Design Control

In Pursuit of Engineering Excellence

A Quick Reference Guide 

for NRC Inspectors

Region III
“All men by nature desire knowledge.”

Aristotle

“The important thing is not to stop questioning.

Curiosity has its own reason for existing.”

Albert Einstein
i

FOREWORD

Knowledge management and transfer have become increasingly important as the U.S. 

Nuclear Regulatory Commission seeks to share the vast inspection knowledge of its 

experienced inspectors with those who have recently joined the agency. This booklet 

should serve as an inspection reference to further increase the understanding by, and 

development of, NRC inspectors who perform inspections in the areas covered by this 

booklet.

The contents of this design control quick reference booklet resulted from a collaboration 

of all four regional offices in support of the agency’s focus on knowledge transfer, with 

staff from various other NRC offices providing valuable comments and suggestions. The 

principal and contributing authors of this booklet are listed below.

Julio Lara, P.E., Region III 

Jennifer Tifft, Region I 

Frank Arner, Region I 

Randy Moore, Region II 

George Replogle, Region IV

Note that the guidance contained in this booklet is not intended to be all inclusive but 

rather to supplement existing inspection procedures, to heighten inspectors’ awareness, 

and further enhance the effectiveness of engineering inspections. The various inspection 

manual chapters and other regulatory documents discussed herein, provide the official 

agency inspection policies and guidance.

iii

Appendix B, “Quality Assurance Criteria for Nuclear 

Power Plants and Fuel Reprocessing Plants,” to 

Title 10, Part 50, “Domestic Licensing of Production and 

Utilization Facilities,” of the Code of Federal Regulations 

(10 CFR Part 50) 

Criterion III, Design Control

Measures shall be established to assure that applicable regulatory requirements and the 

design basis, as defined in 10 CFR 50.2 and as specified in the licensee application, for 

those structures, systems, and components to which this Appendix applies are correctly 

translated into specifications, drawings, procedures, and instructions. These measures 

shall include revisions to assure that appropriate quality standards are specified and 

included in design documents and that deviations from such standards are controlled. 

Measures shall also be established for the selection and review for suitability of application 

of materials, parts, equipment, and processes that are essential to the safety-related 

functions of the structures, systems, and components.

Measures shall be established for the identification and control of design interfaces and 

for coordination among participating design organizations. These measures shall include 

the establishment of procedures among participating design organizations for the review, 

approval, release, distribution, and revision of documents involving design interfaces.

The design control measures shall provide for verifying or checking the adequacy of 

design, such as by the performance of design reviews, by the use of alternate or simplified 

calculational methods, or by the performance of a suitable testing program. The verifying 

or checking process shall be performed by individuals or groups other than those who 

performed the original design, but who may be from the same organization. Where a 

test program is used to verify the adequacy of a specific design feature in lieu of other 

verifying or checking processes, it shall include suitable qualification testing of a prototype 

unit under the most adverse design conditions. Design control measures shall be applied 

to items such as the following: reactor physics, stress, thermal hydraulic, and accident 

analysis; compatibility of materials; accessibility for inservice inspection, maintenance and 

repair; the delineation of acceptance criteria for inspections and tests.

Design changes, including field changes, shall be subject to design control measures 

commensurate with those applied to the original design and be approved by the 

organization that performed the original design unless the applicant designates another 

responsible organization.

v

CONTENTS

FOREWORD ..................................................................................................................................i

PURPOSE.................................................................................................................................... 1

DEFINITIONS............................................................................................................................... 2

GUIDANCE FOR ENGINEERING INSPECTIONS ...................................................................... 5

A. NRC Inspections Involving Engineering Design Activities.............................................. 5

B. 10 CFR 50.55a, “Codes and Standards”........................................................................ 5

C. Standard Review Plans.................................................................................................. 7

D. Systematic Evaluation Program..................................................................................... 8

E. Safety Evaluation Reports.............................................................................................. 9

F. Technical Guidance—Manual Chapter Part 9900.......................................................... 9

DESIGN AND LICENSING BASIS ..............................................................................................11

A. Design Basis (Bases)....................................................................................................11

B. Licensing Basis ............................................................................................................ 12

DESIGN CONTROL GUIDANCE DOCUMENTS....................................................................... 15

CRITERION III OF APPENDIX B TO 10 CFR PART 50 CROSS-REFERENCE TABLE............ 19

PLANT ENGINEERING PRODUCTS......................................................................................... 21

A. Overview ...................................................................................................................... 21

B. Calculations.................................................................................................................. 21

C. Modifications and Temporary Modifications ................................................................. 24

D. Operability Determinations........................................................................................... 25

E. Procedures................................................................................................................... 26

F. Corrective Action Determinations................................................................................. 27

G. Dedication of Commercial-Grade Parts for Safety-Related Applications ..................... 27

ROBUST APPLICATION OF ENGINEERING PRINCIPLES...................................................... 29

VALUE-ADDED FINDINGS AND STARS DOCUMENTS........................................................... 31

REFERENCES........................................................................................................................... 33

ACRONYMS............................................................................................................................... 37

1

PURPOSE

This booklet was developed primarily for new U.S. Nuclear Regulatory Commission (NRC) 

inspectors who conduct engineering-focused inspections. This booklet may also serve as a quick 

reference for other experienced inspectors. In user-friendly language, it provides inspection 

guidance and contains useful inspection tips. 

The main purpose of the booklet is to develop a fundamental understanding of what constitutes 

design control, as required by Criterion III, “Design Control,” of Appendix B, “Quality Assurance 

Criteria for Nuclear Power Plants and Fuel Reprocessing Plants,” to Title 10, Part 50, “Domestic 

Licensing of Production and Utilization Facilities,” of the Code of Federal Regulations (10 CFR 

Part 50). The booklet structure begins with basic terminology used in the nuclear industry to 

ensure a common understanding. NRC engineering inspectors frequently rely on industry codes 

and standards and on NRC guidance documents. The booklet discusses the importance of 

these documents and how inspectors can use them. Because inspectors must develop a sound 

understanding of design and licensing basis documents, these documents are introduced. After 

touching on engineering terminology, codes and standards, and design and licensing basis, the 

booklet then delves into Criterion III of Appendix B to 10 CFR Part 50 to help inspectors understand 

the various requirements of the criterion and into NRC and industry guidance documents regarding 

the various elements of design control. Licensee engineering design programs and activities 

should ensure compliance with the requirements in Criterion III with respect to engineering work 

products, such as calculations, plant modifications, engineering analysis, and procurement and 

dedication. The booklet discusses these various engineering work products to enhance inspectors’ 

knowledge and understanding. Lastly, the booklet briefly explores the meaning of the commonly 

used term “robust application of engineering principles.” 
2

DEFINITIONS

The definitions below came from various sources, including NRC regulations and guidance 

documents and industry standards, and reflect generally accepted engineering terminology. 

However, because a number of these terms are not explicitly defined in NRC regulations, some 

variations may exist. Documents used to develop this guidance booklet are listed in the References 

section.

Assumptions

Assumptions are statements that define premises, limitations, or restrictions that are accepted as 

true without a thorough demonstration.

Commercial-Grade Dedication

Commercial-grade dedication is an acceptance process that provides reasonable assurance 

that a commercial-grade item designated for use as a basic component will perform its intended 

safety function and, in this respect, is deemed equivalent to an item designed and manufactured 

under a 10 CFR Part 50, Appendix B, Quality Assurance Program. This assurance is achieved 

when a third-party dedicating entity or the purchaser of a commercial-grade item identifies the 

critical characteristics of the item and verifies the acceptability of these characteristics through 

inspections, tests, or analyses.

Configuration Management

Configuration management is an integrated management process used to ensure that the licensee 

maintains the plant’s physical and functional characteristics in conformance with its design and 

licensing basis; that operating, training, modification, and maintenance processes are consistent 

with the conditions prescribed by the design and current licensing basis; and that the licensee 

operates and maintains the plant within these conditions. 

Critical Characteristics

Critical characteristics are those important design, material, and performance characteristics of 

a commercial-grade item that, once verified, will provide reasonable assurance that the item will 

perform its intended safety function(s).

Defense in Depth

Defense in depth is a design and operational philosophy for nuclear facilities that calls for multiple 

layers of protection to prevent accidents and to mitigate the effects of accidents. Defense in depth 

includes the use of controls, multiple physical barriers to prevent release of radiation, redundant 

and diverse system functions, and emergency response measures. 

Degraded Condition

A degraded condition is one in which the qualification of a structure, system, or component (SSC) 

or its functional capability is reduced. Examples of degraded conditions are failures, malfunctions, 

deficiencies, deviations, and defective material and equipment. Examples of conditions that 

can reduce the capability of a system are aging, erosion, corrosion, improper operation, and 

maintenance.

Definitions
3

Design Bases

Design bases is the information that identifies the specific functions that an SSC is to perform and 

the specific values or ranges of values chosen for controlling parameters as reference bounds 

for design. These values may be (1) restraints derived from generally accepted “state-of-theart” practices for achieving functional goals or (2) requirements derived from analysis (based on 

calculations and/or experiments) of the effects of a postulated accident for which an SSC must 

meet its functional goals. 

Design Change

A design change is a change to a final design that affects the performance of an SSC.

Design Input

Design input includes those criteria, parameters, bases, or other design information upon which 

the final design is based. This can include technical information, design bases, performance 

criteria, regulatory requirements, codes, standards, analysis, and calculations.

Design Output

Design output includes documents such as drawings, specifications, and other documents that 

define the technical requirements of SSCs. 

Design Review

Design review refers to independent verification that ensures that the licensee has incorporated 

important design, material, and performance characteristics into the plant design to ensure that 

it provides or meets safety functions while providing layers of protection to prevent and mitigate 

accidents.

Engineering Judgment

An engineering judgment is a determination based on prior examples, experience, or observation 

that has not been subjected to rigorous engineering validation. 

Nonconforming Condition

A nonconforming condition is a condition of an SSC that involves a failure to meet the current 

licensing basis (CLB) or a situation in which the quality of an SSC is reduced because of factors 

such as improper design, testing, construction, or modification.

Operable

An SSC shall be operable or have operability when it is capable of performing its specified function. 

This includes adequate performance of any support instrumentation, controls, electrical power, 

cooling or seal water, lubrication or auxiliary equipment.

Part 21 (10 CFR Part 21) 

In 10 CFR Part 21, “Reporting of Defects and Noncompliance,” the NRC requires a responsible 

officer of a firm that supplies, or a facility that receives, non-complying parts and components, or 

defective components that cause a significant safety hazard, to immediately notify the Commission. 

Qualification

Definitions
4

Qualification refers to documented evidence that shows that a facility conforms to all aspects of 

design basis, including codes, standards, design criteria, and Quality Assurance (QA) regulations.

Quality Assurance

Quality Assurance comprises all those planned and systematic actions necessary to provide 

adequate confidence that an SSC will perform satisfactorily in service. Attributes of a QA program 

include programs that preserve quality through procedures, recordkeeping, inspections, corrective 

actions, and audits. 

Redundancy

Redundancy is an alternate, independent, or duplicate method of fulfilling a safety function to 

mitigate the consequences of a design-basis accident.

Safety-Related Function 

A safety-related function applies to the SSCs, procedures, and controls of a facility or process 

that must remain functional during and following design-basis events to ensure the integrity of 

the facility’s reactor coolant pressure boundary, the facility’s capability to shut down the reactor 

and maintain it in a safe shutdown condition, or the facility’s capability to prevent or mitigate 

the consequences of accidents that could result in potential offsite exposure comparable to the 

guidelines in 10 CFR 50.34(a)(1), 10 CFR 50.67(b)(2), or 10 CFR 100.11, “Determination of 

Exclusion Area, Low Population Zone, and Population Center Distance.” An example of a safetyrelated function is a facility’s capability to shut down a nuclear reactor and maintain it in a safe 

shutdown condition. 

Single Failure

A single failure is an occurrence that results in the loss of a component’s capability to perform its 

intended safety functions. Multiple failures resulting from a single occurrence are considered to 

be a single failure. Fluid and electric systems are considered to be designed against an assumed 

single failure if neither of the following failures result in a loss of the system’s capability to perform 

its safety functions:

(1) a single failure of any active component (assuming that passive components function 

properly)

(2) a single failure of a passive component (assuming that active components function properly)

Structures, Systems, and Components (safety-related)

Nuclear power plants are designed with SSCs that prevent or mitigate the consequences of 

postulated accidents which could cause undue risk to the health and safety of the public. See 

also Safety-Related Function.

Verification

Verification refers to the process of checking that the information contained in design-basis 

documents has been correctly and consistently translated from the source documents. 

Definitions
5

GUIDANCE FOR ENGINEERING INSPECTIONS

A. NRC Inspections Involving Engineering Design Activities 

The Reactor Oversight Process (ROP) includes several engineering-focused inspections. 

Engineering remains an important focal point of certain inspections because plant performance 

indicators do not always capture design inadequacies. Additionally, engineering can be a difficult 

area to inspect as design margins can often be compromised without this being readily apparent. 

Engineering-focused inspections allow inspectors to review equipment design and operation to 

verify that normal operation and routine surveillance testing assure equipment functionality under 

worst-case accident conditions. 

Engineering-focused inspections rely on a review of a licensee’s engineering work activities 

and products and remain an important part of the ROP. In the early to mid-1990s, the NRC’s 

findings during inspections and reviews identified broad programmatic weaknesses that were 

the result of design and configuration deficiencies at some plants. These deficiencies impacted 

the operability of required equipment and indicated discrepancies between a plant’s Updated 

Final Safety Analysis Report (UFSAR) and the as-built or as-modified plant or plant operating 

procedures. The current ROP engineering inspections allow the NRC to monitor the licensee’s 

continued effectiveness in maintaining configuration control. To perform effective inspections, 

NRC inspectors need to develop and maintain a fundamental knowledge of requisite engineering 

quality standards and practices.

In summary, the overall goal of engineering inspections is to verify through a review of engineering 

work activities and products that the plant design and analyses, including calculations of record, 

reflect the licensing basis such as the UFSAR, Technical Specifications (TS), and other licensing 

documents. Furthermore, these engineering work products should be consistent with the physical 

plant equipment and how this equipment is operated and maintained. Consistency among 

the engineering and licensing basis documents, when complemented with well-operated and 

-maintained equipment, results in sound configuration management at a licensed facility.

B. 10 CFR 50.55a, “Codes and Standards” 

Codes and standards are an integral part of the NRC’s regulatory process because they provide 

detailed requirements and guidance to implement the NRC’s broad, general design criteria 

prescribed in 10 CFR 50, Appendix A, “General Design Criteria for Nuclear Power Plants.” The 

NRC and nuclear industry use codes and standards to provide greater assurance of safe plant 

design and operations. The NRC endorses codes and standards through Regulatory Guides 

(RGs) (see References Section) and in regulations. In 10 CFR 50.55a, “Codes and Standards,” 

the NRC endorses Sections III and XI of the American Society of Mechanical Engineers (ASME) 

Boiler and Pressure Vessel (BPV) Code and the Institute of Electrical and Electronics Engineers 

(IEEE) Standards 279, “Criteria for Protection Systems for Nuclear Power Generating Stations,” 

and 603, “Criteria for Safety Systems for Nuclear Power Generating Stations.”

Guidance for Engineering Inspections
6

ASME Code

The ASME BPV Code contains 11 sections (I–XI), two of which are nuclear sections. The nuclear 

sections are Section III, which contains rules for the design and fabrication of nuclear power plant 

components, and Section XI, which deals with the inservice inspection (e.g., maintenance of 

nuclear power plant components) and inservice testing 

of nuclear plant components. Both of these industry 

standard sections constitute regulatory requirements 

under 10 CFR 50.55a. 

The ASME Code (also referred to as the “BPV Code” 

or simply “the Code”) is periodically revised to update 

the provisions for design, construction, and inservice inspection of pressure boundary components. 

Inspectors who perform inspections in these areas 

should identify which version of the Code licensees 

have committed to implement as part of the inspection 

efforts.

The BPV Code also contains “Code Cases.” A Code 

case provides an alternative to the BPV Code and 

is specifically written as an alternative to a specific 

paragraph of the Code. The NRC’s endorsement 

of these Code Cases is documented in RG 1.147, 

“Inservice Inspection Code Case Acceptability, ASME 

Section XI, Division 1,” whereas Code Cases that the 

NRC has not endorsed are described in RG 1.193, 

“ASME Code Cases Not Approved for Use.”

Code interpretations are also contained within the BPV Code and differ from Code cases. These 

interpretations are clarifications to the BPV Code. However, these interpretations are part of 

neither the NRC’s regulations nor its endorsed RGs, and hence the NRC is not bound by such.

Within the ASME BPV Code sections, three Code classes pertain to nuclear power plant 

components—Classes 1, 2, and 3. RG 1.26, “Quality Group Classifications and Standards for 

Water-, Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants,” 

describes a quality classification system related to the ASME standard that may be used to 

determine quality standards acceptable to the NRC staff for satisfying requirements for safetyrelated components in light-water-cooled nuclear power plants. Table 1 defines the three Code 

classes. 

A separate subsection in Section XI describes the periodic inspection and repair requirements for 

each Code class. Many plants were built before the first issuance of Section XI. Thus, the design 

of the plant may prevent inspectors from examining specific component locations. Therefore, 10 

CFR 50.55a allows the licensee to request relief from those Section XI Code requirements that 

they cannot meet. The licensee must typically justify why the examination is limited and provide 

an alternative examination that affords a comparable level of assurance.

During access to contaminated and high 

radiation areas, personal safety must also 

be a priority.

Guidance for Engineering Inspections

rg t, _ 

1 ! 

~ 
7

Table 1. Code Classes for Nuclear Power Plant Components

Class Discussion

1 Class 1 (Code category NB) components consist of the pressure-retaining 

boundary for the reactor coolant system, which generally includes all reactor 

coolant system connections to the outermost containment isolation valve 

or to the second of two normally shut valves. In 10 CFR 50.55a(c)(1), 

the NRC specifies that components within these boundaries must meet 

Class 1 requirements.

2 Class 2 (Code category NC) components consist of the pressure-retaining 

boundary for emergency core cooling systems (ECCS) relied on to mitigate 

an accident.

3 Class 3 (Code category ND) components consist of the pressure-retaining 

portion of cooling water systems that support the reactor shutdown function, 

the ECCS function, residual heat removal system functions, or spent fuel pool 

cooling functions.

IEEE Standards 

IEEE 279 and IEEE 603 are incorporated in 10 CFR 50.55a(h). Both of these industry standards 

provide requirements for reactor protection systems. As with other industry standards, most IEEE 

standards are not explicitly endorsed through regulations but rather through RGs. The NRC 

issues RGs to describe methods that the staff considers acceptable for use in implementing 

specific parts of the agency’s regulations, to explain techniques that the staff uses in evaluating 

specific problems or postulated accidents, and to provide guidance to applicants and licensees. 

RGs are not substitutes for regulations, and compliance with RGs is not required, unless explicitly 

incorporated into the facility’s operating license (see References Section for listing of commonly 

applied RGs). 

For nuclear power plants with construction permits issued after January 1, 1971, but before May 

13, 1999, reactor protection systems must meet the requirements stated in either IEEE-279 or 

IEEE-603. For nuclear power plants with construction permits issued before January 1, 1971, 

reactor protection systems must be consistent with their licensing basis. Inspectors should review 

the licensee’s licensing-basis documents to identify the specific commitments and requirements. 

C. Standard Review Plans 

In 1975, the NRC issued a Standard Review Plan (SRP) to define the scope of review and 

acceptance criteria for the NRC’s approval of safety analysis reports. The NRC has a number 

of SRPs for staff use in reviewing proposed licensing actions. These actions may relate to the 

construction or operation of a nuclear facility or to the possession or use of nuclear materials. For 

nuclear power plants, the NRC has a comprehensive SRP (see NUREG-0800 at http://www.nrc.

gov/reading-rm/doc-collections/nuregs/staff/sr0800/).

The Office of Nuclear Reactor Regulation (NRR) and the Office of New Reactors (NRO) are the 

primary users of SRPs; however, inspectors may find them useful when preparing for an inspection. 

SRPs can give inspectors insights into the standard that NRR uses to evaluate licensing actions. 

Guidance for Engineering Inspections
8

Inspectors should keep in mind that SRPs are not regulations, and therefore they cannot propose 

enforcement actions if licensees do not meet the regulatory and safety standard discussed in 

the SRP. SRPs are also not inspection procedures. Inspectors can use SRPs as informational 

tools, but they should perform the inspection in accordance with the guidance contained in the 

applicable inspection procedure. 

D. Systematic Evaluation Program

In 1977, the NRC staff initiated the Systematic Evaluation Program (SEP) to review the designs of 

older operating nuclear power plants. These plants were licensed before the NRC issued its SRP 

in 1975. The SEP was divided into two phases. In Phase I, the staff defined 137 issues for which 

regulatory requirements had changed over time to warrant an evaluation of those plants licensed 

before the issuance of the SRP. In Phase II, the staff compared the design of a select number of 

the older plants to the SRP issued in 1975. Based on these reviews, the staff identified 27 of the 

original 137 issues that required some corrective action at one or more of the SEP plants that 

were reviewed. The staff referred to the issues on this smaller list as the SEP “lessons-learned” 

issues and concluded that the older operating plants that were not in the group of SEP plants 

examined (i.e., non-SEP plants) could improve their safety operations by taking corrective actions 

for these 27 issues. Therefore, the NRC staff concluded that inspectors should consider these 27 

issues at the non-SEP plants to determine whether an adequate level of safety existed at these 

plants. 

During inspections at the plants licensed before 1975, inspectors may find that the licensee’s 

licensing basis is complex, and therefore, they may need to review numerous licensing documents 

to thoroughly evaluate the acceptability of particular issues of concern. Table 2 lists all current 

operating facilities reviewed as part of the SEP and those licensed before 1975. 

Guidance for Engineering Inspections

Table 2. SEP and Pre-1975 Plants

SEP Plants Pre-1975 Plants (non-SEP Plants)

Palisades Arkansas 1 Monticello

Ginna Browns Ferry 1/2 Nine Mile Point 1

Oyster Creek Brunswick 2 Oconee 1/2/3

Dresden 2 Calvert Cliffs 1 Peach Bottom 2/3

Cook 1 Pilgrim

Cooper Point Beach 1/2

Dresden 3 Prairie Island 1/2

Duane Arnold Quad Cities 1/2

FitzPatrick Robinson 2

Fort Calhoun Surry 1/2

Hatch 1 Three Mile Island 1

Indian Point 2/3 Turkey Point 3/4

Kewaunee Vermont Yankee

Millstone 2
9

Generic Letter (GL) 95-04, “Final Disposition of the Systematic Evaluation Program LessonsLearned Issues,” discusses the final disposition of the lessons-learned issues found in the SEP. 

References in GL 95-04 provide additional background information about the SEP.

E. Safety Evaluation Reports 

Safety evaluation reports (SERs) document the results of NRR reviews of proposed licensing 

actions, such as initial licensing reviews and subsequent license amendments. SERs are part of 

the licensing basis of the plant and are useful for inspectors to review so that the full scope of the 

plant’s licensing basis is understood. SERs give insights into technical issues that a licensee has 

already resolved and can help inspectors focus the inspection.

F. Technical Guidance—Manual Chapter Part 9900 

Part 9900 of the NRC Inspection Manual contains technical guidance on various areas of interest 

to inspectors ranging from technical to licensing issues. This guidance was developed in specific 

areas to address NRC staff or industry questions in those areas.

As an example, technical guidance exists 

regarding Sections III and XI of the ASME BVP 

Code. Specifically, Code interpretations, use 

of engineering judgment, and flaw evaluations 

are discussed. Other notable areas where 

technical guidance is available include other 

Code-related issues, equipment repair issues, 

notices of enforcement discretion (NOEDs), 

TS guidance, treatment of degraded or 

nonconforming conditions, and equipment 

testing interpretation issues.

Typically, engineering-focused inspectors use 

the guidance on the treatment of degraded 

or nonconforming conditions along with the 

testing and Code guidance.

Guidance for Engineering Inspections

Field inspections and verifications are an important 

element of the NRC inspection program.

11

DESIGN AND LICENSING BASIS

A. Design Basis (Bases)

Inspectors should recognize that the design 

basis is a subset of the licensing basis.

RG 1.186, “Guidance and Examples for 

Identifying 10 CFR 50.2 Design Bases”

RG 1.186 discusses the history of inadequate 

maintenance of design-basis documentation 

(DBD) by licensees and addresses the NRC’s 

initiative requesting licensees to reconstitute 

their design-base information. The guide 

also endorses Appendix B, “Guidance and 

Examples for Identifying 10 CFR 50.2 Design 

Bases,” to Nuclear Energy Institute (NEI) 97-

04, “Design Bases Program Guidelines.” 

NEI 97-04 Appendix B, Guidance and Examples for Identifying 10 CFR 50.2 Design Bases

This document clarifies and expands the definition of design bases and discusses the relationship 

of the design bases to other 10 CFR requirements. It clarifies the definition of design bases to 

include those bounding conditions under which SSCs must perform design basis functions.

It further discusses how a facility’s design bases can change. The design bases for a plant can 

change as a result of new NRC requirements after the approval of the original operating license 

and from changes to ensure compliance with NRC requirements. NEI 97-04 can be obtained from 

the NRC’s Agencywide Documents Access and Management System (ADAMS Accession No. 

ML003771698).

Final and Updated Safety Analysis Reports

Final safety analysis reports (FSARs) and UFSARs can contain design- and licensing-basis 

information and describe the licensee’s implementation of the general design criteria in Appendix A 

to 10 CFR Part 50. A license application originally includes the plant’s Preliminary Safety Analysis 

Report. During the licensing stage, the licensee submits an FSAR. The licensee must describe 

the plant’s individual safety systems in the FSAR in sufficient detail to allow the NRR reviewer 

to prepare an SER to support the granting of an operating license. As a result of routine updates 

after the granting of the license, the FSAR is often referred to as an UFSAR (or USAR). The 

UFSAR is both a licensing-basis document and a design-basis document because it describes 

how the plant was designed, constructed, maintained, and operated, but it does not by itself 

contain the regulatory requirements. The UFSAR is unique to each nuclear power plant station 

and, in the case of multiple-unit sites, to each unit of a station. UFSARs vary in the amount of 

detail they contain, which mostly depends on the age of the plant. Older plants may have as few 

as four volumes, whereas newer plants have many more. 

Design Basis

Design basis is the information that identifies 

the specific functions that a facility’s SSC 

is to perform and the specific values or 

ranges of values chosen for controlling 

parameters as reference bounds for 

design. These values may be (1) restraints 

derived from generally accepted “state-ofthe-art” practices for achieving functional 

goals or (2) requirements derived from 

analysis (based on calculations and/or 

experiments) of the effects of a postulated 

accident for which an SSC must meet its 

functional goals.

Design and Licensing Basis
12

Suggested Sources for Design-Bases 

Information

• UFSARs

• DBD

• system descriptions

• design calculations

• design analyses

• piping and instrumentation drawings

• significant design drawings

• significant surveillance procedures

• preoperational test documents

• vendor manuals

Locating Design-Basis Information

At the site, the UFSAR and the station DBD or 

system descriptions are good starting points 

for locating design-basis information. Because 

of the design-base reconstitution effort, most nuclear stations have a system DBD or System 

Description document for each safety-related system. Topical DBDs will address program areas 

such as fire protection and station blackout. These documents will address the design bases 

requirements for a system, describe components, often state actual limiting parameter values, 

and provide references that support design-basis information such as calculations, analyses, and 

testing documentation that support the design bases for that system and its components. The 

DBD usually lists licensing commitments related to the subject system or topical program.

B. Licensing Basis 

A facility’s licensing basis is also referred to as the current licensing basis (CLB). The concept of 

the CLB was first introduced in the regulations through 10 CFR 50.54(f), as a result of agency 

considerations in the development of the NRC’s backfit rule. The CLB again became an issue in 

the NRC’s decisions on extending the licenses for plants beyond the original design life, and it is 

defined in 10 CFR 54.3, “Definitions.” Although established in 10 CFR Part 54, “Requirements 

for Renewal of Operating Licenses for Nuclear Power Plants,” the definition represents the NRC’s 

understanding of the scope of the CLB and generally applies to all reactor licensees.

10 CFR 54.3(a)

The current licensing basis (CLB) is the set of NRC requirements applicable to a specific 

plant and a licensee’s written commitments for ensuring compliance with and operation 

within applicable NRC requirements and the plant-specific design basis (including all 

modifications and additions to such commitments over the life of the license) that are 

docketed and in effect. The CLB includes the NRC regulations contained in 10 CFR Parts 

2, 19, 20, 21, 26, 30, 40, 50, 51, 54, 55, 70, 72, 73, 100, and appendices thereto; orders; 

license conditions; exemptions; and technical specifications. It also includes the plantspecific design-basis information defined in 10 CFR 50.2 as documented in the most 

recent FSAR, as required by 10 CFR 50.71 and the licensee’s commitments remaining in 

effect that were made in docketed licensing correspondence such as licensee responses to 

Validation of engineering judgment and 

assumptions can provide insights into the 

adequacy of engineering products.

Design and Licensing Basis
13

NRC bulletins, generic letters, and enforcement actions, as well as licensee commitments 

documented in NRC safety evaluations or licensee event reports.

Facility Operating License

A facility’s operating license describes the conditions and requirements for the operation of the 

nuclear unit, including the receipt, use, and possession of special nuclear material and byproducts. 

It also states the maximum authorized power level for power operation. The license includes 

appendices, such as the TS.

Appendix A to 10 CFR Part 50 

Appendix A, “General Design Criteria for Nuclear Power Plants,” to 10 CFR Part 50, establishes 

the regulatory requirements for the design of nuclear power plants. The criteria are stated generally 

and grouped in the following categories:

I. Overall Requirements

II. Protection by Multiple Fission Product Barriers

III. Protection and Reactivity Control Systems

IV. Fluid Systems

V. Reactor Containment

VI. Fuel and Radioactivity Control

Because of the age of some licensed facilities, inspectors should verify that Appendix A to 10 CFR 

Part 50 applies to the facility being inspected.

Locating Current Licensing-Basis Information

Although licensing-basis information can sometimes 

be found in ADAMS, the word searches needed to 

find specific documentation can be time consuming 

and challenging. However, the licensee does have a 

collection of station-specific licensing-basis information 

in its database and should be able to provide inspectors 

with this information more readily than would a word 

search of ADAMS. The station DBDs or system 

descriptions usually list licensing-basis commitment 

references applicable to the subject system or topical 

program. The UFSAR is readily available and is also a 

source of some, but not all, licensing commitments.

The following are suggested sources of licensing-basis 

documentation:

• NRC Title 10 Federal regulations 

• plants’ TS

• UFSARs

• NRC SERs

Design and Licensing Basis

Safety is paramount when conducting field 

walkdowns near high energy or hazardous 

plant areas. 
14

• licensees’ responses to GLs 

• licensees’ responses to notices of violation

• licensee event reports 

• technical requirements manuals

• TS bases

• organizational topical reports

• QA plans

• fire protection reports

• offsite dose calculations

• physical security plans

• radiological emergency plans

• Core Operating Limits Report

• Pressure and Temperature Limits Report

There are various regulatory processes that include options for licensees to make changes to 

their licensing basis. These include 10 CFR 50.59, “Changes, Tests, and Experiments,” and 

operating license conditions such as fire protection, quality assurance, emergency preparedness, 

and physical security plans; regulatory exemption requests; and license amendments under 10 

CFR 50.90, “Application for Amendment of License, Construction Permit, or Early Site Permit.” 

In 10 CFR 50.59, the NRC establishes the conditions under which licensees may make changes 

to the facility or procedures and may conduct tests or experiments without prior NRC approval. 

The NRC must review and approve the proposed changes, tests, and experiments that satisfy the 

definitions and one or more of the criteria in the regulation before licensees can implement them.

Design and Licensing Basis
15

Design Control Guidance Documents

DESIGN CONTROL GUIDANCE DOCUMENTS

In 1967, the NRC’s predecessor, the Atomic Energy Commission, published for comment general 

design criteria for light-water-cooled nuclear power plants (i.e., draft Appendix A to 10 CFR Part 50). 

Appendix A to 10 CFR Part 50 requires licensees to establish a QA program, whereas Appendix 

B to 10 CFR Part 50 defines the requirements of that program. These appendices were approved 

as final rules in 1971 and 1970, respectively. Before their issuance, the NRC held meetings and 

exchanged correspondence with applicants individually to resolve issues. The staff’s positions 

eventually became RGs, which provide guidance on acceptable methods that licensees can use 

to meet regulations, but they are not regulations themselves. 

As part of the licensing process, licensees must submit a QA plan to the NRC. Compliance with 

the QA plan is required because it is an attachment to the facility’s operating license. The QA 

plan is structured to follow the requirements of Appendix B to 10 CFR Part 50, and it provides 

a general description of the established measures that licensees need to comply with Appendix 

B requirements, including Criterion III, “Design Control.” As part of the preparation for an 

engineering-focused inspection, an inspector should periodically review the licensee’s QA plan 

to better understand the licensee’s design control measures. The inspector can further review 

design control implementing procedures during the inspection because they describe in greater 

detail how the licensee implements design control measures.

The American National Standards Institute (ANSI) published ANSI N45.2, “Quality Assurance 

Requirements for Nuclear Power Plants,” to describe the requirements of Appendix B in greater 

detail. Industry standards, such as ANSI N45.2 and daughter standards (e.g., ANSI N45.2.11, 

“Quality Assurance Requirements for the Design of Nuclear Power Plants”), serve as “howto” documents for the nuclear industry to meet NRC requirements, including Criterion III. The 

NRC endorsed this standard in RG 1.28, “Quality Assurance Program Requirements (Design 

and Construction).” Inspectors should review the licensee’s QA plan, which will describe the 

established design control process, including commitments to industry standards; the SRP; safety 

analysis reports; other industry codes; regulations; and RGs to obtain a thorough understanding 

of the licensee’s design control process. ANSI (sometimes coded as ASME) standards may be 

obtained through the NRC’s electronic Technical Library. 

RG 1.28, “Quality Assurance Program Requirements (Design and Construction)” 

In RG 1.28, Revision 3, the NRC endorsed industry standard ASME NQA-1-1983, “Quality 

Assurance Program Requirements for Nuclear Facilities.” The standard mirrors, and in some 

cases provides greater guidance than that in ANSI N45.2.11 as an acceptable method that 

licensees can use to implement QA programs during the design and construction phases of a 

nuclear power plant. Inspectors should remain informed with respect to the RG position and the 

specific licensee commitments in this area. Additionally, the NRC may later endorse more recently 

issued versions of the ASME standard, as reflected in draft RGs.

ASME NQA-1-1983 has three main sections: (1) Basic Requirements, (2) Supplements, and (3) 

Appendices. The Basic Requirements section provides the basic requirements for establishing 

and executing QA programs. The Supplements section amplifies the individual requirements of 

the Basic Requirements section. The Appendices section provides non-mandatory guidance for 

meeting the Basic Requirements and Supplements sections. 
16

Design Control Guidance Documents

ANSI N45.2.11, “Quality Assurance Requirements for the Design of Nuclear Power Plants”

ANSI N45.2.11 describes the minimum QA 

requirements that licensees must implement 

during the design of nuclear power plant SSCs. 

The SSCs are those that are required to prevent 

accidents that could cause undue risk to the 

health and safety of the public or those that are 

required to mitigate the consequences of an 

accident. During the licensing phase of nuclear 

plants, most licensees commit to following the 

guidance contained in the standard. As is the case 

with all licensing actions, inspectors should verify 

actual licensee commitments through a review 

of the facility’s UFSAR and other licensing-basis 

documents. Licensees structure their QA program 

regarding design control to incorporate the 

guidance contained in the standard and to meet 

the requirements of Criterion III of Appendix B to 

10 CFR Part 50. The standard itself covers various 

elements of an effective design control program. 

A brief discussion of salient portions of ANSI 

N45.2.11 follows. The discussion summarizes the 

guidance contained within the standard. Inspectors 

should periodically review the entire standard to 

develop a comprehensive understanding of all design control standard elements.

Design Process and Input Requirements

Criterion III of Appendix B to 10 CFR Part 50 requires, in part, the following:

Measures shall be established to assure that applicable regulatory requirements and the 

design basis, as defined in 10 CFR 50.2 and as specified in the licensee application, for 

those structures, systems, and components to which this Appendix applies are correctly 

translated into specifications, drawings, procedures, and instructions.

Sections 3 and 4 of ANSI N45.2.11 provide the following guidance on this criterion:

Applicable design inputs, such as design bases, regulatory requirements, codes and 

standards, shall be identified, documented, and their selection reviewed and approved.

Design activities shall be prescribed and accomplished in accordance with procedures 

of a type sufficient to assure that applicable design inputs are correctly translated into 

specifications, drawings, procedures, or instructions.

A multi-discipline team inspection approach is 

warranted for complex plant modifications.
17

Design Control Guidance Documents

Inspectors should recognize that design input requirements include the following (but the list is 

not all inclusive):

• basic functions of SSCs

• performance requirements such as capacity, rating, and system output

• codes and standards

• design conditions, such as pressure, temperature, and voltage

• loads, such as seismic, thermal, and dynamic

• environmental conditions anticipated during operation

• operational requirements under various plant conditions

Additionally, licensees are to develop procedures that include requirements for the control of design 

analyses (calculations). Licensees must identify calculations by subject, originator, reviewer, and 

date or other means to assure calculations are retrievable. Calculations should document known 

and unverified assumptions.

Design Verification

Criterion III of Appendix B to 10 CFR Part 50 requires, in part, the following:

The design control measures shall provide for 

verifying or checking the adequacy of design, 

such as by the performance of design reviews, 

by the use of alternate or simplified calculational 

methods, or by the performance of a suitable testing 

program. The verifying or checking process shall 

be performed by individuals or groups other than 

those who performed the original design, but who 

may be from the same organization. Where a test 

program is used to verify the adequacy of a specific 

design feature in lieu of other verifying or checking 

processes, it shall include suitable qualification 

testing of a prototype unit under the most adverse 

design conditions.

Section 6 of ANSI N45.2.11 provides guidance on “verifying 

or checking the adequacy of design”:

Measures shall be established to verify the adequacy 

of design. Design verification is the process of 

reviewing, confirming, or substantiating the design 

but one or more methods to provide assurance that 

the design meets the specified design inputs.

Design verification shall be performed by any competent individuals or groups other than 

those who performed the original design but who may be from the same organization.

Walkdowns in sensitive plant areas 

may require coordination with licensee 

staff in Operations or Plant Safety.
18

Design Control Guidance Documents

Cursory supervisory reviews do not satisfy the intent of this standard. Design verification 

may vary from spot checking of calculations to actual tests in the field.

The extent of the design verification required is a function of the importance to safety of 

the item under consideration, the complexity of the design, the degree of standardization, 

the state-of-the-art, and the similarity with previously proven designs.

Acceptable verification methods include, but are not limited to, design reviews, alternate 

calculations, qualification testing. 

For design reviews, “the depth of review can range from a detailed check of the complete design 

to a limited check of such things as the design approach and the results obtained.” For alternate 

calculations, verification can be accomplished by performing simplified hand calculations. For 

qualification testing, “all pertinent operating modes shall be considered in determining the design 

conditions where it is intended that the test program confirm the adequacy of the overall design.”

Design Change Control

Criterion III of Appendix B to 10 CFR Part 50 requires, in part, the following:

Design changes, including field changes, shall be subject to design control measures 

commensurate with those applied to the original design and be approved by the 

organization that performed the original design unless the applicant designates another 

responsible organization.

Section 8 of ANSI N45.2.11 provides guidance on “design changes”:

Documented procedures shall be provided for design changes to approved design 

documents, including field changes.... The changes shall be justified and subjected to 

design control measures commensurate with those applied to the original design.
19

Criterion III of Appendix B to 10 CFR Part 50 Cross-Refence Table

CRITERION III OF APPENDIX B TO 10 CFR PART 50 

CROSS-REFERENCE TABLE

Table 3 identifies specific RGs and ANSI standards that provide regulatory guidance with respect 

to specific requirements in Criterion III of Appendix B to 10 CFR Part 50. The format in Table 3 can 

serve as a quick reference for inspectors.

Table 3. Criterion III of Appendix B to 10 CFR Part 50 Cross-Reference Table

 ANSI 

Requirement RG Standard

III.1 Measures shall be established to assure that applicable 1.28 N45.2

regulatory requirements and the design basis, as defined N45.2.10

in 10 CFR 50.2 and as specified in the licensee application, N45.2.11

for those structures, systems, and components to which this 

Appendix applies are correctly translated into specifications, 

drawings, procedures, and instructions.

III.2 These measures shall include revisions to assure that 1.26 N45.2

appropriate quality standards are specified and included in 1.28 N45.2.10

design documents and that deviations from such standards N45.2.11

are controlled. 1.29

1.75 IEEE 279

III.3 Measures shall also be established for the selection 1.28 N45.2

and review for suitability of application of materials, N45.2.10

parts, equipment, and processes that are essential to the N45.2.11

safety-related functions of the structures, systems, and 

components.

III.4 Measures shall be established for the identification and 1.28 N45.2

control of design interfaces and for coordination among N45.2.10

participating design organizations. N45.2.11

III.5 These measures shall include the establishment of 1.28 N45.2

procedures among participating design organizations for N45.2.10

the review, approval, release, distribution, and revision of N45.2.11

documents involving design interfaces.

III.6 The design control measures shall provide for verifying 1.28 N45.2

or checking the adequacy of design, such as by the N45.2.10

performance of design reviews, by the use of alternate or N45.2.11

simplified calculational methods, or by the performance of 1.89 IEEE 323

a suitable testing program.
20

Criterion III of Appendix B to 10 CFR Part 50 Cross-Refence Table

III.7 The verifying or checking process shall be performed by 1.28 N45.2

individuals or groups other than those who performed the N45.2.10

original design, but who may be from the same organization. N45.2.11

III.8 Where a test program is used to verify the adequacy of a 1.28 N45.2

specific design feature in lieu of other verifying or checking N45.2.10

processes, it shall include suitable qualification testing of a N45.2.11

prototype unit under the most adverse design conditions. 1.40 IEEE 334

1.73 IEEE 382

1.89

1.100

III.9 Design control measures shall be applied to items such 1.28 N45.2

as the following: reactor physics, stress, thermal hydraulic, N45.2.10

and accident analysis; compatibility of materials; N45.2.11

accessibility for inservice inspection, maintenance and 

repair; the delineation of acceptance criteria for inspections 

and tests. 

III.10 Design changes, including field changes, shall be subject 1.28 N45.2

to design control measures commensurate with those N45.2.10

applied to the original design and be approved by the N45.2.11

organization that performed the original design unless 

the applicant designates another responsible organization.
21

Plant Engineering Products

PLANT ENGINEERING PRODUCTS

A. Overview 

Engineers play a vital role in the design, maintenance, and operation of commercial nuclear 

power plants. Engineering work products include, but are not limited to, the following:

• calculations

• modifications

• temporary modifications

• operability determinations

• procedures

• corrective actions

• dedication of commercial-grade parts for safety-related applications

B. Calculations 

A facility’s operating license requires licensees to implement their operational QA program. As 

discussed earlier, ANSI N45.2.11 parallels the design control requirements and commitments 

in most QA program manuals. In general, the ANSI standard specifies that licensees take the 

following seven actions:

(1) Activities affecting design control shall be specified and accomplished in accordance with 

approved procedures.

(2) Responsibilities shall be specified in procedures.

(3) Applicable design inputs, such as design bases, regulatory requirements, and 

codes and standards, shall be identified. 

(4) Design inputs shall include, where applicable, the following:

• the functions of each SSC 

• performance requirements, such as capacity, rating, and system output

• industry codes, industry standards, and regulatory requirements

• design conditions, such as pressure, temperature, fluid chemistry, 

and voltage

• loads, such as seismic, wind, thermal, and dynamic

• environmental conditions anticipated during storage, construction, and

operation and during a design-basis accident (e.g., pressure, temperature, 

humidity, corrosiveness, site elevation, wind direct, nuclear radiation, 

electromagnetic radiation, and duration of exposure)

• interface requirements

• material requirements

• mechanical requirements, such as vibration, stress, shock, and 

reaction forces

• structural requirements covering such items as equipment foundations 

and pipe supports
22

Plant Engineering Products

• hydraulic requirements, such as pump net positive suction head (NPSH), 

allowable pressure drop, and allowable fluid velocity

• chemistry requirements

• electrical requirements, such as source of power, voltage, raceway 

requirements, electrical insulation, and motor requirements

• layout and arrangement requirements

• operational requirements under various conditions, such as plant startup, 

normal plant operation, plant shutdown, plant emergency operations, 

special or infrequent operations, and system abnormal or emergency operations

• instrumentation and control requirements, including controls and alarms

• access and administrative control requirements for security 

• redundancy, diversity, and separation requirements (single-failure proof)

• identification of those events and accidents against which the SSCs must be 

designed to withstand 

(5) Design analysis shall be performed in a planned and controlled manner. Analyses

shall be sufficiently detailed as to purpose, method, assumptions, design input, references, 

and units such that a person technically qualified in the subject can review and understand 

the design analysis and verify its adequacy.

(6) Assumptions and references shall be documented.

(7) Design verification may include design reviews, alternate calculations, or qualification 

testing, as follows:

• Design Reviews: Design reviews are critical reviews to ensure that design documents, 

such as drawings, calculations, analyses, or specifications, are correct and satisfactory. 

Design reviews can range from multiple-organization reviews to single-person reviews. 

The results of the reviews shall be documented.

• Alternate Calculations: The use of alternate methods of calculations or analyses can 

help verify, through comparison, some types of calculations or analyses. A person or 

persons other than those who performed the original calculations shall perform these 

alternate methods. The alternate method used may be a more simplified approach 

(e.g., using a hand calculation to check a computer code output). A simplified method 

must provide results that are consistent with the original calculations or analyses.

• Qualification Testing: Design verification for some design or specific design features 

can be achieved by suitable qualification testing of a prototype or initial production unit. 

The testing shall demonstrate the adequacy of the performance of a design under the 

most adverse design conditions.

The review and inspection of engineering calculations are important inspection activities. 

Calculations will exist in various forms and for different purposes. In general, most calculations 

are documented in a manner to ensure that applicable design control requirements are met. 

Specifically, 10 CFR 50, Appendix B, Criterion III requires, in part, measures for verifying or 

checking the adequacy of a design. A licensee may accomplish this through design reviews, 

alternate calculations, and/or qualification testing. If qualification testing is used to demonstrate 

the adequacy of a design, the testing of a prototype unit must be performed under the most 

adverse anticipated conditions.
23

Plant Engineering Products

In addition, calculations will often validate the quantitative values contained in the TS, USAR, 

and other commitments to the NRC. Of critical importance is the consistent relationship between 

design calculations, operational surveillances, procedures, and the licensing basis (i.e., NRC 

commitments, TS, and FSARs). Once minimum thresholds of performance are identified, 

procedures may use acceptance criteria that are more conservative, but not less conservative, 

than those specified in calculations. 

A typical calculation may have the following sections:

• Purpose: This section discusses the reason for the calculation.

• Assumptions: This section discusses the starting points for the calculation. Inspectors 

should verify that assumptions are consistent with the licensing-basis and engineeringreference documents. Inspectors should discuss questionable assumptions with the licensee 

staff to fully evaluate their appropriateness, and, where possible, the assumptions should 

be independently verified. Fundamentally, the stated assumptions define the boundaries for 

the calculation and generally make the calculation “simpler.” The inspector should focus on 

whether the assumption is valid and whether it could have a significant impact on the results 

of the calculation. 

Inspectors should verify that the licensee’s “bounding analysis” approach truly bounds 

the worst-case scenarios and events. Often, licensees evaluate Loss of Coolant Accident 

scenarios as worst-case events, but occasionally “smaller” events provide the most limiting 

system conditions and demands.

• References: Calculation references are an important element of a good calculation. Recently 

published papers are not necessarily the most appropriate references—just as time-proven 

“industry classics” are not necessarily valid either. Inspectors should consider whether the 

NRC has endorsed the stated reference in an RG, NUREG, or other regulatory document. 

Absence of this endorsement does not necessarily make the reference inappropriate for use. 

However, the NRC has found that some standards endorsed by bona fide organizations (e.g., 

ASME) do not provide acceptable guidance. Agency endorsement of a code or standard 

provides a high level of confidence in the applicable guidance. Conversely, licensees are 

not expected to follow all standards. Inspectors have little regulatory leverage to require 

licensees to follow a code or standard where there is no corresponding licensee commitment 

or specific NRC regulation. In fact, imposing such a requirement upon licensees could 

constitute a regulatory backfit (see 10 CFR 50.109, “Backfitting,” and associated agency 

guidance documents regarding the backfit process). 

• Design Inputs: An important consideration in a calculation is the inputs. Inputs can include 

licensing-basis information (e.g., minimum flow rates, voltages, operating, and accident 

conditions) and system drawings. Inspectors should verify the accuracy of the calculation 

inputs, as practical as possible, and should develop a good understanding of the pertinent 

system drawings before and during the review of the calculation.

• Analysis: Engineering theory and the mathematics of the methodology are two distinct 

elements that inspectors should consider when reviewing the technical adequacy of an 

engineering calculation. 
24

Plant Engineering Products

Inspectors should verify the validity of the engineering theory. For instance, many calculational 

formulas are valid only under certain conditions. The inspectors should verify that the 

applicable calculational limitations are properly considered. Additionally, inspectors should 

often perform independent, simplified calculations to verify that the licensee’s calculational 

result is correct. 

Criterion III, “Design Control,” requires that “design control measures shall provide for 

verifying or checking the adequacy of design, such as by the performance of design 

reviews, by the use of alternate or simplified calculational methods, or by the performance 

of a suitable testing program....” Inspectors should be able to identify and verify the method 

by which the calculation was independently verified.

The mathematics of the methodology should also be independently checked, where possible. 

Plant engineers are not immune to making errors in their computations. 

Additionally, the licensee’s calculation should also have considered defense-in-depth 

elements. These include single failure, common-mode failure, separation (physical and 

electrical), and redundancy, where applicable. The facility’s licensing-basis documents will 

discuss the applicability of these elements to particular systems. 

• Conclusions: At this stage in the inspection process, the inspector should review the 

calculation conclusions and consider whether the calculation results seem reasonable or 

whether they conflict with other existing approved calculations. The inspector should focus 

on performing critical reviews of the calculations and on verifying their adequacy.

C. Modifications and Temporary Modifications 

Plant modifications are one of the major 

engineering work products throughout the life 

of the plant. Modifications are implemented 

for various reasons, including to satisfy new 

regulatory requirements (e.g., station blackout 

and Appendix R), to improve the reliability of 

SSCs, to resolve degraded or nonconforming 

SSCs, and to improve plant efficiency. 

Plant modifications are important because these 

engineering activities can affect the facility’s 

conformance to the design and licensing basis. 

Control of design basis and plant configuration 

is important to ensure that the plant’s design, 

operation, maintenance, and modifications 

remain consistent with the facility’s FSAR and 

other licensing-basis documents. 

The difference between a modification and a 

temporary modification is the duration of the 

change. Modifications are permanent changes, 

whereas temporary modifications are invoked 

Observation of in-progress activities provides 

opportunities to objectively assess licensee 

performance. 
25

Plant Engineering Products

for a limited duration. Design control requirements apply to all forms of modifications, including 

temporary modifications. The requirements in 10 CFR 50.59 also apply if an SSC is described in 

the FSAR. 

Examples of modifications include, but are not limited to, the following:

• increasing the size of a motor operator

• changing the gear ratio on a motor operator

• installing a different pump into an existing system

• installing a new type of breaker into an existing breaker cubicle

Temporary modifications tend to be less invasive than permanent modifications, but they still can 

be significant. Typical temporary modifications include the following:

• installing temporary power sources to non-safety-related buses during 

an outage

• installing a temporary pressure gauge

• installing temporary temperature instruments

• lifting leads on nuisance alarms in the control room

For changes to the facility that do not last longer than the next outage, refer to Regulatory 

Issue Summary (RIS) 2005-20, “Revision to NRC Inspection Manual, Part 9900: Technical 

Guidance, ‘Operability Determinations & Functionality Assessments for Resolution of Degraded 

or Nonconforming Conditions Adverse to Quality or Safety’,” for important guidance with respect 

to operability and 10 CFR 50.59.

D. Operability Determinations

In many cases, a licensee will identify that an SSC is in a 

degraded condition or a nonconforming condition (i.e., not 

in compliance with an accepted code or standard). The 

inspectors should verify that the licensee has entered the 

concern into their corrective action program. In some cases, 

an engineering evaluation will be performed to support 

an operability determination. This assessment may be 

qualitative or quantitative (i.e., a calculation). The inspectors 

should evaluate the validity of any assumptions and the 

accuracy of calculations.

Agency expectations with respect to SSC operability are 

contained in RIS 2005-20 and its associated attachment, 

NRC Inspection Manual, Part 9900, “Technical Guidance, 

‘Operability Determinations & Functionality Assessments 

for Resolution of Degraded and Nonconforming Conditions 

Adverse to Quality or Safety.” Inspectors should be familiar 

with this guidance. Ensuring that licensees perform adequate 

operability evaluations is an important responsibility of NRC 

inspectors. The following high-level concepts are provided:

Verification of engineering inputs 

and assumptions is critically 

important. 
26

Plant Engineering Products

• Technical Specification Compliance: If a licensee is not meeting a TS requirement, 

the licensee must implement the applicable TS action requirements. In some cases, the 

licensee may be able to show that the risk to the public is not compromised by failure to 

follow the TS action statement. In those cases, the licensee may request a NOED to allow 

a temporary noncompliance with the TS action statement. These requests require NRC 

approval. Specific guidance regarding NOEDs is located in the NRC Inspection Manual, 

Part 9900, “Technical Guidance, ‘Operations - Notices of Enforcement Discretion.”

• Operable and Operability: A system, subsystem, train, component, or device shall be 

“operable” or have “operability” when it is capable of performing its specified function(s) 

and when all necessary attendant instrumentation, controls, electrical power, cooling or seal 

water, lubrication, and other auxiliary equipment that are required for the system, subsystem, 

train component, or device to perform its function(s) are also capable of performing their 

related support functions.

It is important to stress that the SSC must be capable of performing its safety function for the 

required licensing-basis mission time (which may be different than that of the probabilistic risk 

analysis). 

• Inoperable: Inspectors should recognize that under certain circumstances, an SSC can 

be “inoperable” even if it meets its TS surveillance requirements. For example, a diesel 

generator could pass a TS surveillance load test (i.e., carrying the required kilowatts) while 

operating with high cylinder temperatures. In this case, the TS-required surveillance testing 

does not by itself envelop the degraded or nonconforming condition (i.e., high temperatures). 

Upon discovery of a degraded or nonconforming condition, the licensee must evaluate 

the condition and make an operability determination regarding whether the equipment 

can perform the intended design function under worst-case anticipated conditions for the 

duration of the mission time. 

E. Procedures

Regulations, such as Criterion V, “Instructions, Procedures, and Drawings,” of Appendix B to 10 

CFR Part 50 and the administrative sections of the TS, require that all safety-related activities 

be performed in accordance with approved procedures and work instructions. Engineers write 

surveillance and test procedures to verify that SSCs continue to perform satisfactorily in service. 

Engineers also write post--modification test instructions to ensure that the modification did not 

adversely affect the performance of SSCs. When reviewing surveillance and test procedures, 

inspectors should, as a minimum, perform the following actions:

• Verify that test procedure acceptance criteria are consistent with the licensing basis (e.g., 

FSARs and TS).

• Verify that the tests are performed without undue preconditioning. Review guidance 

contained in NRC Inspection Manual, Part 9900, “Technical Guidance, Maintenance - 

Preconditioning of Structures, Systems, and Components Before Determining Operability.” 

Some preconditioning is allowed in a few limited instances (e.g., emergency diesel generator 

testing).

• To the extent practicable, verify that testing reasonably verifies that SSCs will perform 

satisfactorily under design-basis (worst-case) accident conditions. For example, licensees 

will not likely be able to test heat exchangers under worst-case conditions; therefore, they 

must perform an extrapolation of the test data to verify design-basis capabilities.
27

Plant Engineering Products

• Verify that the instruments used in surveillances are calibrated (see Criterion XII, “Control 

of Measuring and Test Equipment,” of Appendix B to 10 CFR Part 50). Test instrument 

uncertainties must be accounted for either explicitly (i.e., in a calculation) or implicitly (i.e., 

there is obviously sufficient margin so that a rigorous calculation is not needed).

F. Corrective Action Determinations 

Plant engineers may specify corrective measures to correct degraded or nonconforming 

conditions. In accordance with Criterion XVI, “Corrective Action,” of Appendix B to 10 CFR Part 50, 

the corrective measures must be prompt and correct the condition adverse to quality. In addition, 

for significant conditions adverse to quality, the licensee must identify the cause of the condition 

and take sufficient corrective measures to prevent repetition. Equipment failures (e.g., emergency 

diesel generator and ECCS pump failures) are significant conditions adverse to quality.

G. Dedication of Commercial-Grade Parts for Safety-Related Applications 

In some instances, safety-grade replacement components are no longer available for purchase. 

As components wear out and need replacement, licensees will obtain commercially available 

replacement parts and will need to qualify the components for safety-related applications. When 

this is done, licensees must take the following actions:

• Assume responsibility for reporting pursuant to 10 CFR Part 21, “Reporting of Defects and 

Noncompliance.”

• Ensure that QA program, seismic, and environmental qualifications requirements are met.

• Establish a means to verify that the component has the ability to perform design-basis 

conditions (e.g., identify safety functions and perform qualification testing, post-maintenance 

testing, and in-service testing). 

• Establish necessary controls to ensure that the component will remain operable.

• Verify that the vendor has acceptable QA practices.

• Establish storage and preventive maintenance requirements.

• Establish a service life for the components.

• Perform periodic audits of the program.

29

Robust Application of Engineering Principles

ROBUST APPLICATION OF ENGINEERING PRINCIPLES

Inspectors should keep in mind that the overall intent of the design control process is to develop a 

well-supported solution to a defined problem affecting an SSC. The adequacy and completeness 

of the design documents are critically important elements and can be accomplished only through 

a well-defined and -controlled process. Effective implementation of such a process will provide 

greater assurance that safety- and risk-significant SSCs will be able to perform when required to 

ensure plant safety. 

Often inspectors will review licensee conclusions 

and analyses and, after discussions with 

licensee engineers, will find that the engineering 

work products rely on “engineering judgment.” 

Engineering judgment refers to technical 

judgments made by knowledgeable engineers 

experienced in the particular subject matter. 

The inspector should consider whether objective 

data exist to support the engineering judgment. 

Reliance on engineering judgment is predicated 

on the consistent and appropriate use of 

engineering “rules of thumb” and reliance on welldeveloped, understood, and widely accepted 

industry practices and standards. Engineering 

judgment does not mean that such conclusions 

cannot be questioned or are indisputable. When 

considering the appropriateness of engineering 

judgment, inspectors should consider the 

following questions:

• Does the engineering judgment rely on assumptions or data that are not relevant to this 

issue (i.e., reliance on past successes to justify current assumptions)?

• Did the licensee consider other data that contradict the engineering judgment?

• Does the engineering judgment apply to “rules of thumb” alone?

• Does the conclusion seem reasonable?

• Does the conclusion account for industry operating experience? 

• Is there conservatism incorporated in the engineer’s conclusion?

• Is unsupported engineering judgment being relied on in lieu of testing to demonstrate design 

adequacy or system operability?

The licensee’s engineering organization should recognize that engineers are guardians of design 

margin and represent the safety conscience of the plant. All engineering actions should reflect a 

thorough knowledge of systems and components, including applicable design and licensing basis. 

Engineering evaluations should be thorough and timely, and the conclusions reached should be 

based on sound engineering practices and reflect a robust application of engineering principles. 

Design Margin

How do you tell the difference between 

a risk analyst, a mathematician, a 

scientist, and an engineer?

First, you ask them, “What does 2 plus 

2 equal?”

The risk analyst will respond, “Whatever 

number you would like it to be, but ithas 

to be between 0 and 1.”

The mathematician will respond, “4.”

The scientist will respond, “The 

evidence points to 4.”

The engineer will respond, “Let’s make 

it 5 to be on the safe side.”
30

Robust Application of Engineering Principles

Engineering activities should reflect meticulous attention to detail to ensure that the plant’s design 

basis and licensing commitments are maintained. Control of design basis and plant configuration 

is critically important to ensure that the plant’s design, operation, maintenance, and modifications 

are consistent with the facility’s FSAR and supporting documents. 

As discussed in previous sections, rigorous design verifications and checking activities are a 

normal part of the design process. Design-basis documents should be continuously used to 

support engineering activities. The licensee’s engineering organization should constantly monitor 

the plant’s operation within the design bases. 

Rigorous engineering activities should be characterized by an effort to ensure safe and correct 

actions rather than the removal of safety margins in an effort to justify the acceptability of existing 

degraded conditions. 
31

Value-Added Findings and STARS Documents

VALUE-ADDED FINDINGS AND STARS DOCUMENTS

Regional staff often issue value-added findings (VAFs) and STARS documents as a method to 

communicate and share inspection findings and work techniques with other inspectors. These 

documents generally describe topics that provide additional benefits to other inspectors by 

describing a generic issue that may be applicable at other facilities or a unique inspection method 

or technique. Table 4 lists VAFs and STARS documents related to design control. 

Table 4. VAFs and STARS Documents

Number Title Region/Site

2000-03 Potential Service Water Pump Inoperability Due to Inadequate 

Forebay Level

2001-08 Modifications to Nonsafety-Related Systems Cause RIII-Prairie Island

Safety-Related Inoperability

2003-01 Inadequate Temporary Cooling for 480Vac Vital Switchgear RI-Millstone 

2004-17 Shutdown Service Water System Pipe Support Deficiencies RIII-Clinton

2004-18 Inappropriate Use of a Common Nonsafety-Related Power 

Source To Feed Two Redundant Safeguard Electrical RIII-Kewaunee 

Control Circuits

2005-14 Incorrect Sprinkler System for the Emergency Feedwater RIV-ANO

Pump Room

2005-15 Higher Room Temperatures Adversely Impacting Cable RIII-Duane Arnold

Ampacity Margins

2005-25 Failure To Perform Adequate Oversight of a Modification RIII-Dresden

Performed by a Contractor

2005-38 Strong Ownership of Issues Successfully Brings Closure RIII-Duane Arnold 

to Old Issues

2005-41 Reactor Core Isolation Cooling Governor Not Qualified for 

Station Black Out Temperatures and Questionable Breaker RIII-LaSalle

Short Circuit Interrupting Capability

2005-42 Failure To Provide Electrical Coordination To Ensure That RIII-Fermi

Fire-Induced Electrical Faults Would Not Result in the Loss 

of Post-Fire Alternative Safe Shutdown Equipment

2006-03 Single Cell Charging Design Control RIV-Columbia

2006-09 Containment Spray Suction Pipe Voiding RIV-Wolf Creek

2006-10 Net Positive Suction Head of the Reactor Core RIII-LaSalle 

Isolation Cooling Pump

2006-12 Inadequate Freeze Protection RIV-Grand Gulf

2006-14 Inadequate Technical Review of Permanent Modification RIII-Clinton

2006-18 Temporary Alteration Fails To Meet Design Analysis RI-Indian Point 2

Assumptions

33

References

REFERENCES

References discussed within this booklet are listed below. Users should thoroughly understand 

the licensee’s licensing basis and commitments to understand the applicability of the reference 

documents. The industry standards discussed within the booklet are those which were generally 

in-use during the licensing phase of a significant number of nuclear power plants. Additionally, 

industry standards often are revised and the specific content may change; hence, specific revisions 

and dates of such standards, and other references, are omitted. 

Code of Federal Regulations

10 CFR 21 “Reporting of Defects and Noncompliance”

10 CFR 50 “Domestic Licensing of Production and Utilization Facilities”

10 CFR 50.2 “Definitions” 

10 CFR 50.34 “Contents of Applications; Technical Information”

10 CFR 50.54 “Conditions of Licenses”

10 CFR 50.55a “Codes and Standards”

10 CFR 50.59 “Changes, Tests, and Experiments” 

10 CFR 50.67 “Accident Source Term”

10 CFR 50.71 “Maintenance of Records, Making of Reports”

10 CFR 50.90 “Application for Amendment of License, Construction Permit, or Early Site 

Permit” 

10 CFR 50.109 “Backfitting”

10 CFR 50 “General Design Criteria for Nuclear Power Plants”

Appendix A

10 CFR 50 “Quality Assurance Criteria for Nuclear Power Plants and Fuel

Appendix B Reprocessing Plants”

10 CFR 54 “Requirements for Renewal of Operating Licenses for Nuclear 

 Power Plants”

10 CFR 54.3 “Definitions” 

10 CFR 100.11 “Determination of Exclusion Area, Low Population Zone, and

 Population Center Distance”

NUREGs

NUREG-0800 “Standard Review Plan for the Review of Safety Analysis Reports for 

 Nuclear Power Plants” 

NUREG-1055 “Improving Quality and the Assurance of Quality in the Design and 

 Construction of Nuclear Power Plants” 

NUREG-1397 “An Assessment of Design Control Practices and Design 

Reconstitution Programs in the Nuclear Industry,” February 1991
34

Regulatory Guides

RG 1.1 “Net Positive Suction Head for Emergency Core Cooling and 

Containment Heat Removal System Pumps” (Safety Guide 1)

RG 1.6 “Independence Between Redundant Standby (Onsite) Power 

Sources and Between Their Distribution Systems” (Safety Guide 6)

RG 1.9 “Application and Testing of Safety-Related Diesel Generators in 

 Nuclear Power Plants”

RG 1.26 “Quality Group Classifications and Standards for Water-, Steam-, 

 and Radioactive-Waste-Containing Components of Nuclear 

 Power Plants”

RG 1.28 “Quality Assurance Program Requirements (Design and 

 Construction)”

RG 1.29 “Seismic Design Classification”

RG 1.40 “Qualification Tests of Continuous-Duty Motors Installed Inside the 

 Containment of Water-Cooled Nuclear Power Plants” 

RG 1.47 “Bypassed and Inoperable Status Indication for Nuclear Power 

 Plant Safety Systems”

RG 1.53 “Application of the Single-Failure Criterion to Safety Systems” 

RG 1.63 “Electric Penetration Assemblies in Containment Structures for 

 Nuclear Power Plants”

RG 1.73 “Qualification Tests of Electric Valve Operators Installed Inside the 

 Containment of Nuclear Power Plants” 

RG 1.75 “Criteria for Independence of Electrical Safety Systems”

RG 1.89 “Environmental Qualification of Certain Electric Equipment 

 Important to Safety for Nuclear Power Plants”

RG 1.97 “Criteria for Accident Monitoring Instrumentation for Nuclear 

Power Plants”

RG 1.100 “Seismic Qualification of Electric and Mechanical Equipment for 

Nuclear Power Plants”

RG 1.106 “Thermal Overload Protection for Electric Motors on MotorOperated Valves”

RG 1.147 “Inservice Inspection Code Case Acceptability, ASME Section XI, 

Division 1”

RG 1.186 “Guidance and Examples for Identifying 10 CFR 50.2 

Design Bases”

References
35

References

RG 1.187 “Guidance for Implementation of 10 CFR 50.59, Changes, Tests, 

and Experiments” 

RG 1.193 “ASME Code Cases Not Approved for Use”

Regulatory Issue Summary

RIS 2005-20 “Revision to NRC Inspection Manual, Part 9900 Technical

Guidance, ‘Operability Determinations & Functionality Assessments 

for Resolution of Degraded or Nonconforming Conditions Adverse 

to Quality or Safety’”

Information Notices

N 84-54 “Deficiencies in Design Base Documentation and Calculations Supporting 

Nuclear Power Plant Design” 

IN 91-29 “Deficiencies Identified During Electrical Distribution System 

Functional Inspections”

IN 97-81 “Deficiencies in Failure Modes and Effects Analyses for 

Instrumentation and Control Systems”

IN 98-22 “Deficiencies Identified During NRC Design Inspections” 

Generic Letters

GL 88-15 “Electric Power Systems—Inadequate Control Over Design

Processes” 

GL 91-05 “Licensee Commercial-Grade Procurement and Dedication

Programs”

GL 92-03 “Compilation of the Current Licensing Basis”

GL 95-04 “Final Disposition of the Systematic Evaluation Program 

Lessons-Learned Issues”

Industry Standards and Guidance Documents

ANSI/ISA-67.04.01 “Setpoints for Nuclear Safety-Related Instrumentation”

ANSI/ASME N45.2 “Quality Assurance Program Requirements for Nuclear 

 Facilities”

ANSI/ASME N45.2.10 “Quality Assurance Terms and Definitions”

ANSI/ASME N45.2.11“Quality Assurance Requirements for the Design of Nuclear 

Power Plants”

ANSI/ASME NQA-1 “Quality Assurance Program Requirements for Nuclear 

Facilities”
36

References

NEI 97-04 “Design Bases Program Guidelines”

NEI 96-07 “Guidelines for 50.59 Evaluations”

IEEE 279 “Criteria for Protection Systems for Nuclear Power Generating 

Stations”

IEEE 308 “Criteria for Class 1E Power Systems for Nuclear Power Generating 

Stations” 

IEEE 323 Qualifying Class 1E Equipment for Nuclear Power Generating 

Stations”

IEEE 334 “Standard for Type Tests of Continuous-Duty Class 1E Motors for 

Nuclear Power Generating Stations”

IEEE 379 “Standard Application of the Single-Failure Criterion to Nuclear 

Power Generating Station Safety Systems”

IEEE 382 “Trial Use Guide for Type Test of Class 1E Electric Valve Operators 

for Nuclear Power Generating Stations”

IEEE 450 “Recommended Practice for Large Lead Storage Batteries for 

Generating Stations and Substations”

IEEE 484 “Recommended Practice for Installation Design and Installation of 

Large Lead Storage Batteries for Generating Stations and 

Substations”

IEEE 603 “IEEE Standard Criteria for Safety Systems for Nuclear Power 

Generating Stations” 
37

Acronyms

ACRONYMS

ANSI American National Standards Institute

ASME American Society of Mechanical Engineers

BPV Boiler and Pressure Vessel

CFR Code of Federal Regulations

CLB Current Licensing Basis 

DBD Design Basis Documentation 

ECCS Emergency Core Cooling Systems 

FSAR Final Safety Analysis Report

GL Generic Letter

IEEE Institute of Electrical and Electronics Engineers

NOED Notice of Enforcement Discretion

NRC U.S. Nuclear Regulatory Commission

NRR Office of Nuclear Reactor Regulation

NRO Office of New Reactors 

NEI Nuclear Energy Institute

NPSH Net Positive Suction Head 

NUREG Nuclear Regulation

PSAR Preliminary Safety Analysis Report

QA Quality Assurance

RG Regulatory Guide

RIS Regulatory Issue Summary

ROP Reactor Oversight Process

SEP Systematic Evaluation Program

SER Safety Evaluation Reports

SRP Standard Review Plans

SSC Structures, Systems, and Components

TS Technical Specifications

UFSAR Updated Final Safety Analysis Report 

USAR Updated Safety Analysis Report

VAF Value Added Findings

NOTES

NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER 

(.·200<11 (Assigned by NRC, Add Vol., Supp., R,y .. 

NRCM03.7 _nd Add ....... '" Nu_, I' any.) 

BIBLIOGRAPHIC DATA SHEET 

{$eft otI5tnICtIOtIS on rI>e feW1fStI} 

NUREG-1913 

2 TITLE AND SUBTITLE J. DATE REPORT PUBLISHED 

DESIGN CONTROL """'" """ 

A Quick Reference Guide For NRC Inspectors August I 2009 

4. FIN OR GRANT NUMBER 

5. AUTHQR(S) 6. TYPE OF REPORT 

Julio F. Lara, P.E. NUREG 

Jennifer Tifft 

Frank Arner 7. PERIOD COVERED (inciusr-e ~te51 

Randy Moore 

George Replogle 

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (I/NRC. pro_ OM..",. Ol'/iat or Reg;on. IJ S. NIIt:IHr R'fgul.IOf)I CommisSlOll •• nd m&ing address. if oon/,KfO(. 

pro_ ".me .nd ~ .oar.n I 

Region 111 

U. S. Nuclear Regulatory Commission 

2443 Warrenvile Rd. 

Lisle, IL 60532 

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (If NRC. !)ope -S-me.s ebQ>e-. i1«J1!ltectot. ~ NRC~. cw.c. or~. IJ S NtJdeet Regu/Mory~, 

..,.,....-.g I<bItJl) 

Same as above 

10 SUPPLEMENTARY NOTES 

11 . ABSTRACT (200 words orleu) 

Knowledge management and transfer have become increasingly important as the U.S. Nuclear Regulatory Commission seeks 

to share the vast inspection knowledge of its experienced inspectors with those who have recently joined the agency. This 

booklet should serve as an inspection reference to further increase the understanding by, and development of, NRC inspectors 

who periorm inspections in the areas covered by this booklet. 

12. KEY WORDSIOESCRIPTORS (lAr 'oOCtOs or phr'S#J$ IhIl .... $$<$1 <eS1NlO'lerS ".b::ellnsl1IHI repOrt.) 13 AVAlLABlJTY$TATEMfNT 

Knowledge Management unlimited 

Design Control 1~ SECURITY CLASSIFICATION 

Quick Reference Guide (ThIs Pege) 

unclassified 

(ThIs RePOt!I 

unclassified 

15, NUMBER OF PAGES 

15. PRICE 

NRC FORM 335 (9-20()01) PRINTED ON RECYCLED PAPER 
NOTES
NOTES
NOTES
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IN NRC PUBLICATIONS 

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As of November 1999, you may electronically access 

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NRC publications in the NUREG series, NRC 

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These standards are available in the library for 

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NUREG-0750), 
NUREG-1913

August 2009