taken and adequate justification provided. 

Note that indices less than (more negative than) -1 should not be assigned to controls 

unless the configuration management, auditing, and other required management 

measures are of high quality because, without these measures, the controls may be 

changed or inadequately maintained. The reviewer should be able to determine this 

from a tabular summary of safety controls provided in the application. This summary 

should include identification of the process parameters to be controlled and their safety 

limits and a thorough description of the control and its applied management measures. 

4. Determination of Failure Probability Index Numbers in Table A-5 

Occasionally, information concerning the reliability of a safety control may be available 

as a probability on demand. That is, a history may exist of tests or incidents during 

which the system in question is demanded to function. To quantify such accident 

sequences, the applicant must know the demand frequency, the initiating event, and 

the demand failure probability of the safety control. This table provides an assignment 

of index numbers for such controls in a way that is consistent with Table A-4. The 

probability of failure on demand may be the likelihood that it is in a failed state when 

demanded (availability), or that it fails to remain functional for a sufficient time to 

complete its mission. 

5. Determination of Failure Duration Index Numbers in Table A-6 

The failure duration index is important because of reasons discussed above. It 

represents the window of opportunity after failure of the first preventive control, during 

which the failure of the second could lead to adverse consequences. Cases in which 

the loss of the first control may remain undetected for long periods of time (such as 

leakage of hidden or baffled piping when credited as primary containment, or failure of 

items tested only when challenged) will typically not credit the failure duration in 

reducing the probability of the accident scenario. In this case, the duration index D 

should be taken as 0. Duration indices of less than -1 should be based on periodic 

surveillance and/or maintenance periods, or the fact that failure would be readily 

apparent within a certain time frame. For example, a failure duration index of -2 would 

be based on the fact that a weekly surveillance requirement has been established for 

that item. A failure duration index of -3 may be based on a requirement to perform a 

certain measurement once per shift, or the fact that failure would immediately reveal 

itself to operators who are required to be continually present.

NUREG-1718 A-22
Appendix A

Table A-8: Criticality Safety Limits and Controls

ROFS for the Second-Pass Solvent Extraction system

NUREG-1718

IROFS Parameters IROFS Description Management QA 

Identifier* and Limits** I I Measures*- Grade*.... 

2SX-PE1 VOLUME: SX Pump PMPX-001 has a safe volume chamber. Configuration control B 

< 4.5 1(1.2 gal) 

2SX-AE1 GEOMETRY: SX Pump PMPX-001 has an active level switch on 1. Configuration control A 

< 7.6 cm (3") the oil reservoir, which automatically shuts the 2. Control room con depth recirculation valve and sounds an audible alarm in stantly monitored 

the control room if the slab depth is exceeded. 3. Biweekly functional 

check 

2SX-AE2 PROCESS Pressure differential gauge on heat exchanger 1. Configuration control A 

VAR: HX-001 is set to alarm if P, - P, < 0.34 atm (5 2. Control room con

AP < 0.34 atm psi). stantly monitored 

(5 psi) 3. Functional check each 

shift 

2SX-AE3 PROCESS Conductivity probe on heat exchanger shell side to 1. Configuration control A 

VAR: detect intrusion of plutonium. Setpoint will be 2. Functional test weekly 

not applicable, sufficient to detect a concentration of 0.1 gPu/l. 

2SX-ADM1 PROCESS Procedures require operator response to 1. SOP 5349 B 

VAR: differential pressure gauge alarm. 2. Tralning/postings 

AP < -0.34 atm 

(-5psi) 

2SX-PE2 MASS: Siphon break installed in nitric acid supply line. 1. Configuration control C 

0 mass in nitric 

acid supply 

2SX-ADM2 MASS: Utility (in this case, nitric acid) supply gauges are 1. SOP 9483 B 

0 mass in nitric continually monitored in the control room 2. Tralning/postings 

acid supply whenever fissionable material is being processed. 

Facility procedures require shutdown when utility 

pressure lost. 

2SX-PE3 MASS: Siphon break installed on Deionize Water (DIW) 1. Configuration control C 

0 mass in DIW line. 

supply 

2SX-ADM3 MASS: Utility (in this case, DIW) supply gauges are 1. SOP 6879 A 

0 mass in DIW continually monitored in the control room 2. Training/postings 

supply whenever fissionable material is being processed. Facility procedures require shut down when utility 

pressure lost. 

2SX-ADM4 PROCESS Acid molarity is adjusted to _<O.1M in the strip 1. SOP 0292 A 

VAR: column. This ensures the plutonium will be 2. Training/postings 

Acid stripped back into the aqueous phase. 

moladty•0.05.

A-23
Appendix A

IROFS Parameters IROFS Description Management QA 

Identifier* I and Limits- I I Measures*** Grade**... 

2SX-ADM5 CONCENTR: Procedures require weekly check of solvent 1. SOP1929 B 

< 0.1 gPu/I in regeneration columns by dual independent 2. Training/postings 

the solvent sampling. In addition, at the start of each batch, a 3. QA Lab procedure 

regeneration checklist requires operators to visually check the ensures independent 

columns columns for observed plutonium intrusion 

(greenish color). 

2SX-PE4 MASS: Backflow preventer (BFP) installed on bulk 1. Configuration control B 

0 mass in bulk chemical and DIW lines to prevent backflow to 2. Annual surveillance 

chemical supply organic solvent supply tanks. 

2SX-PE5 GEOMETRY: Columns must be composed of no greater than 1. Configuration control C 

diam < 10.2 cm 10.2 cm (4") diameter glass (extraction, scrub, 

(4") strip, and precipitation). 

2SX-PE6 GEOMETRY: Catch pans beneath columns must be no more 1. Configuration control C 

depth < 5.2 cm than 5.2 cm (2") deep. In addition, they must have 

(2") an area of 4.65 m2 (50 ft2

) or more to ensure that 

Area > 4.65 m2 they are capable of handling the largest spill from 

(50 ft2) the columns. 

2SX-PE7 SPACING: Drawings require columns be installed no more 1. Configuration control C 

columns > 61 than 61 cm (24") c-to-c. 

cm (24") center

to-center (c-to-c) 

2SX-ADM6 MODERATION: Facility emergency response procedures prohibit 1. Emergency Plan. C 

water not the use of water in fighting fires in the solvent 2. Training/postings. 

allowed in extraction area when plutonium is being 3. Annual drill. 

fighting fires processed. There is no automatic sprinkler 4. Configuration control. 

system in this area. Foams and fogging agents 

may be used. 

2SX-PE8 MODERATION: Overhead water lines are prohibited in the solvent 1. Configuration control. C 

no overhead extraction area. 

lines in SX area 

2SX-PE9 GEOMETRY: Width of the solvent regeneration M/Ss must be 1. Configuration control. C 

width < 7.6 cm less than 7.6 cm (3"). 

(3") 

2SX-PE10 GEOMETRY: Diameter of the floor drains must be less than 7.6 1. Configuration control. C 

diameter < 7.6 cm (3"). 

cm (3 ") 

2SX-PE1 1 GEOMETRY: Floor must be sloped to drain into the favorable 1. Configuration control. C 

depth < 2.54 cm geometry floor drains; variation in floor level must 2. Annual audit. 

(1") not allow solution more than 2.54 cm (1") deep to 

accumulate. 

2SX-AE4 CONCEPT: Inline monitor interlocked to isolation valve to 1. Weekly calibration and A 

< 0.015 gPuAl terminate feed if concentration exceeds limit, functional source check. 

Safety grade items are the monitor, the interlock 2. Configuration control. 

electronics, and the isolation valve.

NUREG-1718 A-24
Appendix A

IROFS Parameters IROFS Description Management QA Identifier* and Limits-* Measures.*. Grade*.... 

2SX-ADM7 CONCEPT: Dual independent samples must be drawn and 1. SOP 9045 A 

< 0.015 gPu/I confirmed before transfer of refined to the 2. QA Lab procedure 

wastewater tanks is permitted. The results of 3. Training/postings 

sampling must be reviewed by the operator and a 

supervisor (who maintains control of the key to the 

valve lock). 

2SX-PE12 SPACING: Structural supports must be designed to withstand 1. Prestart-up load C 

columns > 61 credible loads with a safety factor Ž 2. Must be testing. 

cm (24") c-to-c designed to withstand seismic loads >0.1g. 2. Configuration control. 

2SX-ADM8 CONCEPT: Acid molarity is adjusted to approximately 3M (via 1. SOP 3934 B 

< 0.015 gPu/I grab sample) in the strip column to ensure Pu 2. Lab QA procedure 

separation. PH may be used as a secondary 

measurement. Needed to keep refined 

concentration at a sufficiently low level. 

2SX-ADM9 CONCEPT: Concentration in second pass extraction limited to 1. SOP 0945 B 

< 0.015 gPu/A 350 gPu/1. Along with 2SX-ADM8, needed to 2. Lab QA procedure 

ensure extraction efficiency to keep refined 

concentration sufficiently low. 

2SX-ADM10 SPACING: Facility procedures require that fissile material 1. Supervisor walk- C 

containers > containers and portable equipment be maintained through 

30.5 cm (12") at least 30.5 cm (12") from columns and pumps. during each shift. 

from columns Reinforced by postings and blue lines painted on 2. Facility directive 07. 

floor (Limited Movement Areas). 3. Training/postings. 

2SX-ADM1 1 MATERIAL The amount of oil in the oil reservoir of any pump 1. Configuration control. C 

FORM: shall be limited to 4 1 (1.1 gal). This limits the 

oil <4 1 (1.1 gal) concentration of hydrogenous moderators other 

than water to ensure subclinical calculations are 

bounding. 

2SX-ADM12 MATATERIAL Uds to bulk chemical supply tanks must be locked 1. Facility directive 29. C 

FORM: and controlled by supervisors, to ensure against 2. Training/postings. 

no precipitating the inadvertent addition of precipitating agents. 

agents Addition of all reagents must be certified by a 

facility chemical engineer prior to fissionable 

material processing. 

2SX-PE13 MATATERIAL BFP installed on the line connecting the 1. Configuration control. B 

FORM: precipitation columns and the second pass 2. Annual surveill. 

no precipitating evaporator. This prevents the backflow of oxalic 

agents acid into the SX operation. 

* IROFS Identifier: cross-referenced with Preventive Controls in Table A-9. 

** Parameters and Limits: describe actual parameter limits, and all attributes of the IROFS that 

are important to criticality safety. 

*** Management Measures: These are the measures needed to ensure IROFS availability and 

reliability. 

**** QA Grade: This is optional; all controls may be classified as Grade A. If there is a graded 

QA Program, this signifies not the relative safety significance of the control, but the degree of 

management attention needed once the item is installed to ensure its availability and reliability 

(e.g., the siphon break is Grade C, not because its failure is of minor NCS significance, but 

because once installed it requires essentially no maintenance.)

A-25 NUREG-1718
Appendix A

Note: Engineered features such as alarms and instrumentation needed to trigger an 

administrative response should be categorized as separate IROFS from the administrative 

controls; these design features are required to be maintained as IROFS. 

6. Determining Management Measures for Safety Controls 

Table A-8 is an acceptable way of listing those IROFS in all the accident sequences 

leading to consequences that do not comply with the performance requirements. The 

IROFS listed should include all safety controls and all external events whose low 

likelihood is relied on to meet the performance requirements of 10 CFR 70.61. Staff 

reviews this list to determine whether measures have been applied to each safety 

control adequate to ensure their continual availability and reliability in conformance to 

10 CFR 70.62(d). The types of management measures include maintenance, training, 

configuration management, audits and assessments, quality assurance, etc. These 

management measures are indicated in the Baseline Design Criteria (BDC) and 

described in greater detail in SRP Chapters 6.0 through 12.0 and 15.0. Safety controls 

meeting all the provisions of these chapters have acceptable management measures, 

that is, they comply with § 70.62(d). Safety controls may, with justification, have lesser 

management measures than those described. However, every item relied on for 

safety in accident sequences leading to consequence categories 2 or 3 should be 

assigned at least a minimal set of management measures. Specifically, to defend 

against common mode failure of all controls on a process, this minimal set of 

measures must include an adequate degree of: (a) configuration management; (b) 

regular auditing for the continued effectiveness of the control; (c) adequate labeling, 

training, or written procedures to assure the awareness of the operating staff of the 

safety function performed; (d) surveillance and corrective maintenance; and (e) 

preventive maintenance, if applicable. 

If lesser or graded management measures are applied to some controls, Tables A-8 

and A-9 and the narratives preceding them, to be acceptable, must identify to which 

controls these lesser measures are applied. In addition, information indicating that 

acceptable reliability can be achieved with these lesser measures must be presented. 

It is not necessary that the specifics of these measures, such as the surveillance 

interval, type of maintenance, or type of testing, be described as applied to each 

control. It is recognized that such specific measures must be applied differently to 

each control to whatever degree is necessary to achieve adequate reliability. It is the 

formality, documentation, and QA requirements applied to these direct management 

measures that may be graded generically in a risk-informed manner. 

The following describes the application of management measures to IROFS based on 

the risk importance of the item in an accident sequence, as defined by (1) the risk 

index, and (2) the failure likelihood index, "T." In summary, items relied on to prevent 

or mitigate accidents with consequences in the two highest categories identified in 

§ 70.61 should satisfy the applicable BDC of § 70.64.

NUREG-1718 A-26
Appendix A

For each of the accident sequences evaluated in Table A-9 as being in an acceptable 

risk category (a risk index of less than or equal to 4): 

(1) If the initiating event is not a control failure, then management measures for that 

event are not necessary. For sequences claimed to be highly unlikely or unlikely, 

the assessment that the initiating event has such a low frequency must be 

adequately justified in the application. 

(2) Regardless of the degree to which this initiating event is relied on in the accident 

sequence, for accident sequences resulting in nuclear criticality, double 

contingency should still be established. This requires at least one more IROFS in 

the accident sequence, in addition to the initiating event, that requires 

management measures to ensure compliance with the double contingency 

principle. 

(3) If the initiating event is a control failure, management measures for that IROFS 

should be applied sufficient to maintain the claimed failure frequency. The 

selection of management measures should take into consideration the failure 

likelihood assumed in finding the accident sequence risk acceptable, as well as 

the inherent nature of the control. 

[Basis: If the required failure frequency index for a control with management 

measures applied (assumed in the accident sequence) is comparable to the 

failure index without management measures, such as for rigid dimensions of 

equipment not susceptible to changing, a relatively low level of management 

measures may be warranted.] 

(4) If the initiating event is a control failure, management measures may be graded 

less than the highest level depending on the importance of the control to the 

overall risk of the accident sequence. 

[Basis: If the unavailability of the IROFS makes a negligible increase in the overall 

risk, then that IROFS has a relatively low importance in the accident sequence. 

Assigning weights to the various IROFS in terms of management measures may 

be done by comparing the overall risk with and without (mitigated versus 

unmitigated) that particular IROFS.] 

7. Risk-Informed Review of Safety Controls 

The staff reviews the safety controls and external events listed in Table A-8 in a risk

informed manner as described in Section 5.5. The procedure for identifying systems 

of safety controls having higher risk significance is described in Section 5.5. These 

controls will be subject to a more detailed review by staff to ensure their adequacy.

A-27 NUREG-1718
Appendix A

Table A-9: Example Accident Sequence Summary and Risk Index Assignment

P3 (Ph itnniiim PItrifictiAn Pm Node: 2SX ('Second-Pass Solvent Extraction)

NUREG-1718

Accident Initiating Event Preventive Preventive Likelihood' Conse- Risk Comments 

Sequence (a) Control 1 Control 2 Index T and quence Indices & 

(b) (c) Category C Category (I = d x e) Recommendations 

(d) (e) 

2SX-001 Pump chamber leaks 2SX-PE1: Pump chamber is safe volume 2SX-AEI: Level switch keeps oil level at safe T =-5 3 3 

F1 = -1. Regular maintenance prevents frequent pump slab depth. Automatically actuates isolation 

failure, valve and alarms if level exceeded. C = 1 

D1 = -3. Pump failure would be detected by oil F2 = -2. Regular maintenance ensures low 

presence in the solution in clear glass columns. failure rate. 

Process continually monitored by operators. D2 = -2. Biweekly surveillance. I 

2SX-002 Heat exchanger tube 2SX-AE2: Differential gauge alarms if pressure 2SX-ADMI: Operator required to respond to T = -4 3 6 This scenario requires 

leaks differential across evaporator not maintained, alarm if pressure differential lost. other controls to ensure 

F1 = -1. Regular maintenance ensures low failure rate. F2 = -2. Failure to evaporate would be noticed C = 2 adequate low 

Df = -3. Failure would be detected during one shift by operators on floor, and control room operator likelihood. 

because concentration monitored frequently for CA. required by training and procedure to respond 

to alarm. Control room manned by two Recommend 

operators at all times, installation of a drain 

D2 = 0. Failure of this control may not be line on the steam 

readily noted. Credit not taken. supply to prevent liquid 

accumulation. 

2SX-003 Motive force causes 2SX-PE2: Siphon break Installed on supply line. 2SX-ADM2: Utility supply pressure not T =5 3 3 

potential backflow to F1 = -4. The most likely scenario is that the siphon maintained above atmospheric. 

nitric acid break was never Installed In the first place. There is a F2 = -2, Utilities used throughout facility for C = 1 

rigorous configuration control program for safety grade many different purposes. They are used 

items. frequently and so are tested on a continual 

D1 = 0. All safety grade items audited annually to basis. 

confirm their continued presence. D2 = -2. Control room continually manned; 

these process variables are monitored 

constantly for GA purposes. 

2SX-004 Motive force causes 2SX-PE3: Siphon break installed on supply line. 2SX-ADM3: Utility supply pressure not T = 5 3 3 

potential backflow to see 2SX-003 for explanation, maintained above atmospheric. 

DIW see 2SX-003 for explanation. C = 1

A-28
Appendix A

Accident Initiating Event Preventive Preventive Likelihood' Conse- Risk Comments 

Sequence (a) Control 1 Control 2 Index T and quence Indices & 

(b) (c) Category C Category (f = d x e) Recommendations 

(d) (e) 

2SX-005 Concentrated 2SX-ADM4: Sufficient DIW added to ensure acid 2SX-PE4: BFP on organic bulk chemical T = -8 3 3 

plutonium not molarity low enough to guarantee stripping. Must be supply line. 

stripped from organic sampled and checked before stripping. F2 = -2. Regular maintenance ensures low C = 1 

2SX-ADM5: Solvent regeneration columns M/Ss failure rate. 

monitored weekly for uranium buildup. D2 = -2. Failure would be detected during 

F1 = -2. Process variables (acid molarity and weekly surveillance. 

concentration) monitored shiftly and monitored 

continuously for QA purposes, ensuring their reliability 

DI = -3. Major process upset would be noted by 

operators almost immediately. 

2SX-006 Solution spills from 2SX-PE5: Columns are favorable geometry glass, 2SX-PE6: Catch pans are safe slab and have T -7 3 3 

column F1 = -1, Columns have capacity to break even though sufficient area to hold the contents of more than 

they are sealed within a steel scaffold; operational two columns when filled to the maximum. C =1 

history shows that this Is an infrequent occurrence. F2 = -4. For this control to fail would require 

DI = -3. Breakage would be readily apparent. The either improper installation, or the breakage of 

process floor is continually manned and good several columns. Configuration management 

housekeeping practices are instituted, reliability is judged to be -4. 

D2 = -3. See Control 1 explanation. 

...additional accident sequences would follow this... 

2SX-008 Earthquake occurs of 2SX-PE7: Columns separated at sufficient distance to 2SX-PE8: There are no water lines or other T = -12 3 3 This scenario takes 

sufficient strength to ensure subcriticality. sources of water installed to burst in the event credit for an external 

cause structural 2SX-PES: Columns are favorable geometry. of an earthquake. C = 1 event. 

failure F1 = -3. If columns are subjected to extreme stresses F2 = -4. This Is the standard frequency used 

F = -5. they will tend to break rather than displace. Probability elsewhere where a passive design feature that Site characteristics 

of displacing so that the columns would come to rest relies only on configuration management is provide the likelihood of 

with their axes parallel is very low. used, when there are no other failure seismic activity and 

D1 = -2. Several days is the longest that the condition mechanisms, flood levels quoted. 

would be likely to persist before control of the site was D2 = -2. The presence of water would be 

reestablished, readily detected by responders following the 

I earthquake. 

* Likelihood index T is a sum. Uncontrolled: T=frqi or frql; Controlled: includes all indices T=a+b+c+d 

Note 1: For these sequences, the initiating event is failure of one of the controls, hence the frequency is assigned under that control.

A-29 NUREG-1718
Appendix A

The final results column of Table A-9 gives the risk index for each accident sequence 

that was identified in the ISA. The risk index will be used by staff to identify all risk 

significant sets of controls. These sets of controls will be reviewed with greater 

scrutiny than controls established to prevent or mitigate accident sequences of low 

risk. 

8. Accident Summary and Risk Index Assignment for Table A-9 

The definitions for the contents of each column in the accident summary tabulation, 

Table A-9, are provided below. 

(1) Accident Sequence 

This column is provided to list the accident sequences identified by the applicant in 

the ISA. It is important to the proper documentation of the ISA that the applicant 

subdivides the facility into a set of uniquely identified units, referred to here as 

"nodes." The applicant should give symbols, names, or numbers to these nodes 

that permit them to be uniquely identified. For example, the Plutonium Purification 

process described in Section A6 has the unique identifying symbol p 3. The 

specific node corresponding to second-pass solvent extraction has the unique 

identifier 2SX. Additional identifier characters have been added to form the 

identifier, 2SX-001, to identify the first accident sequence identified in that node. 

Because the applicant should list all the facility safety controls of significance used 

elsewhere in the ISA, tabulations of the unique node (and accident) identifiers can 

be used to find the accidents that these safety controls have been shown to 

prevent. By reviewing this table, the reviewer can then evaluate (1) the adequacy 

of the controls for preventing accidents and (2) the bases for making the 

consequence and likelihood assignments in the table. 

(2) Initiating Event or Control Failure 

This column is provided to list initiating events or control failures, typically 

identified in the process hazard analysis phase of the ISA, that may lead to 

consequences that do not comply with the performance requirements. Initiating 

events are of several distinct types: (1) external events such as hurricanes and 

earthquakes, (2) facility events external to the node being analyzed (e.g., fires, 

explosions, failures of other equipment, flooding from facility water sources), (3) 

deviations from normal of the process in the node (i.e., credible abnormal events), 

and (4) failures of safety controls of the node. The tabulated initiating events 

should consist only of those that involve an actual or threatened failure of safety 

controls, or that cause a demand requiring controls to function in order to prevent 

consequences that do not comply with the performance requirements. The 

frequency index number for initiating events is referred to in the table using the 

symbol "F." Table A-4 provides criteria for assigning a value to F. Usually, 

insufficient room is present in a tabular presentation like Table A-9 to describe 

accurately the events indicated. Consequently, the applicant should provide

NUREG-1718 A-30
Appendix A

supplementary narrative information to adequately describe each accident 

sequence of Table A-7. Cross-referencing between this information and the table 

should be adequate; for instance, the unique symbolic accident sequence 

identifiers can be used. Table A-7 is an example of a list of supplementary 

accident sequence descriptions corresponding to Table A-9. 

(3) Preventive Control 1 

This column is provided to list a control designed to prevent consequences that do 

not comply with the performance requirements. If separate controls are used to 

prevent different consequences, separate rows in the table should be defined 

corresponding to each type of consequence. Sequences in which two controls 

must simultaneously be in a failed state require assignment of three index 

numbers: the failure frequency of the first control, F1, the duration of this failure, 

D1, and the failure frequency of the second control, F2. For such sequences, the 

initiating event is failure of the first control. In these cases, F, is assigned using 

Table A-4. The failure duration of the first control is assigned using Table A-6. 

Other sequences may be more easily described as a failure of the safety controls 

on demand after the occurrence of an initiating event. In these cases, the failure 

probability index number, prfl, is assigned using Table A-5. The symbol "b" is 

used in the column heading for the indices associated with this control. 

(4) Preventive Control 2 

This column is provided in case a second preventive control exists. The failure 

frequency or failure probability on demand is assigned as for Preventive Control 1. 

The symbol "c" is used in the column heading for the indices associated with this 

control. 

In cases where no second preventive control exists (especially when the BDC 

requires double contingency), this column should contain a description of the 

consequences resulting from the first control failure. For example, there are 

generally two ways to demonstrate double contingency; either by i) specifying a 

second independent control that has to fail concurrently before criticality is 

credible, or ii) showing by calculation that the worst credible physical conditions 

resulting from the control failure remain subclinical. References identifying the 

consequence calculations that relate to the accident sequences should be 

included somewhere in the table, such as in column "c" or "e." 

(5) Likelihood Category 

This column is provided to list the likelihood category number for the risk matrix, 

which is based on the total likelihood index for a sequence. The total likelihood 

index, T, is the Sum of the indices for those events that comprise a sequence 

These events normally consist of the initiating event and failure of one or more 

controls, including any failure duration indices. However, accident sequences may

A-31 NUREG-1718
Appendix A

consist of varying numbers and types of undesired events. Methods for deciding 

what frequencies and failure durations need to be considered will be. described 

later in this appendix. Based on the sum of these indices, the likelihood category 

number for the risk matrix is assigned using Table A-3. The symbol "d" is used 

for this category number in the column heading. 

(6) Consequence Category 

This column is provided to assign the consequence category numbers based on 

estimating the consequences of all types (i.e., radiological, criticality, chemical, 

and environmental) that may occur. Based on this estimate, accidents can be 

assigned to the categories defined in 10 CFR 70.61. The symbol "e" is used for 

this category number in the column heading. 

(7) Risk Index 

This column is provided to list the risk index, which is calculated as the product of 

the likelihood category and consequence category numbers. This is shown in the 

column heading by the formula "f = d x e." Sequences with values of Tf" less than 

or equal to 4 are acceptable. The risk index can be calculated as the product of 

the consequence category with the failure index of the first preventive control, 

giving a measure of the unmitigated risk, in the case where the second control is 

not available to perform its function. This is a way to assess the risk significance 

of the second control. 

For sequences in which there is no second control specified, the unmitigated risk 

may be used to demonstrate an acceptable risk category. There may, however, 

be cases in which this is not possible; where there is a continuum of possible 

consequences resulting from occurrence of the accident sequence up to that 

point, credit may be taken for the unlikelihood of achieving an unacceptable 

physical state (e.g., probability that the upset exceeds a subclinical mass). This 

will require a thorough, documented justification for the reviewer to find this 

approach acceptable. 

(8) Comments and Recommendations 

This column is needed to record ISA team recommendations, especially when the 

existing system of controls is evaluated as being deficient. This may happen 

because a newly identified accident sequence is not addressed by existing 

controls, or because a deficiency has been found in the existing controls. 

9. Alternate Methods of Presentation (Fault Trees) 

Table A-9 displays the results of the ISA Summary in a tabular format by accident 

sequence. This approach is commonly developed from a What-If hazard evaluation 

technique, which is but one of several methods available. The methods that may be

NUREG-1718 A-32
Appendix A

used include What-If, HazOp, Failure Modes and Effects Analysis (FMEA), Fault 

Trees, and other methods. The What-If approach may be considered the least 

preferable of these approaches, particularly when there are a large number of accident 

sequences, as it is difficult to demonstrate completeness. That is, it will be difficult for 

the reviewer to verify that all credible accident sequences have been included in the 

hazard evaluation for a very complex process. 

There are additional reasons why a tabular format (Table A-9) may not be the best 

method of displaying the results of the ISA Summary for all processes. A variety of 

different techniques may be used rather than rigidly adhering to one format, if the 

multimethod approach enhances the clarity of the presented data. One of the main 

weaknesses of the tabular format is that it considers the accident sequence to consist 

of the failure of only two discrete controls. The establishment of double contingency 

may require more than two controls to ensure that at least two unlikely and concurrent 

upsets must occur before criticality is possible (and that the overall likelihood of 

criticality is highly unlikely). Several distinct controls may in general be combined into 

a single control system. When grouped as shown in Figure A-2, there may be several 

distinct controls that must be grouped together to ensure each "leg" of double 

contingency is unlikely to fail. This definition of unlikely is in the context of the double 

contingency principle, which numerically is approximately < 10-2, rather than the more 

restrictive value of _<4x104 as used in SRP Section 5.4.3.2(B)(ix). Use of this more 

conservative value would of course be acceptable, although it is highly doubtful that 

many operations would be able to meet this without the virtual elimination of 

administrative criticality controls. Although this information may be presented in the 

table by listing multiple controls in each bin (e.g., scenarios 2SX-005 and -008), it 

would be more efficacious to use a fault tree (Figure A-2). 

In addition, each accident sequence in the table considers the failure of the first control 

followed by failure of the second. Therefore there are actually two complementary 

accident sequences that must be considered in different rows of the table. This 

particular aspect of the logic-and the general logical flow of the accident as it 

unfolds-is masked by using an approach that follows a simple linear development of 

the sequence from initiating event to completion. In addition, the What-If approach 

often does not consider the control failure at a sufficiently high level. The answer to 

the question "What if the pump chamber leaks into the oil reservoir?" is often "The 

pump cannot leak because...." Considering the next-to-the- top-level event in the tree 

to be the loss of volume control ensures that the system will remain adequately 

subclinical even in the event that the pump failure occurs.

A-33 Draft NUREG-1718
Appendix A

Figure A-2: Fault Tree for SX Operation

I backfiow into deionized 

water (ONW) supply 

I 2SX-004 

sX00 2ollows pleen as fn, 2SX.003

NUREG-1718 A-34
Appendix A

Figure A-2: Fault Tree for SX Operation (cont.)

'iMdl-es not summed 

.o-n, common mode 

1.11_, potulated

A-35 NUREG-1 718
Appendix A

The advantages of using a fault tree to present the ISA Summary data include: (1) the 

top-down approach of a fault tree (as opposed to the bottom-up approach of What-If) 

ensures that all credible changes in process conditions-or loss of controlled 

parameters-:-are considered; (2) robustness is ensured by considering the control 

failures at a sufficiently high level; and (3) the logical sequence of events that must 

occur to cause a criticality cannot be described thoroughly using the tabular approach. 

Figure A-2 shows a fault tree for the accident sequences that are described in 

Table A-9; a cursory review demonstrates that these diagrams present a much higher 

level of information than is contained in Table A-9. For example, in order to have a 

criticality due to leakage through the heat exchanger into the steam supply, the 

following events would have occur: (1) the heat exchanger tubes would have to leak; 

and (2) the pressure on the tube side would have to drop below the pressure on the 

shell side; and (3) the leak would have to continue without being noticed, either by 

failure of the alarm to enunciate (mechanical) or failure of the operator to take 

appropriate actions (human error). This combination of events is then required to 

ensure that the overall consequence-getting concentrated uranium solution into the 

unfavorable geometry steam supply-is sufficiently unlikely (in fact, other controls are 

then recommended to reduce the likelihood index below -4). In addition, one can see 

that the loss of integrity of the evaporator tubes and loss of steam pressure are 

comparable events, and that reducing the frequency of mechanical failure of these 

items or the duration of alarm failure would result in the greatest drop in overall 

likelihood. In the event that other controls are credited in this scenario as a result of 

the recommendation, it would be difficult to convey the full amount of all the above 

information in the table. 

A7. Mixed Oxide Blending Operation 

Oxide blending is a process whereby dry U0 2 and PuO 2 powders are combined to produce a 

homogeneous blend suitable for fabrication into mixed oxide (MOX) fuel pellets and 

assemblies. The final mix consists of 20wt% PuO2 (isotopically, -96% 239Pu and -4wt% 2

'Pu) 

and 80wt% U(0.7wt%)0 2. Process equipment downstream of the blending operation is 

designed with subclinical dimensions for 30wt% PuO 2 MOX. The main criticality controls in the 

blending operation are mass, moderation, and plutonium "enrichment" (defined for the purpose 

of this example as the weight percent of PuO2 relative to the PuO2-UO 2 blend). 

A batch of U0 2 blendstock (-112 kg [246 Ib]) is measured out into a favorable geometry feed 

hopper attached to a safety-grade scale. The material in the hopper is weighed and sampled 

for moisture, after which it is gravity fed into the favorable geometry blending hopper. This is a 

conical-bottom hopper which gravity drains into the cylindrical homogenizer. The low feed rate 

of the blendstock and plutonium oxide ensures that the powder attains a high degree of 

homogeneity as the two oxides are blended together. Homogeneity is not credited for criticality 

safety, however, until after the material passes through the homogenizer. In addition to 

ensuring criticality safety, moisture control is important to ensure that the powder will flow 

smoothly to ensure proper transfer and mixing.

NUREG-1718 A-36
Appendix A

PuO2 powder is emptied from the 2-liter (0.45 gal) bottles into a plutonium oxide feed hopper 

through a hole in the bottom of a glovebox. This hopper is a 4-inch (10.2 cm) diameter 

cylindrical stainless steel vessel, which is heated to 150 °C (302 OF) to drive off residual 

moisture that may have accumulated. Several containers are emptied into the hopper until a 

mass of -28 kg (61.6 Ib) is reached. The powder is sampled and then gravity fed down a chute 

into the blending hopper. The flow rate of the plutonium oxide powder is controlled using a 

mass flow totalizer (MFT), which is interlocked to the plutonium feed valve; the feed rate is 

maintained at a slow rate using a stopcock on the input line. If the preset mass of plutonium 

oxide is exceeded, the MFT shuts the valve and prevents the overall plutonium enrichment in 

the blender from exceeding the safety limit of 22wt% PuO 2. After blending, the material is 

agitated for 30 minutes before being sampled; only after two independent samples confirm the 

correct enrichment may the material be transferred to the cylindrical homogenizer for further 

processing. Following this, the master mix is ball-milled and sieved to ensure proper 

consistency before being combined with additional U(0.7wt%)0 2, which results in a final mix of 

-4wt% Pu. 

Table A-1 0 presents the main accident sequences in the oxide blending operation. Table A-i 1 

shows the IROFS credited for double contingency during oxide blending. Table A-12 shows 

the main accident sequences and the preventive controls used. 

A table of accident sequences is used to communicate the ISA Summary information for the 

oxide blending operation; this operation is a simpler process from a criticality safety standpoint 

than the solvent extraction. Since the double contingency logic is based on a relatively simple 

set of controls on moderation, mass, and plutonium isotonic, this system is much more 

amenable to a tabular approach. Fault trees could be used profitably for this system, but there 

is a much lower level of complexity than in the first example, and tables may be adequately 

used. Several tables will in general be needed to summarize the information that must be 

presented; these should be cross-referenced to allow clear traceability of the control logic. The 

contents of the tables and figure for the oxide blending process are summarized below: 

1. Process Criticality Flow Diagram in Figure A-3 

This process is inherently much simpler than the solvent extraction example 

considered above, from the standpoint of criticality safety. Note that in this case, the 

entire operation is conducted in favorable geometry equipment, so there is no attempt 

to distinguish between favorable and unfavorable process steps graphically. One 

should note that labels have been attached to each piece of equipment relied on for 

safety, so that this diagram may be cross-referenced with the tables. Each component 

relied on for safety must be identified for incorporation into the configuration 

management program. For example, not only the mass flow totalizer, but also the 

interlock back to the PuO2 supply valve, and the valve itself, must be controlled to 

ensure that the active feature that prevents too high a plutonium enrichment in the 

blend hopper remains available and reliable to perform its function.

A-37 NUREG-1718
Appendix A

2. Accident Sequences in Table A-10 

By displaying the accident sequences in the manner shown, it is immediately apparent 

that the criticality controls on the process are mass, moderation, and plutonium 

isotonic. Each of the accident sequences describes the initiating event and presents 

such information as the controls that prevent the loss of that controlled parameter, the 

safety significance of the initiating event, the probable cause, and so forth. The 

information should be succinctly provided in the ISA Summary to immediately put the 

accident sequences into the proper viewpoint. 

3. Items Relied on for Double Contingency in Table A-i 1 

Figure A-3 shows how a criticality flow diagram may be used effectively to summarize 

the contents of Table A-1 1. Please see Section A6 for a fuller discussion of this type 

of table. 

4. Summary of Accident Scenarios and Risk Evaluation in Table A-12 

Note that Scenario MOB-001 has a consequence category of 0 (no consequences) 

instead of 3 (for criticality). This is not actually needed, because the likelihood index is 

sufficiently low based on the two preventive controls. However, this was done for 

illustrative purposes. As described in the first entry in Table A-10, the loss of mass 

control due to the failure of both of these preventive controls cannot lead to criticality 

without a concurrent loss of moderation control. This should be documented in 

criticality calculations that would be referenced in the table. This is an acceptable way 

to treat accident scenarios for which there is sufficient defense-in-depth that criticality 

cannot be achieved without the occurrence of additional events. In other words, the 

accident sequence defined by the failure of two preventive controls does not result in a 

criticality. 

Scenarios MOB-006a and -006b (and MOB-O1 Oa and -01 Ob) represent cases in 

which a single initiating event may occur due to two different causes. Generally 

deeper level events than the initiating event are not treated, but in this case it made 

sense to separate the sequences MOB-006 and -010 into more than one 

subsequence because different controls are needed for each pathway. Accident 

scenarios should be considered separate sequences if the controls relied on for safety 

are different, if the consequences are different (two scenarios leading to loss of mass 

control may result in different physical amounts and configurations), or likelihoods are 

different. Two accident sequences may have the same initiating events and the same 

consequences but different intermediate conditions or steps.

NUREG-1718 A-38
.1' 2 

I mK Z 19 I 

-n 

----------------------------------------------------------------- J 

............................. 

• • 0 

00l , A -- -

0 0 "

L ------- ------

m n 

OD 

* 1 ×I ,
Appendix A

Table A-10: Accident Sequence Descriptions

Accident Seuence_ Description 

Loss of Mass 

MOB-001 The initiating event is exceeding the mass limit of the blend tank by adding too 

much UO2 blendstock. This will have the effect of increasing the overall mass 

present, but will simultaneously decrease the plutonium enrichment. The overall 

effect of this is to increase the distance from the subclinical curve of mass as a 

function of plutonium enrichment as more blendstock is added. The system is 

adequately subclinical under conditions of double-batching uranium. To achieve 

criticality, this would have to be followed by a loss of moderation control. 

MOB-002 The initiating event is exceeding the mass limit of the blend tank, by adding too 

much PuO 2. This will have the effect of increasing both the overall mass and 

plutonium "enrichment." At -33 kg PuO2 (72.6 Ib) (and 23 wt%), the subclinical 

mass limit will be exceeded. Therefore, this could lead to criticality without any 

additional upsets and therefore dual controls are established on the plutonium 

mass. 

MOB-003 The initiating event is exceeding the mass limit of the blend tank by performing the 

blending operation while there is still blended oxide present from the previous 

batch in the tank. Assuming the previous batch was properly mixed, it would 

require an additional 50 kg (110 Ib) of PuO2+UO2 to exceed the subclinical mass 

limit. This could lead to criticality without any additional upsets and therefore dual 

controls are established to ensure the blend tank is empty of material before 

another batch is started. 

Loss of Moderation 

MOB-004 The initiating event is exceeding the moderation limit (1 wt% H20) by adding UO2 

that has not been properly sampled. This could lead to criticality without any 

additional failures. Dual independent sampling is required to ensure moisture 

limits are adhered to. Also, material will not freely flow through orifice if wet. 

MOB-005 The initiating event is exceeding the moderation limit (1 wt% H20) by adding PuO2 

that has not been properly sampled. This could lead to criticality without any 

additional failures. Dual independent sampling is required to ensure moisture 

limits are adhered to. Also, material will not freely flow through orifice if wet. In 

addition, both the plutonium feed hopper and the blend hopper are heated. 

Material is added at a sufficiently slow rate that contact with the heated blendstock 

will cause moisture in the plutonium to be driven off. 

MOB-006 The initiating event is exceeding the moderation limit (1 wt% H20) by introduction 

of liquid water from overhead waterlines or roof leaks. The blend tank is 

completely enclosed within an airtight and watertight enclosure. There are no 

overhead waterlines allowed. The most likely cause of this scenario is backflow of 

condensate from the ventilation header, which serves to remove evolved water 

from the heated material. The ventilation header is sloped and equipped with drain 

lines to ensure against condensate backflow. Even in the event of water intrusion, 

the heating is sufficient to drive off any realistic accumulation of liquid water.

NUREG-1718 A-40
Appendix A

NUREG-1 718

Accident Sequence Description 

Loss of Plutonium Enrichment 

MOB-07 The initiating event is exceeding the plutonium enrichment by adding too little 

blendstock to the blending hopper. This will have the effect of increasing 

plutonium enrichment while decreasing the overall mass. This will eventually reach 

criticality without any additional failures, but only when more than half the original 

UO2 blendstock is omitted. 

MOB-08 The initiating event is exceeding the plutonium enrichment by adding too much 

PuO2 feed to the blending hopper. This is identical to Scenario MOB-02 and will 

be discussed as a loss of mass control. 

MOB-09 The initiating event is exceeding the plutonium enrichment by adding PuO2 to the 

blending hopper without first adding blendstock. This is the bounding case of 

scenario MOB-007. Controls are established to ensure that blendstock is added 

and in the correct proportion before addition of PuO2 feed is allowed. 

MOB-1 0 The initiating event is exceeding the plutonium enrichment by the formation of 

clumps of higher enrichment PuO2 in the blending hopper. This can be caused by 

i) too high a plutonium feed rate, ii) failure of the magnetic stirrer, iii) failure of the 

deflection plate, or iv) failure of moderation control, resulting in a more cohesive 

mix. Calculations show there are sufficient controls such that homogeneity is not 

necessary to ensure subcriticality. However, criticality could occur if this were 

followed by a loss of moderation control.

A-41
Appendix A

Table A-11: Criticality Safety Limits and Controls

IROFS Parameters and Limits IROFS Description Management Measures QA Grade 

Identifier 1 1 _ 1 

MOB-ADM1 MASS: Procedures, training, and 1. Procedures and NA 

UO 2 feed - 112 kg (246 Ib) postings require that the mass training. 

be checked and certified by an 2. Operator/supervisor 

operator and supervisor prior to must sign material 

PuO 2 feed allowed, balance sheets. 

MOB-AE1 MASS: Safety grade scale attached to 1. Weekly calibration A 

U0 2 feed -112 kg (246 Ib) feed hopper. using mass 

standards. 

2. Tare weight re

certified whenever 

hopper is emptied. 

MOB-PE1 MODERATION: Blend tank comprises a welded 

Blend hopper is limited to stainless-steel barrier. 

lwt% H20. Blending required to be under 

dry nitrogen atmosphere. 

MOB-AE2 MODERATION: Blending required to be under a 1. Monthly functional B 

Blend hopper is limited to dry nitrogen atmosphere. test. 

lwt% H20. IROFS is a differential pressure 2. Configuration 

gauge interlocked to the feed control. 

supply valves and system 

alarm. 

MOB-AE3 MODERATION: Electric heater maintains 1. Weekly functional A 

Blend hopper is limited to powder at 150 'C (302 'F) in test. 

1 wt% H20. blend hopper. Low-T gauge 2. Configuration 

and alarm interlocked to supply control. 

valves. 

MOB-AE4 MASS: Mass flow totalizer (MFT) 1. Configuration A 

PuO2 feed -28 kg (61.6 Ib) interlocked to PuO2 supply control. 

valve. 2. Weekly function 

test. 

MOB-PE2 GEOMETRY: Diameter of oxide blender must Configuration control. C 

diameter < 12.7 cm (5") be less than 12.7 cm (5F). 

MOB-PE3 GEOMETRY: Diameter of PuO2 feed hopper Configuration control. C 

diameter < 10.2 cm (4") must be less than 10.2 cm (4"). 

MOB-ADM2 MODERATION: Procedures, postings, and 1. Procedures and NA 

Blend hopper is limited to 1 training require the material in training. 

wt% H20 the UO2 feed hopper to be 2. Lab QA procedures 

sampled for moisture before it must be followed. 

is released to the blending 

hopper. Supervisor 

concurrence required. Dual 

independent samples are 

required.

NUREG-1718 A-42
Appendix A

IROFS Parameters and Limits IROFS Description Management Measures QA Grade 

Identifier I I I I 

MOB-AE5 MODERATION: Electric heater maintains 1. Weekly functional A 

Blend hopper is limited to 1 powder at 150 °C (302 'F) in test. 

wt% H20 PuO2 feed hopper. Low-T 2. Configuration 

gauge and alarm interlocked to control. 

supply valves. 

MOB-ADM3 MODERATION: Procedures, postings, and 1. Procedures and NA 

Blend hopper is limited to 1 training require the material in training. 

wt% H20 the PuO 2 feed hopper to be 2. Lab QA procedures 

sampled for moisture before it must be followed. 

is released to the blending 

hopper. Supervisor 

concurrence required. Dual 

independent samples are 

required. 

MOB-ADM4 MASS: Only a limited number of 2-liter 1. Procedures and NA 

PuO, feed -28 kg (61.6 Ib) (0.5 gal) bottles may be training. 

emptied into the PuO2 feed 2. Material control 

hopper, such that the total program; the feed 

mass does not exceed 28 kg hopper is a process 

(61.6 Ib) as indicated on measurement node. 

material balance sheets. 

MOB-ADM5 MASS: Operators are required to Procedures and NA 

Total blend hopper mass < check visually that the blend training. 

140 kg (308 Ib) hopper is devoid of more than 

surface contamination after 

each campaign. 

MOB-ADM6 MASS: Blend hopper must be NDA Procedures and NA 

Total blend hopper mass < scanned after each campaign. training. 

140 kg (308 Ib) 

MOB-PE4 MODERATION: Stopcock on PuO 2 feed line 1. Flow rate checked B 

Blend hopper is limited to controls feed rate to 800 g/hr during run by 

lwt%/. H20. (1.8 lb/hr). This slow flow rate monitoring MFT. 

ensures that any moisture will 2. Configuration 

be driven off on contact with control. 

the heated blendstock. 

VEN-PE13 MODERATION: Ventilation header must be Configuration control. C 

Blend hopper is limited to sloped away from the blend 

lwt% H20. hopper. 

VEN-PE15 MODERATION: Ventilation header must be 1. Configuration B 

Blend hopper is limited to equipped with condensate control. 

lwt% H20. drains at its lowest point to 2. Periodic monitoring. 

prevent condensate backflow to 

the blend hopper. 

BLD16-65 MODERATION: No overhead waterlines are Configuration control. C 

Blend hopper is limited to allowed in Building 16. 

lwt% H20.

A-43 NUREG-1718
Appendix A

NUREG-1 718

IROFS Parameters and Limits IROFS Description Management Measures QA Grade 

identifier I ý 

MOB-ADM7 ISOTOPIC: Supervisor must check that 1. Procedures and NA 

PuO2 content 20wt% UO2 feed hopper is empty and training. 

that the appropriate mass has 2. Material control 

been added before PuO2 program; the 

transfer is authorized. blendstock feed 

hopper is a process 

measurement node.
Appendix A

Table A-12: Accident Sequence Summary.and Risk Index Assignment

Accident Initiating Event Preventive Preventive Likelihood' Conse- Risk Comments 

Sequence (a) Control Control 2 C Index T and quence Indices & 

(b) (c) Category C Category (g=d x e) Recommendations 

(d) (e) 

MOB-001 Too much blendstock MOB-ADM1: Blendstock mass certified before MOB-ADM7: Supervisor must ensure that the T = -7 0 0 Criticality not possible without 

added. Introduction of PuO2 , appropriate mass was emptied from the an additional failure. 

F1 = -3. Material control sensitivity ensures this blendstock feed hopper. C = 1 

receives appropriate attention and supervisor F2 = -2. Required on checklist and reinforced by 

oversight, training and postings. 

Dl = -2. Process is a batch process, campaign is D2 = -2. See MOB-001, Control 1. 

running several days. Failure would be detected at 

start of subsequent campaign. 

MOB-002 Too much PuO2 added. MOB-ADM4: No more than 28kg (61.6 Ib) PuO2 MOB-AE4: MFT limits total Integrated PuO2 , T =8 3 3 

may be charged Into the feed hopper, which is transferred to the blend hopper. 

F1 = -3. Material control sensitivity ensures this F2 = -3. Regular maintenance and testing C = 1 

receives appropriate attention and supervisor ensures reliability. 

oversight. D2 = -2. Functionally tested weekly. 

D1 = -2. See MOB-001, Control 1. 

MOB-003 MOX not cleaned out MOB-ADM5: Visual check that blend hopper is MOB-ADM6: Blend hopper must be NDA T = -5 3 3 

before next batch started, empty before each campaign. scanned before each campaign. 

F1 = -2. Required on checklist and reinforced by F2 = -2. Required on checklist and reinforced by C = 1 

training and postings, training and postings, 

Dl = -2. See MOB-001, Control 1. D2 = -2. See MOB-001, Control 1. 

MOB-004 Moderated blendstock MOB-ADM2: Dual independent samples taken to MOB-AE3: Electric heater maintains T = -8 3 3 

added. confirm moisture level of blendstock. temperature sufficient to drive off moisture in the 

F1 = -3. Requires failure of two operators to follow blend hopper. C = 1 

procedures, and independence of sampling and lab F2 =-4. Past history with this model of heater 

analysis ensures reliability, shows it to be very reliable. 

D1 = -2. See MOB-00t, Control 1. D2 = -2. Functionally tested weekly, 

MOB-005 Moderated PuO2 added. MOB-AE5: Electric heater maintains temperature MOB-AE3: Electric heater maintains T= -10 3 3 

sufficient to drive off moisture in feed hopper, temperature sufficient to drive off moisture In the 

F1 = -4. Past history with this model of heater blend hopper, C = 1 

shows it to be very reliable. F2 = .4. Past history with this model of heater 

Dt = -2. Functionally tested weekly. shows it to be very reliable. 

D2 = -2. Functionally tested weekly.

A-45 NUREG-1718
Appendix A 

Accident Initiating Event Preventive Preventive Likelihood' Conse- Risk Comments 

Sequence (a) Control I Control 2 Index T and quence Indices & 

(b) (c) Category C Category (g=d x e) Recommendations 

(d) (e) 

MOB-006a Water backflow from VEN-PE13: Ventilation header sloped away from VEN-PE15: Ventilation header has condensate T = -6 3 3 

ventilation condensate. blend hopper. drains to prevent backflow. 

F1 = -3. The configuration control program requires F2 = -3. The configuration control program C =1 

Installation according to design drawings and requires installation according to design drawings 

prestartup verification. Several layers of and pre-startup verification. Several layers of 

management controls would have to fail to allow this management controls would have to fail to allow 

to happen, this to happen. 

D1 = 0. Would be checked during annual audit. D2 = 0. Would be checked during annual audit. 

MOB-006b Water intrusion" from MOB-PEI: Blend tank is watertight, MOB-AE2: Differential pressure gauge with T = - 3 3 

external source. Ff = -3. The ability of certified welders to ensure the Interlock prevents introduction of feed if 

integrity of welded vessels has been demonstrated. containment is breached. C I 

D1 = 0. Would be checked during annual audit. F2 = -2. Based on past failure rate data when 

used in combination with high efficiency 

particulate air (HEPA) filters. 

D2 = -1. Though sufficient to detect breach of 

the containment immediately (D--5), its failure 

would be detected during monthly functional test. 

High demonstrated reliability means that 

D = -5 Is actually more realistic. 

MOB-007 Too little blendstock added, MOB-ADMI: Blendstock mass certified before MOB-ADM7: Supervisor must ensure that the T = -7 3 3 

introduction of PuO2 , appropriate mass was emptied from blendstock 

F1 = -3. Material control sensitivity ensures this feed hopper. C 1 

receives appropriate attention and supervisor F2 = -2. Required on checklist and reinforced by 

oversight, training and postings. 

D1 = -2. See MOB-00, Control 1. D2 = -2. See MOB-001, Control 1. 

MOB-008 same as MOB-002 (q.v.) 

MOB-009 PuO2 added before MOB-ADMI: Blendstock mass certified before MOB-ADM7: Supervisor must ensure that the T -7 3 3 

blendstock. introduction of PuO. appropriate mass was emptied from blendstock 

F1 = -3. Material control sensitivity ensures this feed hopper. C = 1 

receives appropriate attention and supervisor F2 =-2. Required on checklist and reinforced by 

oversight, training and postings, 

D1 = -2. See MOB-00t, Control 1. D2 = -2. See MOB-001, Control l.n

NUREG-1718 A-46
Appendix A

Accident Initiating Event Preventive Preventive Likelihood* Conse- Risk Comments 

Sequence (a) Control I Control 2 Index T and quence Indices & 

(b) (c) 'Category C Category (g=d x e) Recommendations 

(d) (e) 

MOB-010a PuO, clump develops by: MOB-PE4: Feed rate controlled by stopcock. T =-6 3 3 

feed rate too high F1 = -3. This is locked into place and tested before 

start-up. Has no moving or wear parts. C= 1 

D1 = -3. Failure would be detected during the 

course of one shift. Process is continually 

monitored by operators. 

MOB-010b PuO 2 clump develops by: MOB-AE5: Electric heater maintains temperature MOB-PE4: Feed rate controlled by stopcock. T = -9 3 3 

failure of moderation control sufficient to drive off moisture in feed hopper. F2 = -3. This is locked into place and tested 

F1 = -4. Past history with this model of heater before start-up. Has no moving or wear pars. C = 1 

shows it to be very reliable. D2 = -3. Failure would be detected during the 

Dl = -2. Functionally tested weekly. course of one shift. Process Is continually 

Smonitored by operators 

*Likelihood index T is a sum. Uncontrolled: T=frqi or frql; Controlled: includes all indices T=a+b+c+d 

Note 1: For these sequences the initiating event is failure of one of the controls, hence the frequency is assigned under that control.

A-47 NUREG-1718
APPENDIX B 

NATURAL PHENOMENA/OTHER EXTERNAL EVENTS 

B1. General Discussion 

Natural phenomena events (i.e., earthquakes, high winds, tornadoes, tornado missiles, floods) 

and other external events (e.g., transportation accidents, airplane crashes, industrial accidents, 

and fires external to the facility) should be addressed in the integrated safety analysis (ISA) as 

initiating events. 

Currently, no regulatory guides (RGs) in Division 3, Fuels and Materials Facilities, address 

natural phenomena events and other external events. Therefore, sections of NUREG-0800, 

Standard Review Plan (SRP), and Division 1 RGs for powereactors describe methods for 

performing evaluations for natural phenomena events and other external events and provide 

useful reference information (See B.2 References). 

The applicant's approach for evaluating natural phenomena events and other external events 

should be in concert with the risk-informed approach described in 10 CFR Part 70, Subpart H. 

Although the references provide useful information for staff use in the review, if consulted, 

some of the analysis methods described therein would need to be adapted to be risk-informed 

and to agree with the approach described in 10 CFR Part 70, Subpart H. 

The applicant's risk from natural phenomena events and other external events shall meet the 

performance requirements described in 10 CFR Part 70, Subpart H. For each external event, 

the goal is to identify the largest magnitude event (i.e., lowest probability of occurrence) for 

which the consequences in 10 CFR 70.61 are met. Once this event is identified, an 

assessment should be performed to determine whether the performance requirements in 

10 CFR 70.61 are met. The applicant's evaluation to determine whether the performance 

requirements are met should be iterative. First, the applicant should perform evaluations to 

describe the likelihoods associated with a suite of magnitudes for each type of natural 

phenomenon or other external event. For example, when assessing earthquakes, the applicant 

should describe likelihoods associated with a suite of maximum accelerations ("g" values); 

when assessing tornadoes or high winds, the applicant should describe likelihoods associated 

with a suite of maximum windspeeds; when assessing floods, the applicant should describe 

likelihoods associated with a suite of maximum water levels and velocities. 

Next, the applicant should select a likelihood for each external event and identify the associated 

magnitude (e.g., water level, windspeed, acceleration level). For each external event, the 

applicant should identify failures of items relied on for safety associated with the magnitude of 

the event, taking into consideration common-cause failures and the likelihoods of the failures, 

given the event. This step involves developing and applying intermediate assessment tools 

such as response spectra and floor response spectra for seismic analysis. The applicant 

should determine the consequences, in terms of radiation and chemical exposures to the public 

and workers and any nuclear criticalities, for each external event. The applicant should 

compare the consequences and the associated likelihoods to the performance requirements in 

10 CFR Part 70, Subpart H. If the likelihood and consequences of the external event

B-1 NUREG-1718
Appendix B

satisfy the performance requirements, then the process should be repeated using a less likely 

event in order to identify the lowest likelihood event (i.e., largest magnitude event) for which the 

consequences in 10 CFR 70.61 are met. At this point, it should be determined whether the 

performance requirements are met. If the performance requirements are met, then the value 

(i.e., acceleration value; windspeed; water level) of the external event magnitude is acceptable. 

Otherwise, if the performance requirements are not satisfied, modifications should be identified 

and the process repeated until the performance requirements are satisfied. This process 

should be performed for each natural phenomenon event and for other external events. 

For example, first a preliminary (or final) design is developed. Then, assume that the 

magnitude associated with a 1 E-3 event is selected as a starting point and applied to the facility 

design. The failures of structures, systems, and components and the resulting consequences 

are determined. If the performance requirements in 10 CFR 70.61 are not met, then the facility 

should be designed for a larger magnitude event and the process repeated. If the performance 

requirements are met, this does not mean, however, that the process is completed since the 

performance requirements may still not be met for a higher magnitude event (i.e., less likely 

event). In this case, no redesign is necessary (since the performance requirements were met) 

but a larger magnitude event (i.e., less likely event) should be selected and the process 

repeated in order to identify the largest magnitude event for which the consequences in 

10 CFR 70.61 are met. At this point, it should be determined whether the performance 

requirements in 10 CFR 70.61 are met. If the performance requirements are met, then the 

value (i.e., acceleration value; windspeed; water level) of the external event magnitude is 

acceptable. Otherwise, if the performance requirements are not satisfied, modifications should 

be identified and the process repeated until the performance requirements are satisfied. 

B2. References 

1. Floods 

Nuclear Regulatory Commission (U.S.) (NRC). NUREG-0800, Standard Review Plan, 

Section 2.4.1, "Hydrologic Description." NRC: Washington, D.C. 

NUREG-0800, Standard Review Plan, Section 2.4.2, "Floods." NRC: 

Washington, D.C. 

NUREG-0800, Standard Review Plan, Section 2.4.3, "Probable Maximum Flood 

(PMF) on Streams and Rivers." NRC: Washington, D.C. 

NUREG-0800, Standard Review Plan, Section 3.4.1, "Flood Protection." NRC: 

Washington, D.C. 

Regulatory Guide 1.59, "Design Basis Floods for Nuclear Power Plants." NRC: 

Washington, D.C.

NUREG-1718
Appendix B 

Regulatory Guide 1.102, "Flood Protection Plan for Nuclear Power Plants." 

NRC: Washington, D.C. 

2. Wind and Tornadoes 

NUREG-0800, Standard Review Plan, Section 3.3.1, "Wind Loadings." NRC: 

Washington, D.C. 

. NUREG-0800, Standard Review Plan, Section 3.3.2, "Tornado Loadings." 

NRC: Washington, D.C. 

. Regulatory Guide 1.76, "Design Basis Tornado for Nuclear Power Plants." 

NRC: Washington, D.C. 

Regulatory Guide 1.117, "Tornado Design Classification." NRC: Washington, 

D.C. 

3. Earthquakes 

NUREG-0800, Standard Review Plan, Section 3.7.1, "Seismic Design 

Parameters." NRC: Washington, D.C. 

NUREG-0800, Standard Review Plan, Section 3.7.2, "Seismic System 

Analysis." NRC: Washington, D.C. 

NUREG-0800, Standard Review Plan, Section 3.7.3, "Seismic Subsystem 

Analysis." NRC: Washington, D.C. 

Regulatory Guide 1.60, "Design Basis Response Spectra for Seismic Design of 

Nuclear Power Plants." NRC: Washington, D.C. 

Regulatory Guide 1.92, "Combining Modal Responses and Spatial Components 

in Seismic Response Analysis." NRC: Washington, D.C. 

Regulatory Guide 1.122, "Development of Floor Design Response Spectra for 

Seismic Design of Floor Supported Equipment and Components." NRC: Washington, 

D.C. 

Regulatory Guide 1.161, "Damping Values for Seismic Design of Nuclear Power 

Plants." NRC: Washington, D.C. 

Regulatory Guide 1.165, "Identification and Characterization of Seismic Sources 

and Determination of Safe Shutdown Earthquake Ground Motion." NRC: Washington, 

D.C.

B-3 NUREG-1718
Appendix B

4. Other External Events 

NUREG-0800, Standard Review Plan, Section 2.2.1-2.2.2, "Identification of 

Potential Hazards in Site Vicinity." NRC: Washington, D.C. 

. NUREG-0800, Standard Review Plan, Section 3.5.1.6, "Aircraft Hazards." 

NRC: Washington, D.C. 

Regulatory Guide 1.91, "Evaluation of Explosions Postulated to Occur on 

Transportation Routes Near Nuclear Power Plants." NRC: Washington, D.C.

NUREG-1 718 B-4
APPENDIX C 

NUCLEAR CRITICALITY SAFETY EXAMPLES 

Nuclear criticality safety (NCS) has become one of the disciplines of greatest safety focus by 

the NRC and the nuclear fuel industry. Many of the safety and regulatory issues that arise at 

the various fuel cycle facilities involve NCS, which has consequently received a considerable 

amount of interest and management attention. This appendix therefore contains examples of 

how the acceptance criteria in the NCS Chapter (Chapter 6.0) of this SRP may be applied at a 

hypothetical mixed oxide (MOX) fuel fabrication facility. 

The examples should not be treated as acceptance criteria by the reviewer or the applicant, but 

provide additional guidance to both the reviewer and the applicant in how the criteria could be 

applied in the context of an actual MOX fuel fabrication facility. The purpose of this appendix is 

to provide examples that illustrate the general and abstract concepts in Chapter 6.0 concretely 

and elucidate the acceptance criteria by showing how they may be applied in a real MOX plant. 

This is based on an imperfect knowledge of the principles of operation of a MOX fuel fabrication 

plant, such as that described in Appendix A, "Example Procedure for Risk Evaluation." 

Thus the examples presented merely give a hypothetical instance of how a given acceptance 

criterion may be applied on a particular process. The details of how the acceptance criterion is 

met may be very different in the actual design or in another process. 

C1. SRP Section 6.4.3.3.1(G) states: 

The applicant should commit to consider the variability and uncertainty in a process and 

the NCS subcritical limit when setting NCS safety limits. If a controlled parameter is 

dependent on other physical parameters (such as mass controlled indirectly by 

concentration and volume), then the uncertainties in each of the independent 

parameters should be folded into the estimated uncertainty of the dependent parameter. 

Suppose the applicant has an unfavorable geometry waste tank, relying on maintaining 

less than a minimum critical mass for NCS. If the mass is controlled indirectly on the 

basis of measurements of concentration of material transferred to the tank, then the 

uncertainties in the volume of the tank, the identity of the solutions transferred, and the 

precision on any in-line monitors or sampling methods (uncertainty in the concentration 

measurement) would be taken into consideration in setting the mass limit. The mass 

safety limit would be set sufficiently low so that subcriticality would be ensured even 

under a worst-case combination of these uncertainties. 

C2. SRP Section 6.4.3.3.2(B) states: 

Based on the performance requirements in 10 CFR 70.61, the applicant should commit 

to the policy that: "No single credible event or failure could result in a criticality 

accident." This commitment should involve an evaluation to identify common-mode 

failures that may simultaneously defeat two or more controls.

C-1 NUREG-1718
Suppose the applicant has a metering pump that controls the flow of plutonium nitrate 

to a uranyl nitrate-plutonium nitrate blending tank. This pump may be credited with 

controlling both concentration in the tank and the "enrichment" (plutonium isotopics) in 

the tank. The applicant would evaluate the accident scenario resulting from failure of 

the pump and this evaluation would include a determination of possible common-mode 

failures due to simultaneous loss of both concentration and enrichment control. The 

failure could constitute a single contingency that could defeat both controls. If any such 

common-mode failure is identified, then additional controls besides concentration and 

enrichment are needed to meet the double contingency principle. 

C3. SRP Section 6.4.3.3.2.1 contains requirements for the use of mass control for NCS. 

Application of criteria (A) and (B) are illustrated below: 

Suppose the applicant has a filter press, and there is a safety limit to remove and handle 

filtercake on a safe batch basis. If the mass limit derived from criticality calculations is 

based on assumptions about the weight percentage of plutonium in the filtercake, then 

the applicant would first measure the proportion of plutonium before the filtercake could 

be removed and packaged. This would be done using calibrated instrumentation, such 

as non-destructive assay (NDA) equipment, through dual independent sampling, or by 

other direct measurement techniques. Reliance could not be placed on historical data 

or the process design in lieu of direct mass measurement, because there is always the 

potential for process upsets. These controls would be unnecessary if the filtercake were 

assumed to be 100% plutonium when deriving the mass limits. The equipment used to 

perform the measurements would be considered safety-related and would be placed in 

the plant's quality assurance program for regular testing and calibration. 

C4. SRP Section 6.4.3.3.2.2 contains requirements for the use of geometry control for NCS. 

Application of criteria (A), (C), and (D) are illustrated below: 

Suppose the applicant establishes a geometry limit on a tray of fuel rods by limiting the 

height of the rods to less than a safe depth. The applicant could use the minimum safe 

slab thickness in setting the safety limit, or could explicitly model the array of rods and 

demonstrate that a certain depth would be subcritical. If the applicant bases its limits on 

computer calculations of the actual configuration, the diameter and any other credited 

attributes of the rods should be conservatively taken into account. Field measurements 

should be performed and the geometrical tolerance should be added to the measured 

dimensions when comparing to the modeled value. If the basis of the slab thickness is 

handbook data, the safety limit should be no more than some specified percentage (say, 

85%) of the minimum critical slab thickness. The depth of the restraining device should 

be designed so as to take any mounding of the fuel rods into account. When all these 

variations are added cumulatively the maximum depth should still be less than the 

established safety limit.2 

2When geometry control is used in conjunction with other controls-such as dimensional 

limits specific to a particular uranium compound or when certain poisons are present-then this 

hybrid control should be considered one control. One could not take credit for a hybrid control 

(such as geometry-neutron poison) as one leg of double contingency, and then credit neutron 

poison as the second leg. An additional independent control is needed to meet the double

NUREG-1718 C-2
C5. SRP Section 6.4.3.3.2.3 contains requirements for the use of density control for NCS. 

Application of criterion (A) is illustrated below: 

Suppose the applicant takes credit for an oxide density less than maximum theoretical 

density in the criticality calculations. Then process variables that could affect the 

density of pellets removed from the oxide sintering furnace may include: the length of 

residence time in the furnace, the temperature of the furnace, the atmosphere in the 

furnace, the additives added to the fuel, the force applied by the pellet press, and the 

reduction in effective density due to geometrical packing of pellets in trays. The 

applicant would be expected to provide assurance that there were sufficient controls on 

process conditions to ensure that the density credited in the evaluation would not be 

exceeded. 

C6. SRP Section 6.4.3.3.2.7 contains requirements for the use of concentration control for 

NCS. Application of criteria (E) and (F) are illustrated below: 

Suppose the applicant credits concentration control in a large geometry plutonium 

nitrate tank containing more than a minimum critical mass. Any process variables which 

are needed to maintain the plutonium in solution form should be specifically identified as 

NCS controls. These may include: acid molarity (excess nitric acid), electric heaters, or 

an air sparger or mechanical stirrer. Also, if precipitating agents are used in the facility, 

then the tank lid should be locked to prevent inadvertent manual introduction, and there 

should be no hard-piped connections of such materials to the tanks. Concentration 

should also be sampled periodically to ensure against accumulation of an unsafe mass. 

C7. SRP Section 6.4.3.3.2.9 contains requirements for the use of neutron absorber control 

for NCS. Application of criterion (B) is illustrated below: 

Suppose the applicant takes credit for the percentage of gadolinium in the oxide pellets 

in setting mass and/or geometry limits. Then the concentration of 64Gd in the fuel would 

be measured and all process variables that could affect this parameter would be 

controlled. If the presence of fixed poison rods are credited for fuel assemblies, then the 

poison loading and dimensions of these rods would be declared as IROFS. If they are 

subjected to an environment (such as an acid pickling operation) where the poison could 

be degraded over time, then they should be placed on a periodic surveillance program 

to monitor its continued effectiveness. 

C8. SRP Section 6.4.3.3.2.11 contains requirements for the use of heterogeneity control for 

NCS. Application of criteria (A), (B), and (C) are illustrated below: 

Suppose the applicant takes credit for maintaining a low-enriched uranium solution in 

homogeneous form in a uranium oxide dissolution tank. Process features that ensure 

the uranium remains in solution and not in suspension, such as a motorized stirrer or 

heater, acid molarity, and post-dissolution filters, should be declared IROFS. Then the 

contingency principle. For true geometry control, the system must still be subcritical when all 

other process parameters are at their worst credible conditions.

C-3 NUREG-1718
system should be evaluated over the entire range of particulate sizes, if heterogeneity is 

an important effect, to determine the optimal configuration. 

C9. SRP Section 6.4.3.3.4(B)(i) discusses the setting of minimum subcritical margins for 

normal and credible abnormal conditions. This paragraph states, in part, "A condition 

that occurs on a regular basis during facility operations would not be considered 

abnormal." 

Example of conditions that would be considered abnormal include mechanical failure or 

the defeat of rigorous administrative controls. Spills outside the tube of a favorable 

geometry sintering furnace due to mechanical breakdown would probably be considered 

abnormal. A spill of SNM when open containers of powder are manually handled would 

not be considered an abnormal occurrence. However, there may be engineered failures 

that could not be classified as abnormal occurrences. An example would be failure of 

evaporator tubing when operational data shows that this occurs on a regular basis 

(several failures per year). 

SRP Section 6.4.3.3.4(B)(i) also states, in part, when discussing crediting abnormal 

events in setting minimum subcritical margins: "the increased risk associated with the 

less conservative margin should be commensurate with and offset by the unlikelihood of 

achieving the condition to begin with." 

As an example of this, the applicant may choose to use a minimum subcritical margin of 

0.05 (with no bias, kif + 2ur _• 0.95), without further justification than having an validation 

study showing there is no measurable bias. However, the applicant may propose using 

a ke, limit of 0.95 for normal conditions and 0.97 for abnormal conditions (without bias). 

The applicant would have to justify that the risk associated with reducing the margin by 

0. 02 is offset by the restrictions imposed on what events may be considered abnormal. 

Suppose the square root of the pooled variance of the benchmark experiments is 0.02. 

A keff of 0.97 represents 1.5g and 0.95 represents 2.5g below the mean kf, = 1.0. Using 

the assumption that the benchmarks are normally distributed, this permits a quantitative 

evaluation of the increase in risk due to the reduction in margin. Only 6. 7x 1c7 2 of critical 

systems will result in a kff more than 1.5d below the mean; only 6x 101 will result in a kf, 

more than 2.5a below the mean. Therefore, increasing the upper safety limit from 0.95 

to 0.97 would result in approximately ten times the risk that a system calculated safe at 

this limit is actually critical. Normal conditions are expected to occur with a frequency 

on the order of - 1/yr. Thus, to compensate for the increase in risk due to the degraded 

margin, abnormal conditions should have a frequency _• 0.1/yr if the upper limit of 0.97 

is to be used. 

O10. SRP Section 6.4.3.3.4(E) addresses the determination of the code's area of applicability 

(AOA). 

As an example of what would be included in the AOA, it would include a range of the 

chemical and physical forms of materials in the benchmark experiments, isotopic 

concentrations, moderation range, other materials and non-fissionable isotopes present, 

neutron energy ranges, and any codes options and statistics used. For example, an

NUREG-1718 C--4
AOA could be defined as: thermal mixed oxide cases with 4-10 wt% plutonium, H/Pu > 

500, fuel-pin lattices in water, gadolinium and zirconium poisons, etc. The benchmarks 

could have been run with 500 generations and 1000 neutrons per generation, using 24

inch water albedos, no code biasing, and 27-group neutron cross sections. All of this 

information would be specified in the definition of the AGA. 

C11. SRP Section 6.4.3.3.5 describes the requirement to comply with the double contingency 

principle and the need to evaluate common-mode failures. 

Suppose a process depends on both geometry and moderation control, or on two 

independent controls on moderation (such as the integrity of the roof and enclosed 

gloveboxes). As an example of possible common cause effects, a major fire in the 

facility may have the potential to defeat both controls on moderation and geometry by 

sprinkler activation and by bringing material together into a more reactive configuration. 

Or dual moderation controls could be defeated by sprinkler activation and burning 

through of material gloveboxes. Such common-mode accident scenarios would be 

considered in the ISA and in the criticality safety evaluations. 

As an example of possible justification for exceptions to the double contingency 

principle, suppose the applicant has a process requiring the processing of large 

quantities of uranium or plutonium oxide. It may not be economically feasible to limit 

throughput to less than several tons of oxide. In this instance, the only practical means 

of preventing criticality may be moderation control Multiple controls on moderation in a 

large geometry hopper or fluidized bed reactor may be sufficient to ensure double 

contingency for most cases. However, there may be a single almost incredible 

event-such as a catastrophic breach in all layers of containment due to an explosion

that could lead to criticality. This would not be double contingency since a single event 

that could lead to criticality exists. However, single contingency may be authorized 

based on the extreme unlikelihood of such an event. Another example may be a system 

that relies only on an unusually rigorous passive geometry control, where there is no 

identifiable accident scenario that could credibly lead to criticality. 

C12. SRP Section 6.4.3.3.7 discusses facility change requirements. Examples of changes 

that would or would not require prior NRC approval are presented below: 

Suppose the applicant has a uranium solution pump with certain controlled attributes, 

such as volume and oil inventory. The attributes important to criticality safety are the 

plenum volume (pump is safe volume) and diameter, volume of the oil reservoir (safe 

volume and limited moderator into process), and pump capacity (required to prevent 

overflowing downstream equipment). A change that would not require prior NRC 

approval would be changing the manufacturer or model number of the pump, provided 

this did not affect any of the above attributes. Changing from a centrifugal to a positive 

displacement pump may not require prior approval, unless this introduces additional 

accident sequences (new failure modes) or reduced the reliability of the pump or any of 

the pump's characteristics important to NCS. 

Replacement of a backflow preventer with a check valve would typically require prior 

approval, because although this did not change any controlled parameters or introduce

C-5 NUREG-1718
any new accident sequences, it reduced the margin of safety by replacing a robust 

control with one which is more likely to fail. 

C13. SRP Section 6.4.3.3.8 discusses criteria for reportable criticality safety events. 

Suppose the applicant has an unfavorable geometry waste water tank that relies solely 

on mass control for criticality safety. An unauthorized transfer to this tank would be 

reportable as a one-hour report, even if it were unknown whether an unsafe mass had 

been transferred (since mass is what is maintaining subcriticality). Even if it were later 

determined that an unsafe mass did not accumulate in the tank, this report should not 

be retracted. In this case, the mass control was lost even if the mass did not physically 

exceed a safe value, because the multiplicity of positive controls needed for double 

contingency was not maintained - thus, it would be reportable under Items C and D 

above. 

As a second example, consider the case of a slab of molybdenum boats containing 

green MOX pellets, which are heat-treated in a sintering furnace. Typical criticality 

controls for this example would include the mass in each boat, the depth of pellets in the 

boat, and moderation. The depth of pellets is controlled (in this hypothetical example) 

by the boat's dimensions, although the boats are demonstrated to be adequately 

subcritical when filled up to the top with pellets at theoretical density. An applicant 

controlling mass in these boats may establish an operating limit lower than the actual 

capacity of the boats in order to ensure that the material is processed uniformly and to 

give the operators a certain amount of margin in filling the boats. The formal subcritical 

limit established in criticality safety evaluations typically exceeds this operating limit by a 

substantial amount. When controls are established with such conservatism, it may take 

several events before criticality is possible, including exceeding the analyzed safe slab 

depth of the boat and adding moderator. Because several events are needed for 

criticality even after an upset occurs, merely exceeding the operating limit would not be 

reportable as an immediate report under Appendix A Paragraph (a)(5). Double 

contingency would not be lost in this hypothetical case. The filling of a boat to two 

grams more than that allowed in the operating limit would not be considered significant 

(since the boats had been shown adequately subcritical even when overfilled well 

beyond this) and would not require reporting. This of course requires a significance 

determination as to what constitutes a significant loss of mass control in the criticality 

safety evaluation. 

However, the filling of a boat until the material mounded over the top would violate both 

geometry and mass control and would be a significant loss of the control, since the 

mass would exceed the pre-analyzed condition. The resultant condition, and that of 

exceeding the subcritical limit, would constitute a twenty-four hour reportable events 

since the IROFS failed to meet the performance requirements of 10 CFR 70.61(e). 

Overflowing the material out of the boat and onto the floor, however, if it were not 

analyzed and shown to be subcritical as an upset condition in the criticality safety 

evaluation, would be reportable as an unanalyzed condition. 

The information to be submitted in these reports would include, to the extent known at 

the time of the event, the quantities and isotopics of the materials involved, their 

moderation levels, and any other pertinent information needed to assess their k., the

NUREG-1 718 C-6
particular procedural failure that led to the event, and the condition of the remaining 

geometry and moderation controls to allow NRC to determine the actual and potential 

significance of the event.

C-7 NUREG-1 718
APPENDIX D 

FIRE HAZARDS ANALYSIS PROCEDURES 

D1. The purpose of the fire hazards analysis (FHA) is to document specific fire hazards, fire 

protection features proposed to control those hazards, and the overall adequacy of facility 

fire safety. The FHA consists of a systematic analysis of the fire hazards, an identification 

of specific areas and systems important to facility fire safety, the development of design

basis fire scenarios, an evaluation of anticipated consequences, and a determination of 

the adequacy of facility fire safety. 

D2. A preliminary FHA should be performed for the mixed oxide (MOX) facility early in the 

design phase to ensure incorporation of an acceptable level of protection in the evolving 

design. 

D3. The FHA should be performed under the direction of a qualified fire protection engineer, 

with support from chemical, electrical, mechanical, and systems engineers, as well as 

operations staff as needed. 

D4. The FHA should contain, but not be limited to, a conservative assessment of the following 

items and safety issues: 

* Descriptions: 

"o Construction (type); 

"o Fire hazards; 

"o Fire protection features; 

"o Critical process equipment; and 

"o Operations. 

* Potential for a hazardous chemical incident1 or radiation incident from a fire; 

* Impact of natural hazards (earthquake, flood, or wind) on fire safety; 

* Protection of items relied on for safety (IROFS); 

"* Life safety considerations that affect radiation safety; 

"* Emergency planning; 

"* Fire department/brigade response; 

"* Security and safeguards considerations related to fire protection; and 

"* Exposure fire potential and the potential for fire spread between two fire areas. 

D5. The FHA should assume and evaluate the consequences of a single, worst-case 

automatic fire protection system malfunction during a fire. This could be a detection 

system that also functions to activate a pre-action-type sprinkler system. The failures 

and/or events postulated in the analysis should be consistent with the probability criteria 

in the ISA. 

1For the definition of a hazardous chemical, refer to 10 CFR 70.4.

D-1 NUREG-1718
Appendix D

D6. If redundant automatic fire protection systems are provided in the fire area being analyzed 

under D5 of this appendix, only the system that causes the most vulnerable condition is 

assumed to fail. Passive fire protection features, such as blank fire-rated walls or 

continuous fire-rated cable wraps are assumed to remain viable in accordance with their 

fire endurance rating to the extent that they are properly constructed and maintained. 

D7. The FHA is normally organized by the individual fire areas that comprise the facility. As 

defined in Section 7.7, a fire area is a location bounded by fire-rated construction, having 

a minimum fire resistance rating of 2 hours. The FHA through fire modeling (if 

necessary) and fire loading analysis should document that the fire ratings are appropriate 

for each fire area boundary. Where a facility is not subdivided by fire-rated construction, 

the fire area should be defined by the exterior walls and roof of the facility. 

D8. The FHA should contain an inventory of IROFS that are susceptible to fire damage from 

credible fires (taking into account transient and temporary conditions) within each fire 

area. Loss of systems such as ventilation, cooling, or electrical power that could cause 

failures elsewhere in the facility should be evaluated. The FHA should also consider the 

improper operation of equipment due to spurious signals induced by fire damage. In 

addition, the effects of combustion products, manual firefighting efforts, and the activation 

of automatic fire suppression systems should be assessed. 

D9. The FHA may need to produce fire-related parameters (temperatures, pressures, and air 

velocities) for evaluating radioactive material dispersion through the facility air distribution 

system as a result of fire. The radiological consequences should then be determined as 

part of the ISA. 

D1)0. The quantity and associated hazards of flammable and combustible material that can be 

expected to be found within each fire area should be factored into the analyses. 

Consideration should also be given to the presence of transient combustibles associated 

with maintenance activities and storage. Average combustible loading, by itself, should 

not be used to estimate fire area fire severity. At minimum, for each designated fire area, 

the following fire hazards should be evaluated for potential fire severity and consequent 

damage: 

1. Fire load from solid combustible materials (both quantity and configuration) 

including those materials of construction, in-situ materials, and anticipated transient 

combustible materials. Combustibles are defined as materials that do not meet the 

definition of noncombustible material as presented in National Fire Protection 

Association (NFPA) Standard 220. For the purposes of the fire load survey, 

combustibles that can be classified as limited-combustible (as per NFPA 220) may 

be so classified. In performing the fire loading survey, the end uses of the survey in 

the FHA and/or ISA should be kept in mind. These uses may include, but not be 

limited to: determining or verifying the proper design basis of the fire suppression 

system, determining the minimum required fire resistance for barriers, ensuring 

adequate prefire planning, and input to fire propagation or radionuclide transport

NUREG-1 718 D-2
Appendix D

modeling. Each of these uses may require the data to be presented in different 

formats or level of detail. 

2. Flammable and combustible liquids and gases used in the processes within the fire 

area (quantities or flow rates). 

3. Process chemicals and materials (both quantity and location) that could present a 

toxic or radiological hazard, or that could significantly affect health or the quality of 

the environment through a release as a result of a fire emergency. 

4. Potential ignition sources. 

Dl . The FHA should contain an assessment of facility fire water requirements including 

capacity, pressure, and storage requirements. The assessment should include a list of 

water-based automatic suppression systems and their maximum demands, interior hose 

stream requirements, and exterior hydrant requirements. With this assessment, the 

facility fire water system layout should also be provided, including the locations and 

characteristics of pumps, lines, tanks, towers, and sectionalizing valves. 

D1 2. For each designated fire area determined to be important to facility fire safety, or for each 

fire area that contains IROFS, the FHA should provide input to the ISA regarding the 

postulated accident sequences caused or aggravated by fire. Either quantitative or 

qualitative methods may be used. Where quantitative analytical methods are used, all 

input data and assumptions are documented. 

D13. The FHA should define those fire protection systems and procedures that provide 

reasonable assurance that the defined consequences of an accident sequence will not 

occur or will be mitigated. The coverage of fire detection and suppression systems 

should be shown within each fire area. For the identified fire protection measures, the 

applicant should specify compensatory measures to be implemented on a temporary 

basis in the event the identified systems are not operable. Both the compensatory 

measure(s) and the time schedule for implementation should be established. 

Most of the guidance in this appendix originated from, "The Implementation Guide for Use with 

DOE Orders 420.1 and 440.1 -Fire Safety Program," (G-420.1/B-0, G-440.1/E-0, 

September 30, 1995). In some cases, the original guidance was modified to reflect specific 

needs for the MOX facility. 

D14. References 

Department of Energy (U.S.) (DOE). G-420.1/B-0, G-440.1/E-0, "Implementation Guide for 

use with DOE Orders 420.1 and 440.1-Fire Safety Program." DOE: Washington, D.C. 

September 30, 1995. 

National Fire Protection Association, Inc. (NFPA). Standard 220, "Standard on Types of 

Building Construction."

D-3 NUREG-1718
APPENDIX E 

FIRE PROTECTION GUIDANCE FOR NUCLEAR FILTER PLENUMS 

El. Filter Plenum Construction 

All high-efficiency particulate air (HEPA) filters should meet the requirements of American 

Society of Mechanical Engineers (ASME) ASME-AG-1, Section FC, and be listed as 

tested in accordance with Underwriters Laboratories, Inc. (UL) 586. Entrance filters and 

prefilters located upstream or made part of final HEPA filter exhaust plenums should be 

listed as Class 1 air filter units as tested in accordance with UL 900. Filter framing 

systems and filter plenum housing should be of noncombustible construction. 

E2. Fire Rating Requirements for Plenum Housing, Openings, and Dampers 

1. Filter plenum enclosures inside buildings or located less than 1.5 m (5 ft) from an 

adjacent building should be of 2-hour fire-rated construction. For enclosures 

greater than 1.5 m (5 ft) from an existing building, the fire rating may be either 

1 hour or as determined by the fire hazard analysis (FHA). 

2. Door openings into a 2-hour fire-rated filter plenum enclosure should be 1.5-hour 

minimum fire rated. Door openings into a 1-hour fire-rated filter plenum enclosure 

should be 0.75-hour minimum fire rated. 

3. For ducts not required to function as a nuclear confinement system: 

(a) A 1.5-hour damper should be used where the duct penetrates a 2-hour fire

rated barrier. 

(b) A fire damper is not necessary where the duct penetrates a 1-hour barrier 

provided that automatic fire suppression is provided on both sides of the 

barrier and the duct passes through the wall and extends into the area outside 

the enclosure. Transfer grills and similar openings without ducting should be 

provided with an approved damper. 

4. Fire dampers should not be used when penetrating fire-rated construction where 

ducting is an integral part of the air filter system equipment that is required to 

continuously function as part of the confinement system. Such duct material may 

be made part of the fire-rated construction by wrapping, spraying, or enclosing the 

duct with an approved material to provide a minimum 2-hour rating, or be qualified 

for a 2-hour fire-rated exposure to the duct at the penetration location using the fire 

damper criteria as specified in UL 555. 

5. All mechanical and electrical penetrations made into fire-rated plenum enclosures 

should be fire stopped by listed materials meeting the requirements of American 

Society for Testing and Materials (ASTM) ASTM-E-814.

E-1 NUREG-1718
Appendix E

E3. Materials and Hazards Inside Plenums 

1. Filter plenum enclosures should be used only for ventilation control equipment. The 

storage and accumulation of combustible materials (including spare filters) as well 

as combustible and flammable liquids should not be permitted. 

2. Electrical equipment should comply with National Fire Protection Association 

(NFPA) Standard 70, and all electrical wiring inside the enclosure should be in 

metal conduit. 

3. The concentration of flammable vapors inside the final filter plenum should not 

exceed 25 percent of their lower flammable limit. If flammable and combustible 

gases are expected as a result of facility processes, fixed combustible gas 

analyzers should be provided with analyzer alarms set to sound at 25 percent of the 

lower flammable limit and transmitted to a continuously manned position. 

E4. Fire Screens for Filter Plenums 

1. Fire screens should be located upstream from the prefilters and final filter plenums. 

2. Fire screens with metal meshes from 3 to 6 openings per cm (8 to 16 openings per 

inch) should be provided and located at least 1.2 m (4 ft) upstream from all prefilters 

and at least 6.1 m (20 ft) upstream from all final filter plenum enclosures. 

3. Where prefilters are located in final filter enclosures, fire screens should be located 

at least 6.1 m (20 ft) upstream from the prefilters. 

E5. Fire Detection Systems 

1. Automatic fire detectors should be rate-compensated-type heat detectors, approved 

for the specific use and in conformance with NFPA Standard 72. The detectors 

should be of the 880C (190 0 F) temperature range unless operations require higher 

temperature air flows. 

2. Heat detectors or pilot sprinkler heads should be provided in the final filter 

enclosure and in ducting prior to the final filter enclosure. Airflow should be 

considered when determining detector or pilot head location in ducting. 

3. Detector installations should be engineered and installed for testing over the life of 

the detector. Where contamination levels permit, detectors can be removed and 

tested externally.

NUREG-1 718 E-2
Appendix E

E6. Deluge Spray Suppression Systems 

1. Automatic and manual water deluge spray systems should be provided inside all 

final filter plenums for protection of the filters where there is a leading filter surface 

area greater than 1.5 m2 (16 ft2). 

2. Automatic deluge systems should be designed as per the applicable provisions of 

NFPA Standard 13 and Standard 15 and as follows: 

(a) Water spray density should be 10.2 Ipm per m2 (0.25 gpm per ft) over the 

entire filter area or 3.8 Ipm per 14 m3 per min (1.0 gpm per 500 ft3 per min) 

airflow, whichever is greater. 

(b) Spray heads should be deluge-type sprinkler heads. 

(c) The spray pattern of the deluge head should be in the form of a downward 

vertical water curtain approximately 15 cm (6 in) in front of the filter. Heads 

should be spaced so that each head does not exceed 1.2 linear m (4 linear ft) 

of curtain coverage. 

3. Manual spray systems should be designed as per the applicable provisions of 

NFPA Standard 15 and modified as follows: 

(a) Water spray density should be 10.2 1pm per m2 (0.25 gpm per ft2) over the 

entire filter area. 

(b) Nozzles should be deluge spray nozzles that form a full circle, solid cone 

discharge. 

(c) Spray nozzles should be horizontally directed at the face of the HEPA filters 

so that all areas of the first-stage filters and framing support system are 

wetted. 

4. If provided, automatic and manual water spray system water supplies should be 

hydraulically calculated and capable of supplying a simultaneous flow of the 

automatic and manual water spray systems as well as the overhead ceiling 

automatic fire sprinkler systems for the fire area, providing air to the plenum for a 

minimum period of 2 hours. 

5. Water for the deluge spray system should be provided by two separate water 

supply connections for reliability. One connection may be a fire department 

connection. 

6. Demisters should be installed to protect the final stage of HEPA filters from being 

wetted by operation of the deluge water spray system.

E-3 NUREG-1718
Appendix E

Most of the guidance presented here is taken from Department of Energy (U.S.) (DOE) 

Standard, "Fire Protection Design Criteria" (DOE-STD-1066-97). The items of guidance 

presented are considered to be pertinent to the filter systems likely to be used at the mixed 

oxide facility. The items presented also represent the NRC responsibility for fire safety as 

related to facility nuclear safety rather than property protection. A more comprehensive 

discussion of nuclear filter plenum fire protection can be found in Chapter 14 of the DOE 

Standard and the references cited in the standard. 

E7. References 

American Society of Mechanical Engineers (ASME). ASME-AG-1, "Code on Nuclear Air and 

Gas Treatment." 

American Society for Testing and Materials (ASTM). ASTM-E-814, "Fire Test of Through 

Penetration Stops." 

Department of Energy (U.S.) (DOE). DOE-STD-1 066-97, "Fire Protection Design Criteria." 

DOE: Washington, D.C. March 1997. 

National Fire Protection Association, Inc. (NFPA). Standard 13, "Standard for the Installation of 

Sprinkler Systems." 

Standard 15, "Standard for Water Spray Fixed Systems for Fire Protection." 

Standard 70, "National Electric Code." 

Standard 72, "National Fire Alarm Code." 

Underwriters Laboratories, Inc. Standard 555, "Standard for Fire Dampers and Ceiling 

Dampers." 

Standard 586, "High Efficiency Air Filtration Units." 

Standard 900, "Standard for Test Performance of Air Filter Units."

NUREG-1718 E-4
APPENDIX F 

THE NATIONAL ENVIRONMENTAL POLICY ACT AND ENVIRONMENTAL REPORTS 

Fl. Introduction 

The Commission promulgated 10 CFR Part 51 to implement the National Environmental 

Policy Act (NEPA) of 1969, which requires an assessment of the environmental impacts 

for all major Federal actions. The NRC staff conducts an independent assessment for all 

licensing actions that may have a significant effect on the environment, based on the 

information provided by the applicant in an environmental report. An environmental report 

is required for actions listed in 10 CFR 51.60(b). This assessment is documented in an 

Environmental Assessment (EA) and Finding of No Significant Impact (FONSI) or 

Environmental Impact Statement (EIS). 

The Commission determined that the actions listed in 10 CFR 51.22(c) have insignificant 

environmental impacts, and these actions are categorically excluded from the 

requirement for an EA and an environmental report. However, if pursuant to 

10 CFR 51.23(c)(1 1), the action involves an amendment that involves changes in process 

operations or equipment, the applicant must justify that the action will not result in 

significant effects on the environment. 

The Office of Nuclear Material Safety and Safeguards (NMSS) consolidated 

environmental review work into the Division of Waste Management (DWM) on 

May 17, 1999. DWM is responsible for preparing all NMSS EIS and reviewing each EA 

prepared in NMSS. The Division of Fuel Cycle Safety and Safeguards (FCSS) retains the 

responsibility to prepare each EA and FONSI and make determinations regarding the 

applicability of categorical exclusions. As a result, DWM is responsible for determining 

whether the applicant's environmental report is adequate to allow the preparation of an 

EIS; FCSS is responsible for determining if the applicant's environmental report is 

adequate to support the preparation of an EA and a FONSI or, as applicable, to make a 

determination regarding a categorical exclusion. 

Staff coordination on the review of environmental reports used to prepare an EIS should 

be obtained through DWM. Supplementary guidance for FCSS staff use on determining 

the adequacy of environmental reports for an EA and FONSI or to justify the applicability 

of a categorical exclusion is provided in Section F2 for licensing actions after receipt of a 

license to possess and use special nuclear material (SNM). 

Information in Section F2 is presented in parallel with the content of an environmental 

report, as specified in 10 CFR 51.45. This includes: 

o Date of application. - Environmental considerations. 

- Description of the proposed action. 

- Purpose of-the proposed action. - Description of the affected environment. 

- Discussion of considerations.

F-1 NUREG-1718
Appendix F

"o Analysis of environmental effects of the proposed action and alternatives. 

"o Status of compliance. 

"o Adverse information. 

The environmental report may include, reference, or supplement the information 

submitted to the NRC for prior licensing actions. 

Section F3 discusses environmental reports for categorical exclusions and Section F4 

reviews the NEPA documentation and coordination necessary for license amendments. 

F2. Environmental Report Content 

1. Date of Application 

The date of an application for a license to possess and use SNM for processing and 

fuel fabrication, scrap recovery, conversion of uranium hexafluoride, or the conduct 

of any other activity, that the NRC has determined pursuant to 10 CFR Part 51, 

Subpart A, will significantly affect the quality of the environment, is acceptable if the 

application is submitted at least 9 months before the commencement of 

construction, as required by 10 CFR Part 70.21 (f). 

2. Environmental Considerations 

An adequate environmental report addresses the requirements of 10 CFR 51.45(b), 

as described b~low. 

(a) Description of the Proposed Action 

The description of the proposed action includes a brief summary of the 

significant characteristics of the proposed facility, including the major site 

features and the major plant design and operating parameters. The 

description includes a complete discussion about how SNM will be processed 

at the facility. If future construction or expansion is proposed, the description 

includes a proposed project schedule showing the dates for initiation of site 

preparation, plant construction, and operation. 

(b) Purpose of the Proposed Action 

The statement of purpose demonstrates a need for the proposed project. 

This demonstration provides at least the following information: (1) the 

quantities of SNM used for domestic benefit, (2) a projection of national and 

foreign requirements for the services, and (3) alternative sources of supply for 

the proposed facility's services. If delay of the proposed project would have 

effects on the nation's material disposition program or on the applicant's 

business (such as loss of contracts, jobs, or future business), the applicant 

should discuss these effects.

NUREG-1 718 F-2
Appendix F

(c) Description of the Affected Environment 

The description of the affected environment includes: 

(i) Site location (including longitude and latitude) and facility layout; 

(ii) Regional demography and land use; 

(iii) Socioeconomic information, including that for low-income and minority 

populations within a 50-mile radius; 

(iv) Regional historic, archaeological, architectural, scenic, cultural, and 

natural landmarks; 

(v) Local meteorology and air quality; 

(vi) Local surface water and groundwater hydrology; 

(vii) Regional geology and seismology; and 

(viii) Local terrestrial and aquatic ecology. 

To the extent possible, this information is current and reflects observations 

and measurements made over a period of years, especially for conditions that 

are expected to vary seasonally (e.g., precipitation, wind speed and direction, 

and groundwater levels). 

(d) Discussion of Considerations 

The reviewer should find that the discussion of considerations is acceptable if 

it includes: 

(i) Impacts of the proposed action on the environment, such as the: 

"* Effects of site preparation and construction on land use and water 

use; 

"• Effects of plant operation on the human population (including 

consideration of occupational and public radiation exposure) and 

important biota; 

"* Any irreversible commitments of resources because of site 

preparation and plant construction and operation, such as 

destruction of wildlife habitat, removal of land from agricultural 

use, and diversion of electrical power;

F-3 NUREG-1718
Appendix F

Plans and policies regarding deactivation, decommissioning, and 

dismantling at the end of the plant's useful life; 

"* Environmental effects of the transportation of radioactive 

materials to and from the site; 

"* Environmental effects of accidents; 

"* Impacts on air and water quality; and 

"* Impacts on cultural and historic resources. 

The environmental report discusses the impacts on the environment in 

proportion to their significance and considers the cumulative impacts of 

the proposed action. In addition, accident analyses provided in the 

environmental report are consistent with the accident scenarios and 

consequences described in the applicant's Integrated Safety Analysis 

Summary. 

(ii) Adverse environmental effects 

The applicant describes any adverse environmental effects that cannot 

be avoided should the proposal be implemented. This description is 

presented in quantitative terms to the maximum extent possible. This 

discussion makes clear which of these effects are unavoidable and 

subject to later amelioration and which are unavoidable and irreversible. 

The description includes specific measures that the applicant could take 

or plan to take to mitigate adverse effects. 

(iii) Alternatives to the proposed action 

The discussion of alternatives to the proposed action is sufficiently 

complete to aid NRC in developing and exploring, pursuant to 

Section 102(2)(E) of NEPA, "appropriate alternatives to recommended 

courses of action in any proposal which involves unresolved conflicts 

concerning alternative uses of available resources." To the extent 

practicable, the environmental impacts of the proposal and the 

alternatives are presented in comparative form. 

The discussion of alternatives includes siting alternatives and design 

alternatives. Comparable levels of information on each site need not be 

presented as long as the applicant presents sufficient information to 

facilitate a fair and reasonable comparison. As appropriate, the 

following factors are considered when comparing alternative sites:

NUREG-1718 F-4
Appendix F

"* Physical characteristics of the area, including demographic, 

geological, hydrological, meteorological, and seismological 

conditions of the site and surrounding area; 

"* Location of power sources and transmission lines; 

Location of the major product market; 

"* Location of raw materials, components, and sources of supply; 

"* Availability of air, rail, roads, and water for transport of raw 

materials and supplies, finished products, and solid wastes; 

"* Commitment of natural resources for site preparation and plant 

construction, including but not limited to the destruction or 

diminution of wildlife habitats, flora, woodlands, and marshlands; 

"* Commitment of capital for site preparation and plant construction; 

"* Cost of operation, including consideration of labor supply, 

prevailing wage rates, and other recurring or nonrecurring costs; 

"* Availability of municipal services or, conversely, the cost of 

providing services such as water and sewage treatment; 

"* Requirements for relocating homes and families; and 

"* Existing and projected land use and economic status of the 

community (e.g., urban, industrial, stable). 

(iv) Relationship between short-term uses and long-term productivity 

The relationship between local short-term uses of the environment and 

the maintenance and enhancement of long-term productivity is 

discussed. Short-term uses are considered to be those that occur 

during the active life of the facility. Long-term productivity represents 

the use of the environment beyond deactivation and decommissioning 

of the facility. 

(v) Irreversible or irretrievable commitments of resources 

Any irreversible environmental commitments and irretrievable material 

resources that would be involved in the proposed action are discussed.

F-5 NUREG-1 718
Appendix F

3. Analysis of Environmental Effects of the Proposed Action and Alternatives 

An adequate environmental report analyzes the environmental effects of the 

proposed action and alternatives. In accordance with 10 CFR 51.45(c), the analysis 

considers and balances the environmental effects of the proposed action and the 

alternatives available for reducing or avoiding adverse environmental effects, as 

well as the environmental, economic, social, and other benefits of the proposed 

action. 

This analysis quantifies, to the fullest extent practicable, the various factors 

considered. If the application involves renewal or amendment of a current license, 

environmental impacts are quantified using environmental monitoring data collected 

by the licensee. To the extent that there are important qualitative considerations or 

factors that cannot be quantified, the analysis discusses those considerations and 

factors in qualitative terms. The analysis contains sufficient data to aid the staff in 

its development of an independent analysis. 

4. Status of Compliance 

As required by 10 CFR 51.45(d), the applicant should list all Federal permits, 

licenses, approvals, and other entitlements, which must be obtained in connection 

with the proposed action. The list is acceptable if it is complete and current as of 

the application date. 

In addition, 10 CFR 51.45(d) requires that the environmental report include a 

discussion of the status of compliance with applicable environmental quality 

standards and requirements including, but not limited to, applicable zoning and 

land-use regulations, and thermal and other water pollution limitations or 

requirements that have been imposed by Federal, State, regional, and local 

agencies having responsibility for environmental protection. The discussion is 

acceptable if it includes a discussion of whether each alternative will comply with 

such applicable environmental quality standards and requirements. The discussion 

includes, but is not limited to, the following Federal laws: 

(a) The National Historic Preservation Act of 1966; 

(b) The Fish and Wildlife Coordination Act of 1966; 

(c) The Wild and Scenic Rivers Act of 1968; 

(d) The Endangered Species Act Amendments of 1978; and 

(e) The Coastal Zone Management and Improvement Act of 1990. 

5. Adverse Information 

In accordance with 10 CFR 51.45(e), the preceding discussions and analyses are 

acceptable if they include information that is adverse to the proposed actions as 

well as information supporting the proposed action.

NUREG-1718 F-6
Appendix F

F3. Categorical Exclusion 

An environmental report is not required for actions identified in 10 CFR 51.60(b)(1) that 

involve amendments to the mixed oxide (MOX) fuel fabrication facility that are not 

expected to result in significant environmental impacts. However, when these 

amendments involve changes in process operations or equipment, the applicant needs to 

justify that the changes will not result in significant environmental effects. 

The information provided by the applicant to justify the categorical exclusion 

determination for changes in process operations or equipment is acceptable if it 

demonstrates the following as specified in 10 CFR 51.22(c)(1 1): 

"o There is no significant change in the types or significant increase in the amounts of 

any effluents that may be released offsite; 

"o There is no significant increase in individual or cumulative occupational radiation 

exposure; 

"o There is no significant construction impact; and 

"o There is no significant increase in the potential for or consequences from 

radiological incidents. 

Review of the environmental report or information presented to support a categorical 

exclusion includes review of occupational exposure information. This review should be. 

coordinated with the health physics reviewer to assess the adequacy of the information 

provided by the applicant. 

F4. NEPA Documentation and Coordination 

Before taking a licensing action such as a license amendment, the NRC will determine 

whether the proposed action qualifies for a categorical exclusion under 10 CFR 51.22 or 

whether an EA or EIS should be prepared. 

1. An EIS will be prepared if the action meets the criteria in 10 CFR Part 51.20. An EA 

is not necessary if an EIS will be prepared. Coordination with DWM must be 

initiated to prepare the EIS. 

2. A categorical exclusion will suffice if the action meets the criteria for categorical 

exclusions as defined in 10 CFR Part 51.22(c). (An action that qualifies for a 

categorical exclusion is usually identified at the start of the licensing review, and an 

EA is not required.) No coordination with DWM is necessary. 

3. An EA will be prepared if the action meets the criteria in 10 CFR Part 51.21. DWM 

will be informed that an EA will be prepared. DWM should review the completed 

EA. On completion, the NRC determines whether to prepare an EIS or a FONSI.

F-7 NUREG-1718
Appendix F

Requirements for the preparation of an EIS, EA, or FONSI are described in detail in 

10 CFR Part 51. Documents prepared in accordance with NEPA will follow pertinent 

NMSS procedures. 

F5. Environmental Assessment (EA) 

The Special Projects Branch (SPB) staff will prepare an EA that identifies the proposed 

action and includes the following, in accordance with 10 CFR 51.30: 

1. A brief discussion of: 

(a) The need for the proposed action; 

(b) Alternatives to the proposed action as required by Section 102(2)(E) of NEPA; 

(c) The environmental impacts of the proposed action and alternatives, as 

appropriate; and 

(d) As required for special case EAs, as defined by NMSS Policy and Procedures 

Letter 1-50, Revision 2, 1999, disproportionately high and adverse human 

health or environmental effects on low-income and minority populations. 

2. A list of agencies and persons consulted and identification of sources used. During 

preparation of an EA, the staff will consult with affected States on environmental 

issues and will document such contact in the EA. This documentation will include 

the following information identified in NMSS Policy and Procedures Letter 1-48, 

January 1995: 

(a) The name of each State, agency (including contacted individual's name), or 

person consulted; 

(b) The date of consultation(s); 

(c) The purpose of the consultation; 

(d) A brief summary of the views or comments expressed by the consulted party 

and the staff's resolution; and 

(e) References to publicly available documents containing additional information, 

if applicable. 

Much of the information used to prepare the EA is provided by the applicant in the 

environmental report. However, the staff will perform independent analyses of the 

environmental impacts of the proposed action and will discuss the conclusions of 

these analyses in the EA. The EA should focus on the impacts of the proposed

NUREG-1718 F--8
Appendix F

action and should be no more than 15 pages, unless necessary to explain any 

complicated environmental issues associated with the proposed action. 

On completion, the EA should be forwarded to DWM for review. DWM reviews the 

EA to ensure consistency among all EAs prepared by NMSS. When DWM 

completes its review, the staff will determine whether to prepare an EIS or a FONSI 

on the proposed action. As provided in 10 CFR 51.33, a determination to prepare a 

draft FONSI may be made. As provided in 10 CFR 51.25, an EA is not necessary if 

it is determined that an EIS will be prepared. 

F6. Finding of No Significant Impact (FONSI) 

When the staff makes a final finding that there are no significant environmental impacts of 

the proposed action, a final FONSI will be published in the Federal Register. The 

Commission will not take the proposed action until after the FONSI is published. 

Requirements for the preparation of a FONSI for materials licensing actions are contained 

in 10 CFR 51.32-51.35. A FONSI will include the following: 

1. Identification of the proposed action; 

2. A statement that the Commission has determined not to prepare an EIS for the 

proposed action; 

3. A brief presentation of the reasons why the proposed action will not have a 

significant impact on the quality of the human environment; 

4. The EA or a summary of the EA; 

5. A note of any other related environmental documents; and 

6. A statement that the finding and any related environmental documents are available 

for public inspection and where the documents may be inspected. 

NRC may make a determination to prepare and issue a draft FONSI for public review and 

comment before making a final determination whether to prepare an EIS or a final FONSI 

on the proposed action. A draft FONSI may be prepared if a FONSI appears warranted, 

but the proposed action is similar to one that normally requires an EIS or is without 

precedent. 

The draft FONSI will be identified as a draft and will contain the information specified 

above for a final FONSI. The draft FONSI will be accompanied by or will include a 

request for comments on the proposed action and the draft findings within 30 days, or a 

longer period as may be specified in the notice of the draft findings. This draft FONSI will 

be published in the Federal Register, distributed as provided in 10 CFR 51.74(a), and 

made available in accordance with 10 CFR 51.123.

F-9 NUREG-1718
Appendix F

When a draft FONSI is issued, a final determination to prepare an EIS or final FONSI will 

not be made until the last day of the public comment period has expired. 

F7. Environmental Impact Statement (EIS) 

When the NRC determines that an EIS will be prepared for the licensing action, 

coordination should be initiated with DWM, which will review the environmental report and 

prepare the EIS. The environmental reviewer should coordinate with the FCSS Project 

Manager and DWM to ensure consistency between the environmental review for licensing 

and the preparation of the EIS. This coordination minimizes potential issues between the 

safety evaluation and the NEPA analysis, and ensures the results of the NEPA analysis 

are appropriately incorporated into the Safety Evaluation Report (SER) for the 

construction approval review and the SER for the review for the license to possess and 

use SNM.

NUREG-1718 F-1 0
APPENDIX G 

CHECKLIST FOR EVALUATING ACCEPTANCE OF QUALITY ASSURANCE ELEMENTS 

G1. Organization 

The organizational elements responsible for Quality Assurance (QA) are acceptable if: 

1. The responsibility for the overall QA is retained and exercised by the applicant. 

2. The applicant identifies and describes the major delegation of work involved in 

establishing and implementing its QA program or any part thereof to other 

organizations. 

3. When major portions of the applicant's QA program are delegated: 

(a) The applicant describes how responsibility is exercised for overall QA. The 

extent of management supervision should be given, including the position 

location, qualifications, and criteria for determining the number of personnel 

performing these functions. 

(b) The applicant evaluates the performance of work by the delegated 

organization. Method and frequency-once per year, although a longer cycle 

is acceptable with other evaluations of individual elements-are stated. 

(c) Qualified individuals or organizational elements are identified by position title 

within the applicant's organization as responsible for the quality of the 

delegated work before activities are started. 

4. Clear management controls and effective lines of communication exist for QA 

activities among the applicant, contractors, and suppliers to ensure direction of QA. 

5. Organizational charts clearly identify all the onsite and offsite organizational 

elements that function under the purview of QA (such as design, engineering, 

procurement, manufacturing, construction, inspection, testing, instrumentation, 

control, operation, and maintenance), the lines of responsibility, and the criteria for 

determining the size of the QA organization, including the inspection staff. 

6. The applicant describes the QA responsibilities of each of the organizational 

elements noted on the organization charts. 

7. The applicant identifies a management position that retains overall authority and 

responsibility for QA. This position may be filled by a person having the title "QA 

Manager" or another individual performing that function. This position has the 

following characteristics: 

(a) The position resides at least at the same organizational level as the position 

of the highest line manager directly responsible for performing activities that

G-1 NUREG-1718
Appendix G

affect the quality/safety of facility operations (such as engineering, 

procurement, construction, and operation) and is independent of operational 

restraints. 

(b) The person in the position has effective communication channels with other 

senior management personnel. 

(c) The person in the position has responsibility for approval of QA manuals. 

8. Conformance to established requirements (except for designs) is verified by 

individuals or groups within the QA organization who do not have direct 

responsibility for performing the work being verified or by individuals or groups 

trained and qualified in QA concepts and practices who are independent of the 

organization responsible for performing the task. 

9. Persons and organizations performing QA functions have sufficient access to 

management at a level necessary to ensure the capability to: 

(a) Identify quality/safety problems. 

(b) Initiate, recommend, or provide solutions through designated channels. 

(c) Verify implementation of solutions. 

Positions with the above authority are identified by position title and a 

description of how the above actions are carried out is provided. 

10. When work contributes to a situation adverse to safety and has to be stopped, the 

following provisions apply: 

(a) Designated QA personnel, sufficiently free from direct pressures resulting 

from operational concerns, have the responsibility, delineated in writing, to 

stop work in unsafe situations and to control further operations until the 

conditions that created the unsafe condition are corrected. 

(b) The organizational positions with stop-work authority are identified. 

11. Provisions are established for the resolution of disputes involving quality of items 

relied on for safety arising from a difference of opinion between QA personnel and 

personnel from other departments (engineering, procurement, manufacturing, etc.). 

12. Designated QA individuals are involved in day-to-day activities relied on for safety of 

facility conditions and operations. QA staff members routinely attend and 

participate in status meetings to ensure that they are kept abreast of day-to-day 

activities and that there is adequate QA coverage of those activities.

NUREG-1718 G-2
Appendix G

13. Policies regarding the implementation of QA are documented and made mandatory. 

These policies are established at the facility management or corporate level. 

14. The position description ensures that the individual directly responsible for the 

definition, direction, and effectiveness of overall QA has sufficient authority to 

effectively implement responsibilities. This position is to be sufficiently free from 

operational responsibilities to ensure independence of action. Qualification 

requirements for this individual are established in a position description that 

includes the following prerequisites: 

(a) Management experience through assignments to responsible positions. 

(b) Knowledge of QA regulations, policies, practices, and standards. 

(c) Experience in performing QA or QA-related activities in design, construction, 

or operation in a fuel cycle, plant, a power reactor, a low-level waste facility, or 

in a similar high-technology industry. 

15. The person responsible for onsite QA is identified by position and has the 

appropriate organizational position, responsibilities, and authority to exercise proper 

control over QA. The duties of this individual are structured such that adequate 

attention can be given to ensuring that QA at the plant site is being effectively 

implemented. 

Additional guidance for organization is given in SRP Section 4.0, "Organization and 

Administration." 

G2. QA Function 

The QA function for items relied on for safety is acceptable if: 

1. The scope of QA includes: 

(a) A commitment that activities affecting the quality of design, construction, and 

operation will be subject to the applicable controls of QA. Activities covered 

by QA are identified on QA-defining documents. 

(b) A commitment that any test program for items relied on for safety (IROFS) will 

be conducted with QA controls and a description of how QA will be applied. 

(c) A commitment that computer programs for functions related to safety will be 

procured/developed, modified, maintained, and used in accordance with QA 

controls and a description of how QA will be applied.

G-3 NUREG-1718
Appendix G

(d) A commitment that special items, environmental conditions, skills, or 

processes will be provided as necessary to ensure the quality of activities 

having an effect on safety. 

2. A brief summary of the applicant's corporate QA policies is given. 

3. The following provisions are established to ensure that quality-affecting procedures 

required to implement QA are consistent with QA commitments and corporate 

policies and are properly documented, controlled, and made mandatory through a 

policy statement or equivalent document signed by the responsible official: 

(a) The QA organization reviews and documents concurrence in the quality

affecting procedures. 

(b) The organizational group or individual responsible for the policy statement is 

identified. 

(c) The quality-affecting procedural controls of the principal contractors are 

provided for the applicant's review with documented agreement of acceptance 

before the initiation of activities relied on for safety. 

4. Provisions are included for notifying the NRC of changes in the implementation of 

QA from that described in the application. 

5. The QA organization and the necessary technical organizations participate early in 

the QA definition stage to determine and identify QA controls and the extent to 

which they are to be applied to items as they relate to safety. This effort may 

involve applying a defined, graded approach to the items in accordance with their 

importance to safety. 

6. A description is provided that emphasizes how the detailed QA will be properly 

implemented and carried out. 

7. A description is provided of how management (above or outside the QA 

organization) regularly assesses the scope, status, adequacy, and compliance of 

QA. These measures should include: 

(a) Frequent appraisals of QA status through reports, meetings, audits and/or self 

assessments. 

(b) Performance of an annual, preplanned, and documented assessment. 

(c) Identification and tracking of corrective actions based on assessment 

findings.

NUREG-1718 G-4
Appendix G

8. Activities that are IROFS (such as design, procurement, and site investigation) 

initiated prior to formal NRC acceptance of the QA program are controlled by a QA 

program in accordance with this SRP section. Approved procedures and 

appropriately trained personnel should be available to implement the applicable 

portion of the QA program prior to the initiation of the activity. 

9. A summary description is provided on how responsibilities and control of quality

affecting activities are transferred from the principal contractors to the applicant as 

the design and construction phase is completed. 

10. Indoctrination, training, and qualification1 are established so that: 

(a) Personnel responsible for performing and verifying activities affecting quality 

are instructed as to the purpose, scope, and implementation of the applicable 

manuals, instructions, and procedures. 

(b) Personnel performing and verifying activities affecting safety and/or quality 

are trained and qualified in the principles, techniques, and requirements of the 

activity being performed. 

(c) For formal training and qualification, documentation includes a statement of 

the training objective and its content, the attendees, and the date of 

attendance. 

(d) Proficiency tests are given to those personnel performing and verifying 

activities affecting quality, and acceptance criteria are developed to determine 

if individuals are properly trained and qualified. 

(e) The certificate of qualifications clearly delineates the specific functions 

personnel are qualified to perform and the criteria used to qualify personnel in 

each function. 

(f) Proficiency of personnel performing and verifying activities affecting 

safety/quality is maintained by retraining, reexamining, and/or recertifying, as 

determined by management or program commitment. 

11. The applicant's ISA is developed and maintained under QA controls. 

G3. Design Control2 

Control of the design of items relied on for safety is acceptable if: 

1 Guidance for training and qualification of plant personnel is given in SRP Section 15.4. 

2 Guidance for configuration management is given in SRP Section 15.2.

G-5 NUREG-1718
Appendix G

1. The scope of design control includes design activities associated with the 

preparation and review of design documents, including the correct translation of 

applicable regulatory safety requirements and associated design bases into design, 

procurement, and procedural documents. 

2. Organizational responsibilities are described for preparing, reviewing, approving, 

and verifying design documents related to an item or its processes, such as system 

descriptions, design input and criteria, design drawings, design analyses, computer 

programs, specifications, and procedures. 

3. Organizational responsibilities are described for planning and conducting site 

characterization, including reviewing, approving, and verifying analyses and 

conclusions. 

4. Errors and deficiencies in approved design documents, including design methods 

(such as computer codes) that could adversely affect the performance of items and 

processes are documented, and action is taken to ensure that all errors and 

deficiencies are corrected. 

5. Deviations from specified quality standards are identified, and procedures are 

established to ensure their control. 

6. Internal and external design interface controls, procedures, and lines of 

communication among participating design organizations and across technical 

disciplines are established and described for the review, approval, release, 

distribution, and revision of documents involving design interfaces to ensure that 

items are compatible geometrically and functionally. 

7. Procedures are established and described requiring documented verification of the 

dimensional accuracy and completeness of design drawings and specifications. 

8. Procedures are established and described requiring that design drawings and 

specifications for items relied on for safety be reviewed by the QA organization to 

ensure that the documents are prepared, reviewed, and approved in accordance 

with company procedures and that the documents contain necessary QA 

requirements, such as inspection and test requirements, acceptance requirements, 

and those pertaining to the extent of documenting inspection and test results. 

These reviews are documented. 

9. Guidelines or criteria are established and described for determining the method of 

design verification (design review, alternate calculations, or tests). 

10. Procedures are established and described for design verification activities that 

ensure the following:

NUREG-1718 G-6
Appendix G

(a) The verifier is qualified, and neither the verifier nor the verifier's immediate 

supervisor is directly responsible for the design. In exceptional 

circumstances, the designer's immediate supervisor may perform the 

verification provided: 

(i) The supervisor is the only technically qualified individual; 

(ii) The need is individually documented and approved in advance by the 

supervisor's management; and 

(iii) QA audits and self-assessments cover frequency and effectiveness of 

the use of supervisors as design verifiers to guard against abuse. 

(b) Design verification is completed before release of procurement, 

manufacturing, or construction to another organization for use in other design 

activities. When this schedule cannot be met, the design verification may be 

deferred, provided the justification for this action is documented and the 

unverified portion of the design output document and all design output 

documents, based on the unverified data, are appropriately identified and 

controlled. Construction site activities associated with a design or design 

change should not proceed without verification past the point where the 

installation would become irreversible (i.e., require extensive demolition and 

rework). 

(c) Procedural control is established for design documents that reflect the 

commitments for construction approval and the license to possess and use 

special nuclear material. Procedural control differentiates between 

documents that undergo formal design verification by interdisciplinary or 

multiorganizational teams and those that can be reviewed by a single 

individual (a signature and date are acceptable documentation for personnel 

certification). Design documents that pertain to plant safety and are subject 

to procedural control include, but are not limited to, specifications, 

calculations, computer programs, system descriptions, and drawings 

(including flow diagrams, piping and instrument diagrams, control logic 

diagrams, electrical single-line diagrams, diagrams of structural systems for 

major facilities, site arrangements, and equipment locations). Specialized 

reviews should be used when uniqueness or special design considerations 

warrant them. 

(d) The responsibilities of the verifier, the areas and features to be verified, the 

pertinent considerations to be verified, and the extent of documentation are 

identified in procedures. 

11. The following provisions are included if the verification method is only by test: 

(a) Procedures provide criteria that specify when verification should be by test.

G-7 NUREG-1718
Appendix G

(b) Prototype, component, or feature testing is performed as early as possible 

before installation of plant items or before such installation would become 

irreversible. 

(c) Verification by test is performed under conditions that simulate the most 

adverse design conditions as determined by analysis. 

12. Procedures are established to ensure that verified computer codes are certified for 

use and that their use is specified. 

13. Design and specification changes, including field changes, are subject to the same 

design controls that were applicable to the original design. 

G4. Procurement Document Control 

Control of procurement documents for the procurement of IROFS is acceptable if: 

1. Procedures are established for the review of procurement documents to determine 

that quality requirements are correctly stated, inspectable, and controllable; that 

there are adequate acceptance and rejection criteria; and that procurement 

documents have been prepared, reviewed, and approved in accordance with QA 

requirements. To the extent necessary, procurement documents should require 

that contractors and subcontractors provide acceptable QA. The review and 

documented concurrence of the adequacy of quality requirements stated in 

procurement documents are performed by independent personnel trained and 

qualified in QA practices and concepts. 

2. Procedures are established to ensure that procurement documents identify 

applicable regulatory, technical, administrative, and reporting requirements; 

drawings; specifications; codes or industry standards; inspection and test 

requirements; and special process instructions that must be met by suppliers. 

3. Organizational responsibilities are described for procurement planning; the 

preparation, review, approval, and control of procurement documents; supplier 

selection; bid evaluations; and the review of and concurrence with supplier QA 

before initiation of activities relied on for safety. The involvement of the QA 

organization is described. 

G5. Instructions, Procedures 3, and Drawings 

Activities related to instructions, procedures, and drawings pertaining to items relied on for 

safety are acceptable if: 

3 Guidance for plant procedures is given in SRP Section 15.5.

NUREG-1718 G-8
Appendix G

1. Organizational responsibilities are described for ensuring that activities affecting the 

quality of IROFS are prescribed by documented instructions, procedures, and 

drawings and accomplished through implementation of these documents. 

2. Procedures are established to ensure that instructions, procedures, and drawings 

that could affect safety include quantitative acceptance criteria (such as those 

pertaining to dimensions, tolerances, and operating limits) for determining that 

activities relied on for the safety of plant operations have been satisfactorily 

performed. 

G6. Document Control 

Control of documents related to IROFS is acceptable if: 

1. The scope of document control is described and the types of controlled documents 

are identified. As a minimum, controlled documents include: 

(a) Design documents (e.g., calculations, drawings, specifications, analyses), 

including documents related to computer codes; 

(b) Procurement documents; 

(c) Instructions and procedures for such activities as fabrication, construction, 

modification, installation, maintenance, testing, and inspection; 

(d) Documents pertaining to as-built conditions; 

(e) QA and quality control manuals, procedures, and reports; and 

(f) Technical reports. 

2. Procedures for the review, approval, and issuance of documents and changes thereto are established and described to ensure technical adequacy and inclusion 

of appropriate safety/quality requirements before implementation. The QA organization, or an individual other than the person who generated the document 

but who is qualified in QA, reviews and concurs with these documents in regard to 

QA-related aspects. 

3. Procedures are established to ensure that changes to documents are reviewed and 

approved by the same organizations as those that performed the initial review and 

approval or by other qualified, responsible organizations delegated by the applicant. 

4. Before commencing work, procedures are established to provide adequate 

assurance that documents are available at the location where the activity will be 

performed or that procedures are readily accessible to all personnel performing the 

work.

G-9 NUREG-1718
Appendix G

5. Procedures are established and described to provide adequate assurance that 

obsolete or superseded documents are removed and replaced by applicable 

revisions in work areas in a timely manner. 

6. A master list or equivalent document control system is established to identify the 

current revision of instructions, procedures, specifications, drawings, and 

procurement documents. When such a list is used, it should be updated and 

distributed to predetermined responsible personnel. 

7. Procedures are established.and described to provide for the preparation of 

drawings pertaining to as-built conditions and related documentation in a timely 

manner to accurately reflect the actual design. 

G7. Control of Purchased Items 

Control of purchased IROFS is acceptable if: 

1. Organizational responsibilities are described for the control of purchased items 

including interactions between design, procurement, and QA organizations. 

2. Verification of suppliers' activities during fabrication, inspection, testing, and 

shipment of IROFS is planned and performed with QA organization participation in 

accordance with written procedures to ensure conformance to the purchase order 

requirements. The procedures, as applicable to the method of procurement, 

provide for: 

(a) The specification of the characteristics or processes to be witnessed, 

inspected, or otherwise verified; the method of verification and the required 

documentation; and the personnel responsible for implementing these 

procedures; and 

(b) Audits, surveillances, or inspections that ensure that the supplier complies 

with the quality requirements. 

3. Procurement of spare or replacement parts for IROFS is subject to QA controls, to 

codes or standards, and to technical requirements equal to or better than the 

original technical requirements, or as required to prevent the procurement of 

defective items. 

4. Selection of suppliers is documented and filed. 

5. Items are inspected when received to ensure: 

(a) The item is properly identified and corresponds with the identification on the 

purchase document and the documentation when the item is received.

NUREG-1718 G-10
Appendix G

(b) The item and acceptance records satisfy the inspection instructions before 

installation or use of the item. 

(c) Specified inspection, test, and other records (such as certificates of 

conformance attesting that the item conforms to specified requirements) are 

available at the facility before installation or use of the item. 

6. Items accepted and released are identified as to their inspection status before they 

are forwarded to a controlled storage area or released for installation or further 

work. 

7. The supplier furnishes the following records to the purchaser: 

(a) Documentation that identifies the purchased item and the specific 

procurement requirements (e.g., codes, standards, specifications) met by the 

IROFS; 

(b) Documentation that identifies any procurement requirements that have not 

been met; and 

(c) A description of those items that do not conform to the procurement 

requirements and that are designated "accept as is" or "repair." 

The procedure for review and acceptance of these documents is described. 

8. For commercial "off-the-shelf" items for which specific QA controls cannot be 

imposed in a practicable manner, special quality verification requirements are 

established and described to ensure that an acceptable item has been received by 

the purchaser. 

9. Suppliers' certificates of conformance are periodically evaluated by audits, 

independent inspections, or tests to ensure that they are valid and that the results 

are documented. 

G8. Identification and Control of Items 

Identification and control of IROFS are acceptable if: 

1. Controls are established and described to identify and control IROFS. The 

description should include organizational responsibilities. 

2. Procedures are established that ensure that identification is maintained either on 

the IROFS or on records traceable to the item, to preclude use of incorrect or 

defective items.

G-1 1 NUREG-1 718
Appendix G

3. Identification of IROFS can be traced to the appropriate documentation, such as 

drawings, specifications, purchase orders, manufacturing and inspection 

documents, deviation reports, and physical and chemical test reports. 

4. Correct identification of items is verified and documented before they are released 

for fabrication, assembling, shipping, and installation. 

G9. Control of Special Processes 

Control of special processes related to IROFS is acceptable if: 

1. Organizational responsibilities, including those for the QA organization, are 

described for the qualification of special processes, equipment, and personnel. 

2. Procedures are established for recording evidence of an acceptable level of quality 

for special processes, using qualified procedures, equipment, and personnel. 

3. Qualification records of procedures, equipment, and personnel associated with 

special processes are established, filed, and kept current. 

G10. Inspection 

Inspection of items relied on for plant or process safety is acceptable if: 

1. The scope of inspection indicates that an effective inspection program has been 

established. Procedures provide criteria for determining the accuracy requirements 

of inspection equipment and criteria for determining when inspections are required 

or for defining how and when inspections are performed. The QA organization 

participates in these functions. 

2. Organizational responsibilities for inspection are described. Individuals performing 

inspections are other than those who performed or directly supervised the 

item/activity being inspected and do not report directly to the immediate supervisors 

who are responsible for the item/activity being inspected. If the individuals 

performing inspections are not part of the QA organization, the inspection 

procedures, personnel qualification criteria, and independence from undue 

pressure, such as operational needs, should be reviewed and found acceptable by 

the QA organization before the initiation of the activity. 

3. A qualification plan for inspectors is established and documented and the 

qualifications and certifications of inspectors are kept current. 

4. Inspection procedures, instructions, or checklists provide for the following: 

(a) Identification of characteristics and activities to be inspected;

NUREG-1 718 G-1 2
Appendix G

(b) A description of the method of inspection; 

(c) Identification of the individuals or groups responsible for performing the 

inspection in accordance with the provisions of Item 2 in this section; 

(d) Acceptance and rejection criteria; 

(e) Identification of required procedures, drawings, specifications, and revisions; 

(f) Identification of inspection personnel, measuring and test equipment used 

(including any data recorders), and the results of the inspection; and 

(g) Specification of the necessary measuring and test equipment, including 

accuracy requirements. 

5. Inspection results are documented and evaluated and their acceptability is 

determined by a responsible individual or group. 

G11. Test Control 

Control of tests of IROFS is acceptable if: 

1 . The description of the scope of test control indicates that an effective test program 

has been established for tests, including proof tests before installation and 

preoperational tests. Procedures provide criteria for determining the accuracy 

requirements of test equipment and for determining when a test is required or how 

and when testing activities should be performed. 

2. Test procedures or instructions provide, as required, for the following: 

(a) The requirements and acceptance limits in applicable design and 

procurement documents; 

(b) Instructions for performing the test; 

(c) Test prerequisites such as calibrated instrumentation, adequate test 

equipment and instrumentation, including their accuracy requirements, 

completeness of items to be tested, suitable and controlled environmental 

conditions, and provisions for data collection and storage; 

(d) Test acceptance and rejection criteria; 

(e) Mandatory inspection hold points for witness by owner, contractor, or 

inspector (as applicable); 

(f) Methods of documenting or recording test data and results; and

G-1 3 NUREG-1718
Appendix G

(g) Provisions for ensuring that test prerequisites have been met. 

3. Test results are documented and evaluated and their acceptability is determined by 

a responsible individual or group. 

4. A qualification plan is established and documented for those individuals conducting 

the tests, so that certifications for those individuals performing the tests are kept 

current. 

G12. Control of Measuring and Test Equipment 

Control of measuring and test equipment (such as instruments, tools, gauges, fixtures, 

reference and transfer standards, and nondestructive test equipment) identified as 

IROFS or used to measure or test other IROFS is acceptable if: 

1. The scope for the control of measuring and test equipment is described, along with 

the types of equipment to be controlled. This information indicates that effective 

calibrations and adjustments have been established. 

2. QA and other organizations' responsibilities are described for establishing, 

implementing, and ensuring the effectiveness of the calibrations and adjustments. 

3. Procedures are established and described for calibration (technique and 

frequency), maintenance, and control of measuring and test equipment. The review 

of and documented concurrence with these procedures are described and the 

organization responsible for these functions is identified. 

4. Measuring and test equipment is identified and traceable to the calibration data. 

5. Measuring and test equipment is labeled, tagged, or otherwise controlled to indicate 

the due date of the next calibration. The method to "otherwise control" measuring 

and test equipment should be described. 

6. Measuring and test equipment is calibrated at specified intervals on the basis of the 

required accuracy, purpose, degree of usage, stability characteristics, and other 

conditions affecting the measurement. The test equipment should have sufficient 

accuracy to ensure that the equipment being calibrated is within required tolerance, 

and the basis of acceptance is documented and authorized by responsible 

management. The management authorized to perform this function is identified. 

7. Calibrating standards have greater accuracy than standards being calibrated. 

Calibrating standards with the same accuracy may be used if they can be shown to 

be adequate to meet the requirements, and the basis of acceptance is documented 

and authorized by a responsible member of the management staff. The 

management staff member authorized to perform this function is documented.

NUREG-1718 G-14
Appendix G

8. Reference and transfer standards are traceable to nationally recognized standards; 

where national standards do not exist, provisions are established to document the 

basis for calibration. 

9. Measurements are taken and documented to determine the validity of previous inspections and the acceptability of items inspected or tested since the last 

calibration when measuring and test equipment is found to be out of calibration. 

Inspections or tests are repeated on items determined to be suspect. 

G13. Handling, Storage, and Shipping 

Handling, storage, and shipping of IROFS are acceptable if: 

1. Special handling, preservation, storage, cleaning, packaging, and shipping requirements are established and implemented by suitably trained individuals in 

accordance with predetermined work and inspection instructions. 

2. Procedures are established and described to control the cleaning, handling, 

storage, packaging, and shipping of items in accordance with design and procedure 

requirements. 

G14. Inspection, Test, and Operating Status 

Inspection, test, and operating status of IROFS are acceptable if: 

1. Procedures are established to indicate the inspection, test, and operating status of 

items. 

2. Procedures are established and described to control the application and removal of 

inspection and welding stamps and status indicators such as tags, markings, labels, 

and stamps. 

3. Procedures are established and described to control the alteration of the sequence 

of required tests, inspections, and other operations relied on for safety. Such 

actions should be subject to the same controls as those for the original review and 

approval. 

4. The status of nonconforming, inoperative, or malfunctioning items and processes is 

documented and identified to prevent inadvertent use. The organization 

responsible for this function is identified. 

G15. Nonconforming Items 

Control of nonconforming I ROFS is acceptable if:

G-15 NUREG-1718
Appendix G

1. Procedures are established and described for the identification, documentation, 

segregation, review, disposition, and notification to affected organizations of 

nonconforming items (including computer codes) if disposition is other than to 

scrap. The procedures identify authorized individuals responsible for the 

independent review of nonconforming items, including their disposition and 

closeout. 

2. QA and other organizational responsibilities are described for the definition and 

implementation of activities related to nonconformance control. This includes 

identifying those individuals or groups with authority for the disposition of 

nonconformance. 

3. Documentation identifies the nonconforming item; describes the nonconformance, 

the disposition of the nonconforming item, and the inspection requirements; and 

includes signature approval of the disposition. Nonconformances are corrected or 

resolved before the initiation of preoperational testing of the item. 

4. Reworked, repaired, and replacement items are inspected and tested in accordance 

with the original inspection and test requirements or acceptable alternatives. 

5. Nonconformance reports are periodically analyzed by the QA organization to show 

quality trends, and significant results are reported to upper management for review 

and assessment. 

G1 6. Corrective Action 

Corrective actions that affect or support IROFS are acceptable if: 

1. Procedures are established and described indicating that effective corrective 

actions have been established. The QA organization reviews and documents 

concurrence with the procedures. 

2. Corrective action is documented and initiated after the determination of a condition 

adverse to safety/quality (i.e., nonconformance, failure, malfunction, deficiency, 

deviation, defective item, a failure to follow operating procedures, or a human error) 

to preclude recurrence. The QA organization concurrence is required regarding the 

adequacy of the corrective action. 

3. Followup action is taken by the QA organization to verify proper implementation of 

corrective action and to close out the corrective action in a timely manner. 

4. Significant conditions adverse to safety, the root cause of the conditions, and the 

corrective action taken to preclude repetition are documented and reported to 

immediate management and upper levels of management for review and 

assessment.

NUREG-1718 G-1 6
Appendix G

G17. QA Records4 

Control of QA records is acceptable if: 

1. QA and other organizations are identified and their responsibilities are described for 

the definition and implementation of QA records. 

2. Inspection and test records contain the following, where applicable: 

(a) A description of the type of observation; 

(b) The date and results of the inspection or test; 

(c) Information on conditions adverse to quality; 

(d) Identification of the inspector or data recorder; 

(e) Evidence as to the acceptability of the results; and 

(f) Action taken to resolve any discrepancies noted. 

3. Suitable facilities for the storage of the records are described. 

G18. Audits and Assessments 

Guidance for audits and assessments is given in SRP Section 15.6. 

G19. Applicant's Provisions for Continuing QA 

The applicant's provisions for continuing QA are acceptable if the submittal addresses 

reviews and updates based on reorganizations, revised activities, lessons learned, 

changes to applicable regulations, and other QA changes that should be reflected in the 

license application's QA program description to keep it current. 

4 Additional guidance for records management is given in SRP Section 15.8. 

G-17 NUREG-1718
APPENDIX H 

CHECKLIST FOR PLANT PROCEDURES 

The list below shows activities that should be covered by written procedures. The list is not 

intended to be all inclusive, nor is it intended to imply that procedures be developed with the 

same titles as those on the list. 

H1. Operating Procedures 

1. Procedures that address startup, operation, shutdown, control of process 

operations, and recovery after a process upset include: 

(a) Ventilation; 

(b) Criticality alarms; 

(c) Shift routines, shift turnover, and operating practices; 

(d) Decontamination operations; 

(e) Plant utilities (air, other gases, cooling water, firewater, steam); 

(f) Temporary changes in operating procedures; and 

(g) Abnormal operation/alarm response: 

(i) Loss of cooling water; 

(ii) Loss of instrument air; 

(iii) Loss of electrical power; 

(iv) Loss of criticality alarm system; 

(v) Loss of containment; 

(vi) Fires; and 

(vii) Chemical process releases. 

2. Maintenance activities that address repair, calibration, surveillance, and functional 

testing include: 

(a) Repairs and preventive repairs of items relied on for safety (IROFS); 

(b) Testing of criticality alarm units; 

(c) Calibration of IROFS; 

(d) High efficiency air particulate (HEPA) filter maintenance; 

(e) Functional testing of IROFS; 

(f) Relief valve replacement/testing; 

(g) Surveillance/monitoring; 

(h) Pressure vessel testing; 

(i) Piping integrity testing; and 

(j) Containment device testing. 

3. Emergency procedures include: 

(a) Response to a criticality, and 

(b) Hazardous process chemical releases.

H-1 NUREG-1 718
Appendix H

H2. Management Control Procedures 

1. Training; 

2. Audits and assessments; 

3. Incident investigation; 

4. Records management; 

5. Configuration management; 

6. Quality assurance; 

7. Equipment control (lockout/tagout); 

8. Shift turnover; 

9. Work control; 

10. Management control; 

11. Procedure management; 

12. Nuclear criticality safety; 

13. Fire protection; 

14. Radiation protection; 

15. Radioactive waste management; 

16. Maintenance; 

17. Environmental protection; 

18. Chemical process safety; 

19. Operations; 

20. Calibration control; 

21. Preventive maintenance; 

22. Design control; and 

23. Test control.

NUREG-1718 H-2
APPENDIX I 

HEALTH AND SAFETY RECORDS 

The requirements for records management will depend on the applicable hazards and risks 

determined for the facility. Examples of the types of records that should be included in the 

system required by 10 CFR Parts 19, 20, 21, 25, and 70 are listed in Section I1 below. 

Section 12 lists examples of the types of records that should be established and maintained to 

provide reasonable assurance that items relied on for safety (IROFS) will be available and 

reliable to perform their function when needed, as required by 10 CFR 70.64(1). Section 12 is 

organized under the chapter headings of the SRP. 

Although Sections I1 and 12 lists examples of records, the lists are not intended to be 

exhaustive or prescriptive in format. Furthermore, the applicant may choose to organize the 

records in ways other than shown here. 

!1. Examples of Records Required by 10 CFR Parts 19, 20, 21, 25, and 70 

(1) Audits; 

(2) Access authorization for personnel; 

(3) Administrative procedures with safety implications; 

(4) Air sample data; 

(5) Bioassay data; 

(6) Change control records for material control and accounting program; 

(7) Radiation dose to individuals of the public; 

(8) Radiation exposure history; 

(9) Individual radiation monitoring data; 

(10) Individual radiation monitoring results; 

(11) Individual intakes of radioactive material; 

(12) Radioactive material storage records; 

(13) Planned special radiation exposures; 

(14) Radiation protection (and contamination control) recordsp; 

(15) Radiation training records; 

(16) Radiation work permits; 

(17) Records of cumulative occupational radiation dose; 

(18) Records of receipt, transfer, and disposal of radioactive material; 

(19) Records of radioactive waste disposal; 

(20) Reports of theft/loss of licensed material; 

(21) Results-of radiation surveys/calibrations; 

(22) Results of measurements used to calculate radioactive effluents; 

(23) Health and safety compliance records, medical records, personnel exposure 

records, etc. 

12. Examples of Records that Should Provide Reasonable Assurance that IROFS Will 

Be Available and Reliable to Perform their Function (Listed by SRP Section) 

(1) General information:

1-1 NUREG-1718
Appendix I

(a) Construction records; 

(b) Facility and equipment descriptions and drawings; 

(c) Design criteria, requirements, and bases for safety-related structures, 

systems, or components, as specified by the facility configuration 

management system; 

(d) Records of facility changes and associated integrated safety analyses, as 

specified by the facility configuration management system; 

(e) Safety analyses, reports, and assessments; 

(f) Records of site characterization measurements and data; 

(g) Records pertaining to onsite disposal of radioactive or mixed wastes in 

surface landfills; and 

(h) Specifications for items relied on for safety. 

(2) Financial qualifications: None. 

(3) Protection of classified matter: 

(a) Procedures to prevent tampering and loss of classified/sensitive records; and 

(b) Employee access authorization lists. 

(4) Organization and administration: 

(a) Administrative procedures with safety implications; 

(b) Change control records for material control and accounting (MC&A) program; 

(c) Organization charts, position descriptions, and qualifications records; 

(d) Health and safety compliance records, medical records, personnel exposure 

records; 

(e) Quality assurance records (see Section 12.15(1) of this appendix); 

(f) Safety inspections, audits, assessments, and investigations; and 

(g) Safety statistics and trends. 

(5) Integrated safety analysis (ISA): 

(a) ISA and revisions; and 

(b) ISA summary. 

(6) Nuclear criticality safety: 

(a) Nuclear criticality control written procedures and statistics; 

(b) Nuclear criticality safety analyses; 

(c) Records pertaining to nuclear criticality inspections, audits, investigations, and 

assessments; 

(d) Records pertaining to nuclear criticality incidents, unusual occurrences, or 

accidents; and 

(e) Records pertaining to nuclear criticality safety analyses.

NUREG-1718 1-2
Appendix I

(7) Fire protection: 

(a) Fire hazard analysis; 

(b) Fire prevention measures, including hot-work permits and firewatch records; 

(c) Records pertaining to inspection, maintenance, and testing of fire protection 

equipment; 

(d) Records pertaining to fire protection training and retraining of response 

teams; and 

(e) Prefire emergency plans. 

(8) Chemical safety: 

(a) Chemical process safety procedures and plans; (b) Records pertaining to chemical process inspections, audits, investigations, 

and assessments; 

(c) Diagrams, charts, and drawings; 

(d) Records pertaining to chemical process incidents, unusual occurrences, or 

accidents; 

(e) Chemical process safety reports and analyses; and 

(f) Chemical process safety training. 

(9) Radiation safety: 

(a) Bioassay data; 

(b) Exposure records; 

(c) Radiation protection (and contamination control) records; 

(d) Radiation training records; and 

(e) Radiation work permits. 

(10) Environmental protection: 

(a) Environmental release and monitoring records; and 

(b) Environmental report and supplements to the environmental report, as 

applicable. 

(11) Plant systems: 

(a) Written procedures and statistics for plant systems; 

(b) Safety analyses and management measures for plant systems; 

(c) Records pertaining to inspections, audits, investigations, and assessments of 

plant systems; and 

(d) Records pertaining to a description of equipment and facilities design 

(electrical systems, structures and components, cooling water systems, 

containment and confinement systems, ventilation systems, etc.).

1-3 NUREG-1718
Appendix I

(12) Human factors: 

(a) Personnel performance trend analyses; and 

(b) Human factor improvements. 

(13) Security and safeguards: 

(a) Physical protection plans; 

(b) Fundamental nuclear material control plans; 

(c) Transportation plans; 

(d) Records pertaining to granting unescorted access; and 

(e) Records pertaining to material control and accounting of special nuclear 

material. 

(14) Emergency protection: 

(a) Emergency plan(s) and procedures; 

(b) Comments on emergency plan(s) from outside emergency response 

organizations; 

(c) Emergency drill records; 

(d) Memorandum of understanding with outside emergency response 

organizations; 

(e) Records of actual events; 

(f) Records pertaining to the training and retraining of personnel involved in 

emergency preparedness functions; and 

(g) Records pertaining to the inspection and maintenance of emergency 

response equipment and supplies. 

(15) Management measures: 

(a) Quality assurance: 

(i) Table 1 in the reference listed second in Section 13 contains a list of QA 

records generated during design and construction of a nuclear power 

plant that should be maintained as QA records. Although that reference 

was developed for nuclear power plants, the QA recordkeeping 

requirements for the design and construction of this facility should be 

comparable in many areas; and 

(ii) Appendix A of the reference listed third in Section 13 contains a list of 

typical procedures for the operation of nuclear power plants. Although 

that reference was developed for nuclear power plants, the QA procedure 

requirements for the operation of this facility should be comparable in 

many areas.

NUREG-1718 1-4
Appendix I

(b) Configuration management: 

(i) Safety analyses, reports, and assessments that support the physical 

configuration of process designs and changes to those designs; (ii) Validation records for computer software used for safety analysis or 

MC&A; 

(iii) ISA documents including facility drawings, specifications, and purchase 

specifications for IROFS; and 

(iv) Approved, current operating procedures and emergency operating 

procedures. 

(c) Maintenance: 

(i) Preventive maintenance records, including trending and root cause 

analysis; 

(ii) Calibration and testing data for IROFS; and 

(iii) Corrective maintenance records. 

(d) Training and qualification of plant personnel: 

(i) Personnel training and qualification records; and 

(ii) Procedures. 

(e) Plant procedures: 

(i) Standard operating procedures; and 

(ii) Functional test procedures. 

(f) Audits and assessments: 

(i) Audits of safety and environmental activities; and 

(ii) Assessments of safety and environmental activities. 

(g) Incident investigations: 

(i) Investigation reports; 

(ii) Discussion of how and when changes recommended by investigation 

reports are implemented; 

(iii) Summary of reportable events for the term of the license; and 

(iv) Incident investigation policy. 

(h) Records management: 

(i) Policy;

1-5 NUREG-1718
Appendix I

(ii) Material storage records; and 

(iii) Records of receipt, transfer, and disposal of radioactive material. 

13. References 

Nuclear Regulatory Commission (U.S.), Washington, DC. "Domestic Licensing of Special 

Nuclear Material, (10 CFR Part 70)." Federal Register: Vol. 64, No. 146. pp. 41338

31357. July 30, 1999. 

Nuclear Regulatory Commission (U.S.)"(NRC). Regulatory Guide 1.28, Rev. 3, "Quality 

Assurance Program Requirements (Design and Construction)." NRC: Washington, D.C. 

August 1985. 

. Regulatory Guide 1.33, Rev. 2, "Quality Assurance Program Requirements 

(Operation)." NRC: Washington, D.C. February 1978.

NUREG-1718 1-6
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER 

(2-89) (Assigned by NRC, Add Vol., Supp., Rev., NRCM 1102. and Addendum Numbers, if any.) 3201.3202 BIBLIOGRAPHIC DATA SHEET 

(See instructions on the reverse) 

2. TITLE AND SUBTITLE NUREG-1718 

Standard Review Plan for the Review of an Application for a Mixed Oxide (MOX) Fuel 

Fabrication Facility 3. DATE REPORT PUBLISHED 

MONTH YEAR 

Final Report August 2000 

4. FIN OR GRANT NUMBER 

5. AUTHOR(S) 6. TYPE OF REPORT 

Enrichment Section Final 

Special Projects Branch Final 

7. PERIOD COVERED (inclusive Dates) 

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC, provide Division, Office or Region, U.S. Nuclear Regulatory Commission, and mailing address; if contractor, 

provide name and mailing address.) 

Division of Fuel Cycle Safety and Safeguards 

Office of Nuclear Material Safety and Safeguards 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-001

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (If NRC, type 'Same as above', if contractor, provide NRC Division, Office or Region, U.S. Nuclear Regulatory Commission, 

and mailing address.)

Same as 8. above.

10. SUPPLEMENTARY NOTES 

11. ABSTRACT (200 words or less) 

This NUREG provides guidance to the NRC staff reviewers in the Office of Nuclear Material Safety and Safeguards who will 

perform safety and environmental impact reviews of the anticipated application for construction approval and license application 

for operations for the Mixed Oxide (MOX) Fuel Fabrication Facility under 10 CFR Part 70 specifically as related to plutonium 

processing and fuel fabrication. The standard review plan (SRP) presented in this NUREG ensures the quality, uniformity, 

stability, and predictability of the staff reviews. It presents a defined basis from which to evaluate proposed changes in the 

scope and requirements of the staff reviews. The SRP makes information about NRC acceptance criteria widely available to 

interested members of the public and the regulated industry. Each SRP section addresses the responsibilities of the persons 

performing the review, the review areas, the Commission's regulations pertinent to specific technical matters, the acceptance 

criteria used by the staff, how the review is accomplished, and the conclusions that are appropriate for the Safety Evaluation 

Report (SER).

12. KEY WORDS/DESCRIPTORS (List words or phrases that will assist researchers in locating the report.) 13. AVAILABILITY STATEMENT 

mixed oxide fuel fabrication plants unlimited 

mixed oxide fuels 14. SECURITY CLASSIFICATION 

plutonium (This Page) 

unclassified 

(This Report) 

unclassified 

15. NUMBER OF PAGES 

16. PRICE

NRC FORM 335 (2-89) This form was electronically produced by Elite Federal Forms, Inc.
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