## **Continuing Evolution of U.S. Nuclear Quality Assurance Principles, Practices and Requirements -- PART I** 

August 2005

###  *This document provides in two parts an accounting of the continuing evolution of quality assurance principles, practices and requirements for nuclear facility applications in the United States from 1954 to 2005.* 

**EXECUTIVE SUMMARY**[^1]***.***

### 

### PART I (this part) describes in SECTIONS 1 through V how nuclear quality assurance (NQA) and its documentation has evolved along four separate yet interrelated paths: (1) AEC\>ERDA\>DOE standards and directives; (2) AEC\> NRC regulations and regulatory guides; (3) ANSI N45.2\>ASME NQA-1 and related standards; and (4) ASME Section III Boiler and Pressure Vessel Code. It tracks the evolution of early AEC quality control and acceptance inspection requirements and practices for nuclear weapons production from the 1950s; the AEC quality assurance requirements for government-owned reactor and technology development programs from the 1960s; the AEC licensing regulations for designing, constructing and operating commercial nuclear power plants and fuel reprocessing plants from the 1970s; the development of ASME national consensus standards for nuclear facilities from the 1970s; the ASME code quality assurance criteria from the 1960s; and the management system and performance-based approach, also from the 1970s to date. SECTION VI outlines ASME NQA management issues. Section VII presents the ASME NQA Committee vision. PART I concludes with an attachment containing the basic NQA requirements of the ASME NQA-1-2004 Standard and an explanation of their implementation practices and benefits to be derived for the continuing assurance of quality in nuclear facilities. 

##### PART II is a tutorial that discusses NQA terms and definitions; paradigm shifts from ineffectual NQA concepts; underlying principles of NQA-1 requirements; lessons learned for effective NQA programs; and a condensed chronology of events.

**At the onset** of nuclear power generation in the U.S., AEC regulators
and nuclear utility owners were primarily concerned with assuring safe
plant operation without due regard to formal management controls that
were essential to ensure quality in achieving both safe and reliable
operation of these complex facilities. Untoward problems during
commercial nuclear facility design and construction phases eventually
caused AEC regulators and nuclear utility plant owners to realize the
importance and interrelationship of NQA to nuclear safety. Similarly,
problems in AEC-owned reactor and test facilities led to the development
of quality assurance standards and practices.

**Early NQA activities** were focused on the design of commercial
nuclear power facilities. This focus then shifted to construction
activities. Current emphasis is on the operations and maintenance of
existing facilities.

Anticipating a revitalized U.S nuclear power plant industry and
associated fuel reprocessing, design, construction and operation, the
American Society of Mechanical Engineers (ASME) NQA Committee

envisions in Section VII: (1) a broader adoption by the nuclear industry
of ASME NQA-1 Standards: (2) a growing application of the Standards to
DOE reactor and non-reactor nuclear facilities, and (3) a more timely
endorsement by the NRC of successive versions of the Standards beyond
1994.

At the April 2004 meeting of the ASME NQA Main Committee approved Task
Proposal Notice (TPN) enabling development of an historical and tutorial
document on the origins, purpose and benefits to be derived from the
principles, practices and requirements of quality assurance standards
for nuclear facilities from 1954 to 2004. This document (PARTS I and II)
is for information only and is not a part of the ASME NQA-1-2004
Standard.

######### TABLE OF CONTENTS 

**EXECUTIVE SUMMARY** 1

**TABLE OF CONTENT** 2

**ACRONYMS** 3

**PREFACE** 4

**TIMELINE** 6

**SECTION I.**

**AEC/ DOE QUALITY ASSURANCE REQUIREMENTS AND RULES** 9

-   **AEC Weapons Quality Policy (QC-1)** 9

-   **AEC Naval Reactor QRC-82C** 11

-   **AEC/DOE RDT F2-2T Standard** 11

-   **DOE Order 5700.6** 14

-   **DOE Order 5700.6C** 14

-   **DOE Order 414.1A** 16

-   **DOE Order 414.1B** 16

-   **DOE Rule 10 CFR 830** 17

-   **DOE Order 414.1C** 17

-   **Implementation Guides** 18

-   **DOE Action Plan** 18

    **SECTION II.**

    **AEC/NRC REGULATIONS** 19

-   **Early Nuclear Power Plants** 19

-   **AEC 10 CFR 50 Appendix A** 20

-   **Zion Station Hearings Impact** 21

-   **Quality Assurance Redefined** 22

-   **10 CFR 50 Appendix B** 22

-   **Fuel Reprocessing Plants** 25

-   **Appendix B Criterion I Amendment** 26

-   **Ford Amendment Study** 26

-   **Three Mile Island** 27

    **SECTION III.**

    **ANSI\>ASME N45.2 STANDARDS** 28

-   Recognition of Need 28

-   ANSI N45.2-1971 Standard 28

-   **ASQC Matrix** 30

-   **AEC Safety Guide 28** 31

-   **N45-2 Daughter Standards** 31

-   **AEC Rainbow Series** 33

-   **ASME N45.2-1977** 33

    **SECTION IV.**

    **ASME AND RELATED NQA STANDARDS** 35

-   **"N" Committees Standards** 35

-   **ANS N18.7 (3.2) Standards** 35

-   **ANSI N46.2 Standards** 35

-   **ASME NQA Committee** 35

-   **ASME NQA-1-1979** 36

-   **ASME NQA-1-1983** 37

-   **ASME NQA-1-1986** 37

-   **ASME NQA-2 Standards** 37

-   **ASME NQA-1-1989** 38

-   **ASME NQA-3-1989** 38

-   **ASME NQA-1-1994** 39

-   **ASME NQA-1-1997** 40

-   **ASME NQA-1-2000** 41

-   **ASME NQA-1-2004** 41

-   **ASME NQA Survey** 41

-   **ASME NQA Standards Benefits** 41

**SECTION V.**

**ASME B&PV CODE SECION III**

**QUALITY ASSURANCE CRITERIA** 42

**SECTION VI.**

**ASME NQA MANAGEMENT ISSUES** 42

**SECTION VII.**

**ASME NQA VISION** 43

**TABLE I.**

**DOD MIL-Q-9858A, NASA NHB 5300.4(1B)**

**AND NRC 10 CFR 50 APPENDIX B COMPARISON** 44

**TABLE II.**

**ASME NQA-1-1979 STRUCTURE** 45

ATTACHMENT.

ASME NQA-1-2004 BASIC REQUIREMENTS

AND PRACTICES 46

ACRONYMS

ACRS Advisory Committee for Reactor Safety

AEC Atomic Energy Commission

AL Albuquerque Operations Office

ANS American Nuclear Society

ANSI American National Standards Institute

ASLB Atomic Safety and Licensing Board

ASME American Society of Mechanical Engineers

B&PV Boiler & Pressure Vessel

CFR Code of Federal Regulation

CO Compliance

CP Construction Permit

CRD Contractor Requirements Document

DOD Department of Defense

DOE Department of Energy

DNFSB Defense Nuclear Facility Safety Board

DP Defense Programs

DRL Director of Reactor Licensing

DRS Director of Reactor Standards

ERDA Energy Research and Development Administration

FFTF Fast Flux Test Facility

GOCO Government-owned, contractor- operated

MWE Megawatts electrical

NASA National Aeronautics and Space Administration

NFPQT Nuclear Facility Personnel Qualification and Training

NNSA National Nuclear Security Administration

NQA Nuclear Quality Assurance

NR Naval Reactors

NRC Nuclear Regulatory Commission

NSMB Nuclear Standards Management Board

OGC Office of General Counsel

OMB Office of Management and Budget

PAAA Price Anderson Amendments Act

QC Quality control

RDT Reactor Development and Technology

SNAP Space Nuclear Auxiliary Power

SFO Santa Fe Operations Office

**PREFACE**[^2]

**Nuclear quality assurance** had its origin in the quality control and
inspection practices of World War II. Quality control and inspection
requirements were exercised through statistical process control
techniques. These techniques were embedded in the early military and
industrial products and nuclear weapons production standards. Quality
assurance emerged as an adjunct engineering practice.

**Early engineering efforts** to design and construct components for
nuclear power plants evolved through nuclear code cases arising from
non-nuclear boiler and pressure vessel codes and standards. In 1963 and
1974, the first nuclear vessel codes emerged and became the foundation
standard for extension to other pressure-retaining nuclear components
rules. With the initial edition of the nuclear vessel code, Section III,
the ASME provided a tutorial guide that was extremely helpful when
orienting new people to the principles and practices underlying the
rules and procedures that governed nuclear component design and
fabrication. Later, as the body of experience and understanding grew and
the nuclear component codes matured, this guidance was no longer
essential and thus no longer promulgated.

**In 1962, Vice Admiral Hyman G. Rickover** [^3], the recognized father
of naval nuclear propulsion, spoke of a "cultural lag" in nuclear power
plant management and manufacturing. He said industry practices were not
geared to the higher standards imposed by the new power reactor
technology. He laid out three principles for improving quality
management:

1.  More effective management and engineering attention must be given to
    routine and conventional aspects of nuclear power.

2.  Specifications must be clearly understood, respected and enforced by
    manufacturers as well as customers.

3.  More effective use must be made of quality assurance program
    requirements.

**An exponential growth** in the nuclear power plant market began in
1965. This growth followed the successful demonstration of commercial
nuclear power at the Shippingport, Pennsylvania, nuclear plant. At that
time, eight reactors with a combined capacity of 4870 megawatts
electrical (Mwe) were on order. In the first eight months of 1966, 15
more reactors with a total capacity of 11,800 Mwe were ordered. By
November 1966, there were 52 civilian power reactors with a total
capacity of 26,890 Mwe on order. The AEC predicted an increase in
capacity of from 80,000 to 110,000 total Mwe by 1980. Plant capacity had
increased in size from several hundred to 1100 Mwe, including multiple
units at some sites such as Commonwealth Edison's Dresden station. Also,
plants were being located in the proximity of largely populated
metropolitan area.

**This rapid growth** in nuclear power plant orders and construction in
the 1960's eventually raised considerable concern among the members of
the U.S. Congress, the AEC Commissioners and their inspectors, and
senior utility industry officials. These concerns focused on the
questions of:

> **Did the nuclear industry have sufficient numbers of skilled people
> to staff these very large and technically challenging projects without
> compromising the high quality standards necessary to protect public
> and worker safety? Conversely, did the AEC have sufficient staff to
> inspect, evaluate and oversee licensee applications and construction
> permits for nuclear power plants?**

**AEC Commissioner James Ramey** and RDT Director Milton Shaw spoke on
numerous occasions about the need for quality assurance in nuclear
reactor design and development projects and facility construction.

**At a meeting of the American Nuclear Society** in 1966, Commissioner
Ramey[^4] defined quality assurance as comprising:

> **"*All actions necessary to provide adequate confidence that a
> product or facility will operate satisfactorily in service."***

**This definition** was consistent with the DOD military
specification[^5] that defined quality assurance as[^6]:

> ***"A planned and systematic pattern of all activities necessary to
> provide adequate confidence that the item or product conforms to
> established technical requirements"***.

**In 1968, Commissioner Ramey** addressed the American Power Conference
expressing his concerns about insufficiently experienced organizations
causing errors and omissions resulting in startup problems and delays in
nuclear power plant construction. He emphasized that these problems and
delays demanded management leadership and urgent attention by the
nuclear utilities.[^7] He referred to his 1966 definition of quality
assurance and the practices necessary for an effective quality assurance
program.

**Commissioner Ramey's concerns**, when coupled with other unplanned
events, led eventually to the development of AEC regulation on nuclear
quality assurance, known as Appendix B to 10 CFR Part 50. In response to
Appendix B, the ASME-sponsored American National Standards N45 Committee
formed a quality assurance subcommittee to develop implementing quality
assurance standards. This subcommittee subsequently became the ASME
Nuclear Quality Assurance (NQA) Committee. Since their inception, these
ASME standards have carefully preserved the early definition and the
0enlightened concepts of nuclear quality assurance.

**Throughout the late 1960s and 1970s**, as nuclear power plant
construction projects continued to grow in size and numbers, groups of
people were trained to conform to the ANSI N45.2 standards and later to
the ASME NQA-1 standards. As we know now, the nuclear industry designed,
constructed and successfully operated over 100 nuclear power plants;
however, no new plants have been ordered for over 20 years. During this
time, many of the skilled workers in the US nuclear industry who
managed, designed and constructed these plants have moved to other
careers, retired, or are nearing retirement. The nuclear industry may
again be approaching the same situation it initially faced in the early
1960's: a lack of knowledgeable and skilled management, technical and
quality assurance professionals.. There has been a tremendous amount of
accumulated experience and best practices have been developed,
documented and codified over the past 50 years. This knowledge must not
be lost to the future designers, constructors and operators of nuclear
power generation facilities.

**With the prospect emerging** again for new nuclear power plant orders
and a new cycle of growth in the nuclear power industry, the ASME NQA
Committee believes it is appropriate and timely to prepare an historical
record of events (PART I) and tutorial (PART II )as guidance for the
next generation of managers, technical specialists and nuclear quality
assurance professionals. The ASME NQA Committee intends that this
publication will be publicly available on the ASME website and used to
acquaint newly involved management, technical and quality assurance
professional with the "what", "how" and "why" of the principles,
practices and requirements that have been defined and documented in ASME
NQA-1 and other standards, as well as with some of the key quality
management issues.

#########  {#section-2}

######### TIMELINE

The following chronology traces significant events and reactions in the
evolution of AEC, ERDA, DOE, ASME and other related NQA standards and
directives from 1954 to 2004:

-   **1954:** The Atomic Energy Act of 1954 amended the Atomic Energy
    Act of 1946, defined the AEC function, and established the AEC in
    Germantown, Maryland, and Washington, D.C., in 1957, encompassing
    both regulatory and developmental functions.

-   **1954:** AEC Santa Fe Operation Office issued QC-1, Weapons Quality
    Policy, prescribing nuclear weapons production quality control and
    inspection practices.

-   **1956**: Dresden 1 and Indian Point 1 received the first AEC
    construction permits (CP) under Title 10, Code of Federal
    Regulation, Part 50, with no specified quality assurance program
    criteria or requirements.

-   **1963:** ASME issued Section III of the Boiler and Pressure Vessel
    (B&PV) Code with no specified quality assurance requirements.

-   **1965-1967:** AEC developed a proposed Appendix A to 10 CFR Part 50
    covering nuclear power plant design criteria. Criterion 1, Quality
    Standards and Records, required (1) quality standards, (2) a quality
    assurance program and (3) quality records for structures, systems
    and components important to safety.

-   **1966:** Fermi 1 incident resulted from unauthorized design changes
    causing partial reactor core meltdown. Also, AEC reported ten
    reactors that were in operation for approximately 2 ½ years, were
    then closed.

-   **1967:** During Turkey Points 3 & 4 CP review, the ACRS asked about
    but did not pursue "methods of quality control".

-   **1967:** ASME published Section III, Appendix IX, of the B&PV Code
    containing 15 quality assurance criteria and requiring ASME review
    and approval.

-   **1967:** Brown's Ferry 1 and 2 CP review found a lesser commitment
    to quality assurance. ACRS was concerned because these were the
    first reactors to exceed 1000 Mwe.

-   **1967:** The AEC regulatory function moved to Bethesda, MD, while
    the development function stayed in Germantown, MD.

-   **1968:** The Atomic Safety and Licensing Board (ASLB) suspended
    public hearings on a Commonwealth Edison's application to design and
    construct a nuclear power plant because (1) the license applicant
    did not have a quality assurance program for the plant and (2) the
    AEC did not have criteria for evaluating the adequacy of the
    applicant's quality assurance program

-   **1969:** The AEC Reactor Development and Technology (RDT) Division
    developed and issued a comprehensive quality assurance program
    standard, RDT F2-2T, for its government-owned and
    contractor-operated (GOCO) reactors and test facilities.

-   **1969:** AEC regulatory function proposed for public comment 18
    quality assurance criteria as Appendix B to 10 CFR Part 50 for
    licensing nuclear power plants,

-   **1969:** Representatives of the AEC and the nuclear industry met to
    begin developing N45.2 standards on quality assurance program
    requirements and guidance for nuclear power plants.

-   **1970:** Following an extensive pubic comment period and a trial
    use at Surry, the AEC issued 18 quality assurance criteria for
    nuclear power plants as Appendix B to 10 CFR 50, thereby expanding
    upon Criterion 1 of Appendix A to 10 CFR 50.

-   **1971:** AEC issued Appendix A of 10 CFR 50.

-   **1971:** AEC expanded Appendix B of 10 CFR 50 to apply the 18
    quality assurance criteria to fuel reprocessing plants as well as to
    nuclear power plants.

-   **1971:** The American National Standard (ANSI) N45-2-1971, was
    issued. Supplementary N45.2 (daughter) standards were issued in
    subsequent years.

-   **1972:** American Nuclear Society (ANS) published ANS 3.2-1972
    standard for administrative controls during nuclear power plant
    operation.

-   **1974:** The Energy Reorganization Act of 1974 abolished the AEC
    and established separate agencies in the Nuclear Regulatory
    Commission (NRC) and the Energy Research and Development
    Administration (ERDA).

-   **1974**: Due to some organizational issues at LaSalle and Midland,
    the AEC\>NRC proposed an amendment to Criterion I, Organization, of
    10 CFR Appendix B with regard to permissible organizational
    relationships; Criterion 1 amendment was approved and issued early
    in 1975.

-   **1975:** ASME established the ASME Committee on Nuclear Quality
    Assurance (NQA) to continue developing, coordinating, and
    consolidating and restructuring nuclear quality assurance standards.

-   **1975**: Browns Ferry fire occurred.

-   **1977:** ASME NQA Committee issued ANSI/ASME N45.2-1977 on quality
    assurance program requirements for nuclear facilities.

-   **1977**: ERDA was abolished with the creation of the U.S.
    Department of Energy.

-   **1978:** ANSI N46-2 Committee issued Revision 1 of its ANSI
    N46.2-1978 standard for Post-Reactor Nuclear Fuel Cycle Facilities,
    which was subsequently withdrawn.

-   **1979:** Three Mile Island Unit 2 suffered severe operational
    casualty due to minor maintenance errors and a stuck pressure-relief
    valve leading to a loss of crew operational awareness and resulting
    in major core damage.

-   **1979:** ASME NQA Committee issued ASME NQA-1-1979 on quality
    assurance requirement for nuclear facilities

-   **1981:** DOE issued Order 5700.6, Quality Assurance, in response to
    deficiencies observed by the DOE Inspector General in DOE nuclear
    facilities. This Order was superseded by DOE Orders 5700.6A, in
    1981, 5700.6B in 1986 and 5700.6C in 1991, which was superseded by
    DOE Order 414.1 in 1998'

-   **1983:** ASME NQA Committee issued ASME NQA-1-1983.

-   **1983:** ASME B&PV Code Section III, Appendix IX, adopted ASME
    NQA-1-1983 edition with addenda through 1992.

-   **1983:** ASME NQA Committee incorporated seven NQA-2 quality
    assurance standards as parts of ASME NQA-2-1983.

-   **1985:** NRC endorsed ASME NQA-1-1983 in Revision 3 of Regulatory
    Guide 1.28.

-   **1986:** ASME NQA Committee issued ASME NQA-1-1986 with minor
    editorial changes to the 1983 edition, with several positions.

-   **1989:** ASME NQA committee issued ASME NQA-1-1989 on quality
    assurance requirements for nuclear facilities.

-   **1989:** NRC move to amend 10 CFR50 Appendix B to address
    counterfeit and fraudulent parts was unsuccessful.

-   **1991:** NRC endorsed ASME NQA-1 and NQA-2 in NRC Standard Review
    Plan 17.3.

-   **1991:** DOE published a proposed Nuclear Safety Management Rule
    under 10 CFR Part 830 and subpart 830.120 on quality assurance; DOE
    also issued DOE Order 5700.6C, Quality Assurance, introducing ten
    performance-based quality assurance criteria including the concept
    of quality improvement. These ten criteria were used in the proposed
    Rule.

-   **1994:** DOE published the Nuclear Safety Management Rule, 10 CFR
    Part 830 and subpart 830.120. The Rule provides for civil and
    criminal penalties similar to the NRC rules for commercial nuclear
    facilities.

-   **1994:** ASME NQA Committee issued the ASME NQA-1-1994 edition,
    quality assurance requirements for nuclear facility applications.

-   **1997:** ASME NQA Committee issued the ASME NQA-1-1997 edition with
    continued restructuring and removal of redundant text.

-   **1998:** DOE Order 414.1, issued in November 1998, was superseded
    by DOE Order 414.1A in May 1999, by DOE Order 414.1B in April 2004.

-   **2000**: ASME NQA Committee issued ASME NQA-1-2000 with minor
    revisions to the 1997 edition.

-   **2001:**DOE revised the Nuclear Safety Management Rule to include
    Safety Basis requirements and minor changes to the quality assurance
    Rule, clarifying its applicability to nuclear weapons and
    radiological facilities.

-   **2004:** ASME NQA Committee issued ASME NQA-1-2004 containing
    numerous revisions to the 2000 edition.

-   **2004:** DOE/NNSA issued Revision 10 of QC-1 on DOE weapons quality
    policy.

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-   **2005:** DOE Order 414.1C, Quality Assurance, supersedes DOE Order
    414.1B

-   **2005:** DOE published Action Plan based on lessons learned from
    the Columbia Space Shuttle accident and Davi-Besse reactor pressure
    vessel head corrosion event.

**SECTION I.**

AEC/ DOE QUALITY ASSURANCE REQUIREMENTS AND RULES

*This section describes the evolution of nuclear quality assurance from
early quality control and inspection requirements for AEC nuclear
weapons production and Naval Reactors programs; to the more
comprehensive quality assurance program requirements for nuclear weapons
production and AEC/DOE reactor development and technology activities;
and to the series of DOE quality assurance directives.*

###  {#section-3 .unnumbered}

### [AEC Weapons Quality Policy (QC-1)]{.underline} {#aec-weapons-quality-policy-qc-1 .unnumbered}

###  {#section-4 .unnumbered}

### AEC quality management policy for nuclear weapons complex activities was first documented in the former AEC Santa Fe Operations (SFO) Weapons Quality Policy (QC-1). Issued in April 1954, QC-1 pre-dated MIL-Q-9858A [^8], which was issued in 1959 and widely the used Department of Defense (DOD) specification for military quality assurance programs. {#aec-quality-management-policy-for-nuclear-weapons-complex-activities-was-first-documented-in-the-former-aec-santa-fe-operations-sfo-weapons-quality-policy-qc-1.-issued-in-april-1954-qc-1-pre-dated-mil-q-9858a-which-was-issued-in-1959-and-widely-the-used-department-of-defense-dod-specification-for-military-quality-assurance-programs. .unnumbered}

The initial QC-1 quality control and inspection principals and
requirements included:

+-----------------------------------------------------------------------+
| -   Specification and drawing control                                 |
|                                                                       |
| -   Quality control procedures                                        |
|                                                                       |
| -   Control of inspection gauging and test equipment                  |
|                                                                       |
| -   Production tooling accuracy                                       |
|                                                                       |
| -   In-process inspection and records                                 |
|                                                                       |
| -   Control of special processes                                      |
|                                                                       |
| -   SFO/DOE source inspection                                         |
|                                                                       |
| -   Raw material and deviation control                                |
+-----------------------------------------------------------------------+

QC-1 prescribed general principles and practices for AEC-SFO acceptance
inspection of nuclear weapons systems and auxiliary equipment from prime
contractors. It required weapons program prime contractors to establish
and implement quality control systems to assure, among other things,
that nuclear weapons materials met minimum quality standards. SFO
expected these principles and requirements to be applied also to
ordnance plants operated by the DOD on the behalf of SFO and to arsenals
that performed work for the SFO under agreements with DOD.

In 1982[^9] and again in 1989[^10], the DOE Assistant Secretary for
Military Applications defined and redefined in greater detail the
quality assurance policy for the DOE nuclear weapons complex. This
policy required the execution and maintenance of procedures that:

1)  Provided control, through pans and actions, over activities
    affecting quality to an extent consistent with defined programmatic
    or organizational objectives;

2)  Had objective, measurable means to assure its effectiveness, which
    shall be used by management for regular assessments;

3)  Emphasized continuous improvement in all activities including
    support as well as operational organizations; and

4)  Applied appropriate elements of recognized standards.

    QC-1 has been revised numerous times from its initial issue in 1954
    to its most recent issue in 2004 to address changes and additional
    DOE weapons quality policy and quality assurance requirements. Thus,
    for example, Revision 6, issued in 1992, added quality system
    requirements for training of manufacturing, inspection and test
    personnel, for quality improvement, for error prevention versus
    detection, and for nonconformance costs.

    The highly classified nature of most DOE Defense Program (DP)
    weapons production activities governed by QC-1 led to some external
    criticism that DP lacked a viable quality assurance effort that
    complied with DOE quality assurance directives. DP quality
    management policy in the 1980s exempted DOE nuclear weapons program
    from complying with Departmental quality assurance Orders on the
    basis of equivalency. Then, in 1992, in a memorandum to DOE field
    office managers[^11], the Assistant Secretary for Defense Programs
    declared that DP would comply with DOE Order 5700.6C, discussed
    below, with certain exemptions for classified weapons-production
    work. The Assistant Secretary decided that it would be to DP's
    advantage to be able to demonstrate to oversight organizations,
    e.g., Congress, NRC, EPA and various public interest groups that DP
    did in fact have a systematic, disciplined quality assurance program
    for its weapons production activities.

    In 1992, the DOE Albuquerque Operations Office (AL) issued a new
    standard, QC-2, to complement QC-1 for nuclear weapons research,
    development and testing. Revision 9 of QC-1, issued in late 1998,
    incorporated QC-1 as well as newer quality assurance requirements of
    QC-2 for nuclear weapons research, design, development, procurement,
    production, dismantlement, maintenance, stockpile evaluation and
    disassembly/disposal.

    From its inception in 1954 to the current 2004 revision, QC-1
    policy, principles and requirements have placed maximum
    responsibility and accountability:

-   On nuclear weapons program prime contractors to maintain effective
    quality control systems;

-   On the AEC/DOE to conduct surveillance and acceptance inspections
    that focus on functional quality evidence presented by the prime
    contractors; and

-   On AEC/DOE for verification of this quality evidence.

    In February 2004, the DOE/National Nuclear Security Administration
    (NNSA) issued Revision 10 of QC-1[^12], superseding Revision 9.
    Revision 10 of QC-1 contains the following significant changes for
    that organization that must comply with DOE weapons quality policy:

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-   Change of ownership from AL to DOE/ DOE/National Nuclear Security
    Administration (NNSA) Headquarters DP

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-   New requirement for a management program (QAP) or WQAP to be
    submitted to DOE/NNSA for approval

-   Major emphasis on a risk-based quality management system for
    decision making

-   Grater emphasis on quality metrics

Revision 10 of QC-1 has been restructured along the lines of NQA-1-2000
to ensure that QC-1 could be implemented using the NQA-1 standard.
DOE/NNSA expanded the scope to include weapons work conducted by the
federal organization and the management controls beyond hardware QC.
Another major change made it clear that QC-1 is the DOE/NNSA method for
implementing DOE Order 414.1A and the DOE Nuclear Safety Management
Rule, 10 CFR 830 Subpart A (discussed below). All federal and contractor
work relating to nuclear weapons is now covered by the DOE/NNSA quality
assurance Order, Rule and contract direction.

While Revision 10 of QC-1 adopts a number of NQA-1 requirements, there
are some differences. For example, QC-1 establishes a two-party GOCO
contractual relationship. Also QC-1 includes the following additional
quality assurance requirements:

11. Metrics

    1.  Quality Cost Management

12. Control of Processes

    3.8.1 Process Control Methods

    1.  Government Furnished Material

    2.  NNSA Accepted Material

13. Senior Management Responsibilities

**[AEC Naval Reactors QRC-82C]{.underline}**

Quality control (QC) requirements for AEC naval nuclear propulsion
programs were prescribed in the AEC document QRC-82C[^13]. This document
supplemented MIL-Q-9858A by imposing QC requirements for material
inspection and testing during manufacturing of naval reactor (NR)
components.

[AEC\>DOE RDT F2-2T Standard]{.underline}

From its beginning, the AEC managed and operated its civilian reactor
and technology development programs as a decentralized agency. AEC
Headquarters developed policy, managed funding and issued broad
programmatic direction to its field organizations. The AEC issued grants
to universities, national laboratories and research and development
contractors.

Notwithstanding the good operational safety records in the late 1960s,
AEC Reactor Development and Technology (RDT) management and engineers
were disturbed to note that important civilian reactor and technology
development objective were not being accomplished as planned. Quality
problems, including equipment failures and irretrievable loss of
important data, were attributed not to the inherent risks of technology
development but to insufficient management and engineering attention to
conventional material and process controls. Fundamental, exacting
engineering standards and quality controls that were essential to
technology development were not being applied.

Some early AEC, DOE and contractor project management misconceptions
about quality assurance were :

-   Some project managers believed it was possible to assure nuclear
    facility quality without a formal, documented and integrated quality
    assurance program. While this approach was used for small basic
    research reactors, quality-related operating problems resulted in
    shutting down production reactors at most DOE sites and prevented
    their restart.

-   Some project managers believed that the quality assurance program
    establishment and implementation was the primary responsibility and
    role of the quality assurance organization. This misconception was
    fostered in part by some quality assurance organizations that
    believed that the quality assurance plans, requirements and
    procedures were written by and for the quality assurance
    organization. They failed to recognize a fundamental quality
    assurance principle:

> Quality and its achievement is a primary management responsibility;
> the quality assurance organization supports top and line management in
> executing their quality assurance programs and by conducting
> independent audits.

Prior to 1968, there were no formal quality assurance requirements
imposed by the AEC and its management and operating contractors on GOCO
nuclear facilities conducting reactor development and technology
activities. This situation presented an early quality management
dichotomy for the National Aeronautical and Space Administration (NASA)
quality engineers who had technical and quality management oversight of
some joint NASA/AEC space exploration programs. For example, in the
mid-1960s, NASA's Space Nuclear Auxiliary Power (SNAP) Programs imposed
rigorous quality assurance and quality control requirements from the
NASA quality program standards [^14]on some of its prime contractors,
e.g., Aerojet General in Azusa, California, and General Electric in
Evendale, Ohio, to name a few. NASA delegated to DOD Air Force, Navy and
other contract administration agencies certain DOD inspection system
requirements[^15] for non-nuclear, non-mission-critical components of
the power conversion system being developed by the contractor. NASA
quality assurance program managers were disappointed to learn that the
AEC did not impose any formal quality assurance or/quality control
requirements on the SNAP reactor-system-development contractors. The
NASA Lewis Associate Director brought this situation to the attention of
the joint AEC-NASA organization.

In mid-1968, the AEC Division of Reactor Development and Technology
(RDT) senior managers acknowledged the need for a comprehensive quality
assurance program standard that could be imposed on GOCO reactors,
technology development programs, national laboratories and test
facilities[^16]. RDT management agreed to develop the new RDT standard
that would be designed to address not only the quality program and
inspection system specifications of MIL-Q-9858A and MIL-I45208A but also
the more comprehensive quality and reliability assurance program
requirements of the NASA publication NHB 5300.4(1B). The proposed
standard would provide for planning, management and engineering
activities as well as additional requirements for design review, design
verification testing, qualification testing and development testing.

It is noteworthy that Admiral Rickover's AEC Naval Reactors programs
employed some of these design control elements as a part of their normal
design engineering activities, although they were not called quality
assurance elements. So it was not too difficult to convince the RDT line
managers who came from Naval Reactors that design control was still
their line responsibility under the broad quality assurance umbrella.

With the participation of the major AEC national laboratories and M&O
contractors, RDT Director Milton Shaw authorized in late 1968 a working
group, managed by the author, to develop the new standard. RDT F2-2T was
issued in June 1969 under the RDT standards program managed by the Oak
Ridge National Laboratory (currently managed by the DOE Office of
Environment, Safety and Health).

RDT F2-2T was written in a phased format to be selectively applied to a
facility or project that flowed from initial quality assurance program
planning, through design and development; procurement; manufacturing,
fabrication and assembly; construction and installation; to facility
operation, maintenance and modification, depending on scope of the
quality assurance program activities.

RDT F2-2T covered all of the basic quality assurance criteria of 10CFR50
Appendix B. Also included were quality assurance program requirements
for design descriptions, development testing, engineering studies,
operational readiness reviews, unusual occurrence reporting, data
collection methods, material certification, alloy verification, and
management reviews. Many of these RDT F2-2 quality assurance
requirements had been proven to be effective in earlier NASA flight
system and ground support operations.

RDT F2-2T was a dynamic standard that was fully endorsed and used by RDT
technical and quality assurance managers. At the request of RDT
managers, RDT F 2-2T was amended ten times from 1969 to 1983, to improve
the effectiveness quality assurance program implementation based on user
experience and on unusual occurrences reporting during design,
construction and operation of AEC/DOE numerous reactors and test
facilities. These amendments included:

1.  Purchaser approval of repairs and waivers

2.  Planning and documentation of independent design reviews

3.  Surveillance of facility operations, maintenance, modifications and
    repairs

4.  Preparation of engineering drawing lists.

5.  Purchaser approval of inspection and test plans and establishment of
    mandatory hold points

6.  Control of handling, lifting and rigging activities

7.  Selective application of quality assurance requirements and
    preparation of a quality assurance program index of procedures

8.  Calibration and control of measuring and test equipment

9.  Indoctrination, training and qualification of personnel

10. Identification, reporting and management of engineering holds

In a large construction project, e.g., the Fast Flux Test Facility
(FFTF), special quality assurance requirements and controls were
incorporated through these amendments to RDT F2-2T. The amended
requirements helped to detect, eliminate, or prevent the installation or
use of improperly identified and mixed weld filler metal and many other
substandard materials purchased for FFTF construction.

Amendment 6 resulted from a rash of handling, lifting and rigging
incidents at DOE facilities. RDT also developed a stringent RDT F 15.2
standard on testing and lifting controls, particularly over reactors.

Amendment 7 to RDT F2-2T abolished costly and voluminous quality
assurance program descriptions that merely repeated the contents of
implementing procedures. Instead, users were instructed to prepare a
quality assurance program index including the organization structure and
a listing of quality assurance procedures.

In his memorandum of March 1972 to RDT technical professionals, RDT
Director Milt Shaw called their attention to the promulgation of RDT
F2-2T, which reinforced the many policy statements and related actions
of the Congress, the Commission, the ACRS, and standards-writing groups
over several years on the need to significantly strengthen quality
assurance in conducting reactor and technology development programs,
whether in the national laboratories or in the commercial sector.

In a 1978 memorandum to DOE field office managers[^17], the DOE Director
for Nuclear Energy stated that DOE preferred the quality assurance
program for civilian nuclear energy technology development programs to
be established and implemented in accordance with applicable
requirements of the nationally recognized, voluntary consensus
standards. Unless otherwise directed, or where there was no cost
advantage, major DOE reactor development program were to employ the
N45.2 standard. Where N45.2 was determined to be insufficient for
technology development activities, it was to be supplemented by
appropriate quality assurance requirements.

Acting on the recommendation of an RDT study group under Dan Garland to
endorse a single national consensus standard for reactor and technology
development programs, in April 1985, RDT management canceled and
withdrew the RDT F2-2 standard when it endorsed the ANSI/ASME NQA-1-1983
standard. The shift to the NQA-1 national consensus standard was
consistent with an Office of Management and Budget (OMB) Circular A-119
regarding the use of such national consensus standards.

A comparison of the RDT F2-2 standard with ANSI/ASME NQA-1-1983 revealed
consistencies in their basic quality assurance program elements but
significant differences in their degree of specificity, particularly
when applied to reactor development and testing activities, and in their
format.

By endorsing the NQA -1-1983 standard, RDT relinquished its technical
and quality management ability to make rapid, timely and substantive
changes to quality assurance program requirements based on urgent needs
back by field experience. The ASME NQA-1 consensus process achieved
thorough reviews of draft standards but did not lend itself to the
processing of rapid, program-specific changes.

[DOE Order 5700.6]{.underline}

In March 1978, the DOE Inspector General (IG) advised DOE Headquarters
senior managers that IG inspectors were observing major continuing
deficiencies in formal quality assurance programs at DOE field sites.
These deficiencies included:

-   Inadequate or nonexistent quality standards;

-   Inadequate design control and design reviews;

-   Inadequate supplier controls; and

-   Inadequate fabrication controls.

    These deficiencies were attributed in part to the lack of a strong
    DOE Headquarters quality assurance policy, organization and
    implementing requirements. For many years, the AEC and DOE national
    laboratories and contractors had operated under a system of grants
    for research and development that only required periodic progress
    reports on activities and how funds were being spent. The AEC
    requested but did not direct quality assurance compliance

    In response to the IG advisory, in 1979 the DOE Secretary of Energy
    appointed a study group headed by Phil Coyle of DP, including the
    author, to develop recommendations on a Department-wide quality
    assurance policy and requirements. In January 1981, with the
    concurrence of the DOE Secretary of Energy, DOE Order 5700.6[^18],
    Quality Assurance, was issued.

    The DOE Nuclear Facility Personnel Qualification and Training
    Committee (NFPQT), which was appointed by the Under Secretary of
    Energy after the accident at Three Mile Island in 1979, reported
    that the DOE quality assurance Order was being implemented in
    varying degrees of rigor at most DOE nuclear facility sites.

    *The NFPQT reported that a common deficiency was the lack of
    management controls and attention paid by senior DOE and contractor
    managers to implementing effective quality assurance requirements at
    DOE nuclear facility sites.*

    As a result of the NFPQT report, the DOE Under Secretary of Energy
    developed a DOE Action Plan. A revised DOE Order 5700.6A was issued
    in an attempt to strengthen the Department's dysfunctional quality
    assurance policy. In September 1986, DOE Order 5700.6B only
    addressed changes in DOE Headquarters responsibility for quality
    assurance oversight.

    **[DOE Order 5700.6C]{.underline}**[^19]

    In August 1991, a completely restructured DOE quality assurance
    Order, 5700.6C, was published as a part of a comprehensive DOE
    directives system. This system includes DOE Policies, Manuals,
    Orders, Notices, Regulatory Rules and Guides for quality assurance
    plus a variety of many other subjects, which may be accessed though
    the DOE Internet at
    [http://www.directives.doe.gov](http://www.directives.doe.gov/).
    Information specific to the Department's quality assurance policy is
    available at http://www.directives.eh.doe.gov/qa.

    DOE 5700.6C established Total Quality Management arrangement of DOE
    quality assurance program requirements into three categories and ten
    criteria, as follows:

+-----------------------------------------------------------------------+
| -   **Management**                                                    |
|                                                                       |
| 1.  Program                                                           |
|                                                                       |
| 2.  Personnel Training and Qualification                              |
|                                                                       |
| 3.  Quality Improvement                                               |
|                                                                       |
| 4.  Documents and Records                                             |
+-----------------------------------------------------------------------+
| -   **Performance**                                                   |
|                                                                       |
| 5.  Work Processes                                                    |
|                                                                       |
| 6.  Design                                                            |
|                                                                       |
| 7.  Procurement                                                       |
|                                                                       |
| 8.  Inspection and Acceptance Testing                                 |
+-----------------------------------------------------------------------+
| -   **Assessment**                                                    |
|                                                                       |
| 9.  Management Assessment                                             |
|                                                                       |
| 10. Independent Assessment                                            |
+-----------------------------------------------------------------------+

> **DOE Order 5700.6C reflected the concept that all work is a process
> that can be managed, performed, assessed, and improved, i.e., by
> adopting a management system approach to quality.**

The ten basic criteria of DOE Order 5700.6C provided general quality
assurance requirements for all work to be performed by DOE and its
contractors. The 10 criteria were stated as expected outcomes
(performance-based), rather that as prescriptive "how to" requirements.
The Order also included definitive responsibilities for federal managers
in their oversight roles.

The quality assurance criteria were fairly well understood by DOE and
contractor organizations. However, Criterion 3, Quality Improvement,
Criteria 9, Management Assessment, and Criterion 10, Independent
Assessment, posed a challenge to organizations unfamiliar with
contemporary quality concepts. So, DOE developed an implementation guide
for DOE Order 5700.6C to illustrate the management system concept for
quality; expand on the performance-based criteria; and define acceptable
approaches to implementing the Quality Improvement, Management
Assessment and Independent Assessment criteria.

DOE Order 5700.6C referenced ASME NQA-1, ASME NQA-2, ASME NQA-3 and a
number of DOE and other standards.

Exempted from DOE Order 5700.6C was work associated with DOE nuclear
weapons production, Naval Nuclear Propulsion Programs, NRC licensing,
and research and development (R&D) publications. This Order did not
exempt work associated with design construction, operation, and
maintenance of facilities and equipment used to produce nuclear weapons.

DOE 5700.6C required the use of appropriate standards, such as ASME
NQA-1 (discussed below) to develop and implement quality assurance
programs. The Order stressed the following three quality management
principles:

1.  Senior DOE and contractor managers are responsible for quality
    assurance program management, implementation, assessment, and
    improvement;

2.  Line organizations achieve quality; and

3.  Overall performance is measured and evaluated using a rigorous
    assessment process.

The following 12 underlying principles and actions were the basis for
the DOE 5700.6C performance-oriented quality assurance criteria:

1.  Define policies and objectives -- ensure they are understood and
    accepted.

2.  Specify roles and responsibilities -- ensure they are understood and
    accepted.

3.  Specify and communicate expectations -- identify and allocate
    resources to achieve them.

4.  Strive to continually improve quality objectives.

5.  Ensure people are competent at the work they perform.

6.  Ensure the right people have the right information at the right
    time.

7.  Seek and use relevant experience.

8.  Plan and control work.

9.  Use the correct materials, tools and processes -- control changes to
    them.

10. Assess work results to ensure it meets what is required and exceeds
    customer expectations.

11. Identify and remedy errors and deficiencies.

12. Periodically review management processes to improve their
    effectiveness and efficiency.

Whereas previous DOE quality assurance directives were applied only to
contractors, DOE Order 5700.6C required both the DOE and contractor
organizational elements to develop and implement quality assurance
programs that complied with the ten criteria.

Attachment 1 to DOE Order 5700.6C provided guidance for developing and
implementing quality assurance programs to satisfy the ten quality
assurance criteria. DOE national laboratories and contractors were
required to prepare and submit for DOE evaluation and approval a Quality
Assurance Program to describe how each organization would comply with
the applicable criteria of DOE Order 5700.6C. Where an organization was
not in compliance, an Implementation Plan was required to describe the
actions and schedules for achieving compliance.

**[DOE Orders 414.1 and 414.1A]{.underline}** [^20]

## In November 1998, DOE issued a new draft quality assurance directive, DOE Order 414.1[^21], for review and comment. The new Order superseded DOE Order 5700.6C. It redefined the scope in conjunction with the new DOE Nuclear Safety Management Rule 10 CFR 830 and subparts 830.120 and 122 on quality assurance discussed below. The Order applied to both DOE and contractor organizations. 

## DOE Orders 414.1 and Revision 414.1A[^22] were developed in coordination with the DOE-chartered Quality Assurance Topical Standards Committee (QA TOPCOM) and the Quality and Safety Management Special Interest Group (QSMSIG). These committees provided an avenue for Gustave (Bud) Danielson, the DOE Quality Policy Manager, to gather broad input from national/international standards bodies, standards users and users of the Order and Guides; Danielson charted these committees, which were composed of representatives of DOE, national laboratories, and contractors with expertise in quality assurance, assessments and management. The QM/SIG has since been disbanded.

##  {#section-5}

## Attachment 1 of DOE Order 414.1A was a contractor requirements document (CRD) that included all of the general quality assurance requirements and the ten quality assurance criteria. For example, the CRD required submittal of a contractor's quality assurance program document to a designated DOE official for approval. The CRD was developed by DOE for attachment to all of its Orders and is intended to be a stand-alone document suitable for use directly in a contract. As such, it does not include section from the Order that only apply to federal organizations (e.g., DOE Headquarters and field office responsibilities.)

##  {#section-6}

## Attachment 2 of DOE Order 414.1A contained supplemental quality improvement requirements for corrective action plans for significant safety issues resulting from DOE Office of Oversight reports and other issues as specified by the Secretary of Energy. This new Corrective Action Management Program was developed in response to the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 98-1. 

General guidance on implementation of DOE Order 414.1A was published in
DOE G 414.1-2. This superseded the original guidance in Attachment I to
DOE Order 5700.6C.

**[DOE Order 414.1B]{.underline}**[^23]

DOE Order 414.1B supersedes DOE O 414.1A. Its purpose is to ensure that
DOE National Nuclear Security Administration (NNSA) products and
services meet or exceed customer expectations. This objective is to be
achieved based on the following principles:

1.  Quality is assured and maintained by a single, integrated, effective
    quality assurance program, i.e., a management system;

2.  Management support for planning, organization, resources, direction
    and control are essential to quality assurance;

3.  Performance and quality improvement require thorough, rigorous
    assessment and corrective action;

4.  Workers are responsible for achieving and maintaining quality; and

5.  Environmental, safety and health risks and impacts associated with
    work processes can be minimized while maintaining reliability and
    performance of work product.

DOE Order 414.1B requirements applied to NNSA and contractor
organizations, except for the exclusion of NNSA Naval Reactor Programs.
The Order referenced ASME NQA-1-2000 for nuclear-related activities and
ANSI/ASQ Q 9001:2000 for non-nuclear activities. Thus, a DOE
contractor's quality assurance program was (and is still) expected to
use the appropriate American national or international consensus
standard, where practicable and consistent with contractual and
regulatory requirements. Attachments 1and 2 contained essentially the
same ten basic quality assurance criteria taken from the superseded DOE
Orders 5700.6C and 414.1A for DOE and contractor organizations,
respectively; Attachment 3 describes the DOE-wide process for
controlling suspect/counterfeit items (S/CI). The Order updated S/CI
requirements based on field experience and was supported by a new guide,
DOE G 414.1-3 [^24] for controlling suspect/counterfeit items.

**[DOE Rule 10 CFR Part 830]{.underline}**

The Price Anderson Amendment Act (PAAA) of 1988 amended the Atomic
Energy Act of 1954 by providing broad, mandatory, indemnification
coverage to all persons, including DOE contractors, subcontractors, and
suppliers, whose activities as related to DOE nuclear facilities might
result in public liability claims under the PAAA.

In December 1991, the DOE published a proposed Rule[^25], 10 CFR PART
830, in the Federal Register, to provide basic requirements for ensuring
nuclear safety at DOE facilities. The initial version of the Rule,
issued in 1994, included quality assurance requirements drawn from DOE
Order 414.1A. It was later revised on January 10, 2001, to add safety
basis requirements and integrate with the DOE Safety Management System
Policy. This version was issued as a final Rule that was effective on
February 9, 2001. The Quality Assurance Rule is now defined in 10 CFR
830, Subpart A, Quality Assurance.

Unlike the contractually enforced DOE Orders, 10 CFR Part 830 provides a
basis for enforcement and for civil and criminal penalties under the
authority established by the PAAA. Information on the Department's
nuclear safety enforcement program is available at
http://eh.doe.gov/enforce/.

> **The DOE Safety Management Rule, 10 CFR Part 830, had finally
> established fully enforceable quality assurance rules with civil and
> criminal penalties for its nuclear facilities similar to the NRC's 10
> CFR 50 Appendix B for commercial nuclear facilities.**

**[DOE Order 414.1C]{.underline}**

The latest revision of the DOE quality assurance Orders 414C[^26],
superseding DOE Order 414.1B, includes new requirements and guidance for
nuclear facility safety software[^27]. The Order invokes ASME NQA-1-2000
for implementing these new requirements. DOE has developed a companion
guide, G 414.1-4, for implementation of the new requirements, which
further describes the requirements, and use of ASME NQA-1-2000. More
information on safety software is available at http://eh.doe.gov/sqa.

**[Implementation Guides]{.underline}**

As previously mentioned, DOE and its contractors have had various
interpretations on how to implement the independent and management
assessment requirements of 10 CFR Part 830.120 and DOE 5700.6C.
Therefore, DOE developed and issued a guide[^28] under Danielson's
direction, on the purpose, types, planning, conduct and reporting of
assessments. Later, Danielson managed the development, issuance and
updating of an implementation guide for 10 CFR 830.120 and DOE Order
414.1[^29], plus new guides for S/CIs and safety software noted above.

One might have concluded that the ten nuclear quality assurance criteria
of DOE Order 5700.6C tended to be masked and diluted by a more
contemporary Total Quality Management (what) format. This argument would
explain the need for DOE to develop attachments and guidance discussed
above on how to implement the Orders and Rule. It would also explain the
need for supplemental requirements and guidance found in the ASME
NQA-1-2000 discussed in SECTION IV below.

**[DOE Action Plan]{.underline}**

In July 2005, the DOE published an action plan[^30] addressing ten
lesson learned from the NASA Columbia Space Shuttle catastrophe of
February 2003 and Davis-Besse pressure vessel head corrosion event
discovered in March 2002 and their applicability to the DOE complex. The
primary focus of the working group preparing the DOE Action Plan was on
nuclear operational safety in response to a DNFSB Recommendation 2004-1.
None of the DOE Action Plan items related specifically to ways of
improving NQA preventive actions.

Some of the lessons identified from these occurrences are summarized and
modified here for their potential NQA applicability:

[Operating experience]{.underline}: Individuals and organizations need
to learn lessons from operating experience to avoid repeating errors and
improving performance.

[External pressures:]{.underline} Budget and schedule priorities must
not override safe and reliable operation decisions.

[Focus on planning and prevention:]{.underline} Safety and quality
efforts should focus more on planning and preventive actions in addition
to investigations and corrective actions after an accident or unexpected
occurrence.

[Technical inquisitiveness:]{.underline} Managers need to encourage
employees to freely communicate safety and quality concerns and
differing professional opinions.

[Complacency:]{.underline} Management must guard against a
self-satisfying attitude brought on by good performance metrics and past
safety records.

[Normalizing deviations:]{.underline} Routine departures from
established standards should not be allowed to create a low-probability
event to occur.

**SECTION II.**

AEC\>NRC REGULATIONS

*This section describes AEC/NRC regulatory initiatives to the extent
that they have had a direct impact on the development of nuclear quality
assurance regulations, standards and guidance on their implementation.
It does not describe the interrelationships of 10 CFR Part 50 Appendix B
with other quality assurance-related regulations such as Parts 70, 71
and 72.*

**[Early Nuclear Power Plants]{.underline}**

Recognition of the need for and the adoption of effective quality
assurance regulations, standards, and guidance on their implementation
did not come rapidly nor voluntarily to designers, constructors,
operators, and the AEC/NRC regulators of licensed nuclear facilities. In
early commercial nuclear power plants, as well as in AEC-owned reactors,
technology development programs and test facilities, quality was
achieved and verified with a minimum of formal, documented practices and
procedures. First-generation commercial nuclear power plants and
AEC-owned nuclear facilities were relatively small in size and capacity,
remotely located, and relatively simple in their design, construction
and operation practices. Close-knit teams of highly competent people
were in charge of every aspect of design, construction and operation.

It was not uncommon for early commercial nuclear power plants to be
designed and constructed as "turnkey" projects by the reactor suppliers
and engineer-constructors. Thus, many of the smaller utility owners had
minimal involvement in and technical knowledge of, early nuclear plant
designs and construction practices until such time that the constructed
nuclear plants were formally turned over to them for operation. Some
utilities were focused on producing electricity, and aside from
generating costs, did not really care whether the plant was run on coal,
oil, gas, hydroelectric or nuclear power. Furthermore, except for
applying the quality assurance provisions of the ASME Boiler and
Pressure Vessel Code, there was little interest in developing and
applying unique quality assurance standards, per se, to other nuclear
safety-related components and activities\...

In March 1967, Vice-Admiral Rickover offered his advice on designing,
constructing and testing to purchasers of central station nuclear power
plants under the popular "turnkey" arrangement[^31]. For example, he
would require the "seller" to define the standards to be used for
design, material, fabrication, etc. He would require the "purchaser" to
retain an independent organization to check and audit all phases of
design and construction. He would require the "seller" to guarantee that
the plant would perform reliably: satisfactory performance of equipment
for one year and 95% availability

As nuclear facilities grew in size, complexity, and number, as less
experienced people became involved, and as their differences from
non-nuclear power plants were fully recognized, it became apparent that
a more systematic, disciplined, engineering approach to assuring nuclear
facility quality was urgently needed.

Sufficient experience in designing and operating first and second
generation nuclear plants had not been accumulated until the early
1970s, when the level of nuclear plant construction activities really
exploded, to prove conclusively that the nuclear industry and the AEC
had no option but to pay the price for more exacting quality standards,
regulations and more disciplined work practices for implementing them.
Major cultural and systemic changes throughout the industry were of
paramount importance if the nation was to succeed in the development and
commercialization of safe and reliable nuclear power.

Accordingly, the AEC undertook certain long-term regulatory initiatives
to formalize quality assurance programs and standards for the licensing
of commercial nuclear power plants. Among these initiatives were the:

1.  Developing quality assurance regulations and safety guides.

2.  Implementing the broad definition of quality assurance expressed by
    Commissioner Ramey that transcended the traditional manufacturing
    quality control and inspection concepts

3.  Providing leadership, financial and technical assistance to an
    infrastructure dedicated to developing urgently needed national
    consensus quality assurance standards.

**[AEC 10 CFR 50 Appendix A]{.underline}**[^32]

In 1965, the AEC issued a press release designating a series of 27
general design criteria for light-water-cooled nuclear power plants.
These criteria were published for public comment in 1967 as Appendix A
to 10 CFR50. They were greatly expanded to 55 and then to 64 criteria
and made effective as an AEC regulation in 1971. Criterion 1 of Appendix
A, \"Quality Standards and Records\", contains three fundamental quality
assurance requirements:

1.  Identification, evaluation and use of appropriate quality standards;

2.  A quality assurance program to assure that structures, systems, and
    components performed their safety functions; and

3.  Maintenance of appropriate records.

Dresden 2 was the first nuclear power plant to be governed by Appendix A
to 10 CFR50.

One might have argued that the three fundamental requirements of
Criterion 1 of 10 CFR Part 50 Appendix A together was a sufficient
regulation for quality assurance in the design and construction of
nuclear power plants; however, on-site AEC regulatory inspections and
audits in the 1950s and 1960s at commercial nuclear power plants belied
this argument by revealed numerous repetitive generic design and
constructions deficiencies, including:

-   Inadequate review of detailed designs

-   Inadequate quality provisions in purchase specifications

-   Inadequate control of suppliers

-   Inadequate process controls for shop and field work

-   Inadequate control of materials handling and lifting operations

-   Inadequate construction inspection

-   Inadequate quality records and their control

    To cite a few specific examples, AEC inspections at Oyster Creek
    found continuing quality-related problems in control rod drives,
    steam separator and cracks in the core shroud and supports. At Big
    Rock Points, AEC inspectors found defects in fuel assembly welds. At
    Ginna, AEC inspectors found equipment and personnel hatch frame
    buckling during concrete placement. The AEC developed internal
    inspector training programs and inspection procedures. An ongoing
    discovery was that off-the-shelf commercial products were not always
    of sufficient quality for nuclear plant service conditions.

    Prior to late 1968, however, quality assurance requirements,
    standards and implementing procedures or the lack thereof was not a
    major licensing issue of the AEC, the ACRS and the ASLB at CP
    hearings.

    Eventually, it became increasingly apparent to the AEC inspectors
    and regulatory staff and to the nuclear utility industry that, as
    the number of applications for nuclear power plant construction
    permits and operating licenses grew and quality problems were being
    discovered at plant sites, more definitive quality assurance
    regulations, standards, and guidance on their application were
    needed beyond Appendix A to 10 CFR part 50. More importantly, a
    major paradigm shift was needed to counter a limited construction
    management attitude that, if the AEC/NRC inspector did not find a
    quality problem, it didn't exist and must not be reported.

    **[Zion Station Hearings Impact]{.underline}**

    On September 17, 1968, the lack of definitive quality assurance
    regulations, standards, and guidance caused the Atomic Safety and
    Licensing Board (ASLB) to suspend public hearings on Commonwealth
    Edison\'s (ComEd) Zion Station Units 1 and 2. The ASLB refused to
    issue a CP by stating:

    ***\"In the opinion of the Board, Commonwealth Edison has not
    presented sufficient evidence pertinent to the provisions that
    should be made for the assurance of the control of quality needed
    for the technology and disciplines of the nuclear reactor field, nor
    has the AEC staff submitted evidence by way of criteria or expert
    testimony adequate to permit a judgment of its evaluation of the
    quality control program\".***

    The ASLB members who presided over the public hearings were:

```{=html}
<!-- -->
```
-   S. Jensch

-   J. Buck

-   S. Forbes

-   Notwithstanding the ACRS asking about but receiving no commitment on
    "methods of quality control" during the previous Turkey Point 3&4 CP
    review in 1967, the Zion hearings was a first and major milestone in
    the history of nuclear power quality assurance program commitment,
    requirements and eventual regulation in the United States. It had
    the potential for giving the highest visibility to the development
    and enforcement of quality assurance program regulations and
    standards against which the regulatory staff and the applicants
    could mutually judge the adequacy of the latter's quality assurance
    programs.

    When the ASLB hearings were suspended, ComEd contracted with the
    Aerojet General Corporation in Sacramento, California, to develop a
    quality plan and procedures. Concurrently, Wilbur (Bill)
    Morrison[^33] of the AEC Division of Reactor Standards (DRS) and
    Gene Langston of AEC/RDT were tasked by the DRS Director to began
    developing a set of quality assurance program criteria to judge the
    adequacy of ComEd's Zion Station quality assurance program. Morrison
    and Langston extracted applicable provisions from MIL-Q-9858A, NASA
    NHB 5500.4(1B) and RDT F2-2T quality assurance documents into a
    series of draft DRS nuclear power plant quality assurance criteria:

```{=html}
<!-- -->
```
-   10/3/68: First draft, 15 criteria A --P.

-   11/12-20/68: Second draft, 20 criteria, 1 -- 21, added purpose,
    definitions, program, control of special processes, mandatory
    inspection hold points and test control.

    On November 26, a joint AEC meeting was held to discuss the scope,
    timing and criteria that had been developed for a special on-site
    quality assurance inspection or survey of Common Edison's quality
    assurance program for Zion reactors.[^34] Attending the meeting held
    in Bethesda, Maryland, were:

```{=html}
<!-- -->
```
-   Chuck. Long, DRL

-   Bill Morrison, DRS

-   Gene Langston, RDT

-   Harry Thornburg, CO

-   Gerry Hadlock, OGC

The Zion reactors construction site was visited on December 3, 1968 by
the inspection team. Commonwealth Edison's engineering and management
offices in Chicago, IL, were visited on December 3, 1968. Chicago Bridge
and Iron Company, fabricator of containment liner plates, was visited on
December 4, 1968. The team met on December 6, 1968, in Bethesda to
correlate inspection results and prepare a report for the AEC and ASLB.

# When the ASLB reconvened hearings on December 17, 1968, one ASLB mentioned he had witnessed a concrete form failure at Dresden and questioned whether a quality assurance program could have prevented this failure. ComEd management responded that it would be doubtful.

#  {#section-7}

# Morrison and Langston testified as expert witnesses for the AEC on the adequacy of Commonwealth Edison's quality assurance program. Salient observations during the ASLB hearings were:

1.  ComEd's quality assurance program had not been very formal or well
    documented prior to this becoming an issue before the Board.

2.  ComEd management contended, without substantial opposition, that the
    quality assurance program was as formal and as well documented as
    other cases previously reviewed and approved by the Board.

3.  Based on the AEC team's limited on-site survey of the Zion Station
    construction site and review of ComEd's Quality Assurance Plan,
    which emphasized documentation and verification, when evaluated
    against the 15 drafts AEC quality assurance criteria, the AEC staff
    concluded that the applicant had developed and documented a
    philosophically acceptable quality assurance plan.[^35]. The CP was
    issued to ComEd for the Zion Station.

4.  It was recognized, however, that (1 the plan had not been fully
    implemented and (2 that some quality assurance elements had not been
    full investigated, and (3) changes to the plan and implementing
    processes and procedures would be needed as experience in their
    implementation was gained.

5.  The applicant committed albeit reluctantly to fully implement the
    Zion Quality Assurance Plan by all contractors working in the site
    by January 1, 1969, and implementing procedures by February 1, 1969.

6.  It should have been more apparent to the AEC Commissioners and the
    nuclear power industry that quality assurance would become an
    increasingly important consideration in the AEC licensing reviews in
    future hearings, especially for plants in the proximity of
    metropolitan areas.

# **Quality Assurance Redefined** 

#  {#section-8}

While NASA and RDT had previously established performance-focused
quality assurance program standards and definitions, other early
government quality standards, such as MIL-Q-9858A and QC-1 were focused
on compliance with contractual quality control and acceptance inspection
systems requirements. 10 CFR 50 Appendix B adopted a more encompassing
definition of quality assurance, which was a slightly modified version
of AEC Commissioner Ramey's:

> \"**All those planned and systematic actions necessary to provide
> adequate confidence that a structure, system, or component will
> perform satisfactorily in service.\"**

This definition infers, for example, that design criteria and standards
are adequately defined and correctly translated into design documents;
that competent persons execute the design in accordance with the design
documents; that tests confirm the design; perform construction and that
the plant is operated and maintained within safe limits established by
the design.

The 10 CFR 50 Appendix B definition of quality assurance included
quality control:

> **"*Those quality assurance actions which provide a means to control
> and measure the characteristics of an item, process, or facility to
> established requirements."***

**[10 CFR 50 Appendix B]{.underline}**

# Following the Zion Station hearings of the ASLB on December 17, 1969, representatives of the AEC DRS and RDT staff continued to develop the quality assurance program criteria for nuclear power plants. Criteria were added and modified for the control of special processes, the control of design, inspection and test and other quality-affecting activities. The list of criteria grew to 21 and finally stabilized at 18.

The proposed AEC 10 CFR 50 Appendix B would require applicant for
licenses to include in preliminary safety analysis report a description
and evaluation of the quality assurance program to be applied to the
design, fabrication, construction and testing of structures, systems and
components of the facility. The proposed 10 CFR 50 Appendix B also would
require that the final safety analysis report contain information
concerning measures to be taken to assure safe operation of the
facility, including such things as management and administrative
controls and plans for operations and maintenance, surveillance, and
periodic testing of structures, systems and components.

The proposed criteria would apply to all structures, systems and
components of nuclear power plants that prevent or mitigate the
consequences of accidents, which can cause undue risk to public health
and safety. The requirements would apply to all activities affecting the
safety-related function of these structures, systems and components
throughout the design, construction and operation phases. Specific
activities covered in these phases would include designing, purchasing,
fabricating, handling, shipping, storing, cleaning, erecting,
installing, inspecting, testing, operating, refueling, maintaining,
repairing, and modifying.

On January 3, 1969, the DRS Director submitted to the ACRS for their
review a draft of the quality assurance program criteria for nuclear
power plants.

Concurrently, the DRS Director provided the draft quality assurance
criteria to RDT for review and comment. RDT formed a steering committee
whose members represented the AEC national laboratories and contractors.
The membership included:

-   Joe Anderson

-   Gordon Beer

-   Jim Bell

-   Stuart Knight

-   Gene Langston (Chair)

-   Jack Norris

-   Herb Ross

-   Ralph Seidensticker

-   Al Squires

    The RDT steering committee met on January 14 and 15, 1969, with W.
    Morrison and others of the DRS staff. RDT provided numerous comments
    and recommendations on the draft quality assurance criteria for
    nuclear power plants:

    The RDT steering committee had also recommended adding two new
    criteria, one for training and another for operation and maintenance
    control that were not accepted by DRS.

    Following the AEC Commissioners' review on March 24, 1969 the AEC
    published the proposed 18 quality assurance criteria of Appendix B
    to 10 CFR 50 in the Federal Register on April 17, 1969 for public
    comment 50[^36].

    During the 15 month-period for public comment, the newly formed
    N45-3 Subcommittee reviewed and provided numerous constructive and
    critical comments on the proposed 10 CFR 50 Appendix B amendment.
    These comments were sent to the AEC regulatory staff for resolution
    and to RDT for information[^37].

    Some of the nuclear industry comments criticized the proposed
    Appendix B quality assurance criteria by saying they contained
    prescriptive requirements instead of general criteria for judging
    the adequacy of an applicant's quality assurance program. Other
    comments criticized an overemphasis on the mechanics and techniques
    of meeting the proposed criteria, as opposed to simply defining
    criteria and leaving implementation details to applicants. Another
    criticism was that the criteria were rigidly tied to the specific
    heading, some of which were seemingly redundant, e.g., Sections IV,
    V and VII for document control. The requirement for exhaustive
    documentation was perceived as resulting in too much attention to
    records to the detriment of performance.

    Because of the numerous nuclear industry and RDT steering committee
    comments, the AEC DRS staff made the following changes to the
    proposed quality assurance criteria of 10 CFR 50 Appendix B:

    **Section III, Design Control[,]{.underline}** was extensively
    revised "to (a) provide provisions to assure that the appropriate
    quality standards are included in design documents and that
    deviations from such standards are controlled, (b) require that
    measures are established for the selection and review for
    suitability of application of materials, parts, equipment and
    components; (c) indicate that design control may include methods of
    verifying or checking the adequacy of design other than the
    performance of design reviews, such as the use of alternate or
    simplified calculational methods or the performance of a suitable
    testing program; and (d) require that design changes be subject to
    design control measures commensurate with those applied to the
    original design".

    **Section IV, Procurement Document Control,** was modified "to
    recognize that all sections of the quality assurance criteria might
    not be applicable to all contractors or subcontractors".

    **Section V, Instructions, Procedures, and Drawings,** was revised
    "to make clear that the criteria for determining that important
    activities have been satisfactorily accomplished need not be
    duplicated on more that one design document".

    **Section VII, Control of Purchased Material, equipment, and
    Services**, was expanded "to require that documentary evidence that
    such items conform to procurement requirements shall be available at
    the nuclear power plant site prior to installation or use".

    Section VIII, Identification and Control of Materials, Parts, and
    Components, was revised "to eliminate the implication that
    traceability of items id required in all cases".

    **Section VIII, Identification and Control of Materials, Parts, and
    Components,** was revised to eliminate the implication that
    traceability is required in all cases.

    **Section X, Inspection,** was revised "(a) to eliminate the
    implication that in-process inspection and mandatory hold points
    are, in all cases, required and (b) to indicate that the inspection
    program shall be established and executed by or for the organization
    performing the inspection activity, and that inspection shall be
    performed by individuals other than those who performed the activity
    being inspected".

    **Section XIV, Inspection, Test, and Operating Status,** deleted the
    requirement for marking nonconforming items "to eliminate
    duplication with the requirements of Section XV". The section was
    also revised "to indicate that tagging valves and switches is one
    way to identify the operating status, but not necessarily the only
    way".

    **Section XVI, Corrective Action**, was revised "to preclude the
    necessity of corrective action measures for those conditions adverse
    to quality which are rarely completely eliminated, such as all weld
    defects prior to final or acceptance inspection. The requirement
    that the cause be determined and corrected to preclude repetition"
    was changed "to apply to significant conditions adverse to quality".

    **Section XVIII, Audits,** "to avoid the implication that personnel
    performing audits should be qualified to specified requirements, the
    term 'qualified personnel' was changed "to: appropriately trained
    personnel".

    On November 18, 1969, Edson Case, AEC DRS Director, addressed the
    ASME Winter Meeting on the proposed Appendix B quality assurance
    criteria for the design of nuclear power plants.[^38] He remarked
    that many of the concepts of quality assurance were being applied in
    the design process of nuclear power plants without the designers
    identifying them as being elements of a quality assurance program.
    For this reason, the application of quality assurance to the design
    and operation phases may have appeared to many organizations to be
    more of a new concept in the nuclear power industry than it really
    was. Nevertheless, he believed there were also important quality
    assurance elements that were not being applied routinely to the
    design, as they should have been. It was time to recognize that
    quality assurance during design was just as important as quality
    assurance during construction. In addition to design review by an
    independent organization, there were other design control
    techniques, including the use of alternate simplified calculation
    methods and prototype testing to confirm design adequacy. He noted
    that design organization interfaces and design change control were
    also important design control measures.

    When 10 CFR 50 Appendix B was issued as a regulation in June
    1970[^39], only 12 nuclear power plants had operating licenses;
    numerous other plants were in varying stages of their applications
    for construction permits and operating licenses. Therefore, older
    plants had no commitment in their licensing applications to
    implement the proposed quality assurance criteria of 10 CFR 50
    Appendix B. After Appendix B was issued, an applicant had to commit
    in a licensing application to its quality assurance criteria.
    Eventually, the AEC regulatory staff obtained commitments to the
    quality assurance criteria of 10 CFR 50 Appendix B for the
    previously licensed nuclear power plants.

    When 10 CFR Section 50.34(a)(7) became mandatory in 1970, it
    required an applicant for a license to build a nuclear power plant
    to submit a description of its quality assurance program per the
    criteria of 10 CFR 50 Appendix B for the design and construction
    phase in a section of a Preliminary Safety Analysis Report (PSAR) or
    in a topical report. The regulatory staff performed a desktop review
    of an applicant\'s quality assurance program description prior to
    issuing a construction permit.

    In 1973, the AEC Director of Regulation announced a revised
    procedure that provided for a more substantive review by the
    licensing staff of the applicant's quality assurance program
    description for design and procurement activities and a site
    inspection by the compliance staff to verify the applicant\'s
    implementation of the quality assurance program as described in the
    construction permit.

    The ASLB prompted this action during the Consumer Power's Midland
    station's public licensing hearings in March 1973 by stating that:

    **\"*No QA program is self-executing. Thus, irrespective of how
    comprehensive it may appear on paper, the program will be
    essentially without value unless it is timely, continuously, and
    properly implemented.\"***

    The Board decided that the AEC must do more than a simple desktop
    review of the applicant's quality assurance program to determine
    whether the Appendix B requirements were met and were being
    implemented.

    Regarding scope, the introduction to 10 CFR 50 Appendix B states
    that it applies explicitly to activities affecting the
    safety-related functions of those structures, systems, and
    components that could cause undue risk to the health and safety of
    the public. It applies to activities including designing,
    purchasing, fabricating, handling, shipping, storing, cleaning,
    erecting, installing, inspecting, testing, operating, maintaining,
    repairing, refueling, and modifying.

    The comparability of Appendix B with MIL-Q-9858 and NASA NHB
    5300.4(1B) is illustrated in TABLE I.

    **[Fuel Reprocessing Plants]{.underline}**

    On Marc 8, 1971[^40], The AEC Director of Regulation proposed to the
    Secretary of the AEC for public comment, an amendment to 10 CFR 50,
    which was requested by the Director DRS, to apply the same quality
    assurance requirements of 10 CFR 50 Appendix B for nuclear power
    plants to the design, construction and operation of those
    structures, systems and components and activities of fuel
    reprocessing plants that prevent or mitigate the consequences of
    accidents which could cause undue risk to the health and safety of
    the public.

    Appendix B to 10 CFR 50 was changed to "Quality Assurance Criteria
    for Nuclear Power Plants and Fuel Reprocessing Plants".[^41]

    The AEC Division of Compliance and OGC concurred in this proposed
    amendment. RDT also concurred in this recommendation with the
    comment that RDT would expect that the Commission rules would be
    further amended to include the same quality assurance requirements
    for plutonium processing and fabrication plants when they are
    subject to licensing requirements similar to those for reactors and
    reprocessing plants. This comment was to be taken into account in
    developing a proposed amendment to 10 CFR Part 70 to provide for
    pre-construction review of plutonium processing and fabrication
    plants.

    **[Appendix B Criterion 1 Amendment]{.underline}**

###  {#section-9 .unnumbered}

On April 19, 1974, the AEC Director of Regulation proposed a rule change
to Criterion I, Organization, of 10 CFR 50 Appendix B to be published in
the Federal Register. The rule change was published in January 1975.

Rapid growth in the number of nuclear power plants being planned or
built had resulted in the entry of many new organizations into this
field. In addition, there had been significant changes in the management
arrangements under which nuclear power plants were being built, which
had resulted in undertaking of project management functions by
organizations having limited experience in the nuclear field. To assist
applicant and licensees in the development and implementation of their
quality assurance programs, the AEC considered it appropriate to present
the Commission's requirements with respect to quality assurance in more
detail to supplement the requirements of Criterion I. The development of
these requirements took into account the experience accumulated in
designing, constructing and operating licensed nuclear power plants and
Commission-owned reactors.

The change under consideration by the AEC would require an organization
assigned responsibility for assuring that a quality assurance program is
established and executed and for verifying that an activity has been
correctly performed must have sufficient authority and organizational
freedom to initiate, recommend, or provide solutions, and to verify
implementation of solutions. This provision is similar to subsection
1.B.201 of NASA's quality assurance standard NHB 5300.4(1B).

### [Ford Amendment Study]{.underline}[^42]  {#ford-amendment-study .unnumbered}

In 1984, the NRC reported on a major study that had been made at the
request of Congress on the improvement and assurance of quality in the
design and construction of nuclear power plants. Referred to as the Ford
Amendment study, NUREG 1055 provided valuable insight into and lessons
learned from the failures and successes of nuclear power plant design
and construction projects.

The study described quality assurance case studies of failures that had
been made in nuclear power plant construction projects in the 1970s at
Diablo Canyon, Marble Hill, Midland, South Texas and Zimmer sites. Major
quality-related problems that contributed to the failures encountered at
these sites might have been avoided if the project managers and the NRC
had implemented more rigorously effective quality assurance and
regulatory inspection programs.

***An important lesson to be learned from this study was that there is a
level of change actions -- technical, regulatory and procedural --
beyond which any project management structure can no longer effectively
implement its quality assurance program*.**

In some construction projects, there was a tacit delegation by senior
management of the responsibility for the ***achievement*** of quality to
the NRC-required quality assurance organization whose responsibility was
to assist in the ***assurance*** of quality.

Perhaps the most disturbing finding in the report was the harassment and
intimidation of quality assurance personnel at one or more construction
sites.

According to this study, the essential quality-related principles of a
successfully managed commercial nuclear power plant projects were:

-   Top-down project management commitment to quality is a principal
    project objective;

-   Understanding by top management of the magnitude, complexity and
    difficulties in designing and constructing a nuclear power plant
    compared to a conventional steam plant, and the importance of
    applying exacting engineering and quality standards;

-   Prior nuclear facility design and construction experience of key
    project personnel;

-   Quality assurance is implemented as an integral part of a
    comprehensive management control system; and

-   Prompt detection, communication and correction of quality problems
    in design, effective management oversight of the design process, and
    the ability to control plant configuration and manage change.

    ***Although often criticized as being too prescriptive and not
    sufficiently performance based, 10 CFR 50 Appendix B has withstood
    the test of time during over thirty-plus years of application.***

**[Three Mile Island]{.underline}**

In the aftermath of the 1979 accident in Unit 2 at the Metropolitan
Edison's Three Mile Island nuclear power plant site southeast of
Harrisburg, Pennsylvania, President Jimmy Carter appointed the Kemeny
Commission to conduct a comprehensive investigation of the accident and
make recommendations based on its findings. The Kemeny Commission
report, issued in October 1979, traced the cause of the accident to
maintenance errors resulting in, a stuck pressure-relief valve, the loss
of crew operational awareness and operator errors that resulted in major
core damage, and the unavailability of onsite technical support.

**SECTION III.**

ANSI\>ASME N45.2 STANDARDS

*This section describes the development of ANSI and ASME N45.2 quality
assurance program standards and its "daughter" standards from 1969 to
1977, the AEC Rainbow" series of guidance documents, and the
safety/regulatory guides that endorsed the N45.2 standards.*

**[Recognition of Standards Needs]{.underline}**

By the late 1960s, it became evident to the AEC and the nuclear industry
that regulations alone were not the most desirable or appropriate way to
define management and technical practices for designing, constructing
and operating nuclear facilities and their components. Using the ANSI
national consensus standards process would permit experts from
government, industry, national laboratories, and other public
institutions to contribute to the definition of these practices.

> **Bob Minogue of the AEC DRS staff once remarked: "I went to the codes
> and standards cupboard and it was bare!"**

With financial backing, significant participation and strong endorsement
from the AEC plus industry leadership, a plan for a whole body of
urgently needed national consensus standards was formulated. This plan
identified and assigned responsibility for the development of a variety
of consensus standards to the appropriate technical societies and other
standards-writing organizations under the American National Standards
Institute (ANSI).

In early 1969, a joint ANSI steering committee consisting of
representatives of the ASME and other technical societies identified,
among other things, the following seven quality assurance-related
standards topics for the construction phase of nuclear power plants:

1.  Pressure System Cleaning

2.  Packaging, Shipping, Receiving, Handling, Storage

3.  Housekeeping (total plant)

4.  Installation, Inspection and Testing - Electrical and
    Instrumentation

5.  Inspection and Testing - Structural Steel and Structural Concrete

6.  Qualification of Personnel

7.  Quality Assurance Program Requirements

Six of these standards were to cover specific work practices associated
with construction and possibly manufacturing activities. The seventh
standard on quality assurance program requirements was significantly
expanded to cover the total range of activities affecting the quality of
nuclear power plant structures, systems and components, from initial
design through construction and operation, exclusive of those
structures, systems and components covered by the ASME Boiler and
Pressure Vessel Code. It would be applicable to the plant owner and
major participating contractors at every level of the plant construction
project.

The ANSI N45 Committee on Reactor Plants and Their Maintenance,
sponsored by ASME, was assigned five of the seven standards, including
the quality assurance program requirements standard.

#  {#section-10}

# **ANSI N45.2-1971 Standard**

During a meeting at Commonwealth Edison in May 1969, the American
National Standards Committee N45 established an Ad Hoc Committee
(N45-3.70 on Quality Assurance Program Requirements. The purpose of this
committee was to prepare a standard for general industry use that would
among other things satisfy the intent and amplify the requirements of
the AEC quality assurance regulations and provide a basis for the
development of detailed quality assurance practices and procedures. The
N45-3.70 Ad Hoc Committee was composed of representatives from the AEC,
its national laboratories and key segments of the nuclear industry,
including utilities, reactor suppliers, plant engineers, and
constructors.

The initial activities of Ad Hoc Committee focused on a critical review
of the draft 10 CFR 50 Appendix B criteria and the preparation of
consensus comments. A number of these comments were considered when the
draft AEC rule was revised and approved for publication.

Following this effort and after extensive discussion, the work group
concluded that, consistent with its purpose, the quality assurance
program standard should be consistent with the format and amplify the
content of the 18 criteria of 10 CFR 50 Appendix B. To expedite the
development process, an editorial team within the committee was formed
and members of the committee were requested to submit their individual
suggestions to the editorial team on the content of the standard. The
team then met in Santa Barbara, California, for an intensive four-day
series of sessions to incorporate input from the committee members into
a coherent document. Practically all of the material submitted by the
committee contributors was used and the resulting draft was unanimously
accepted by the committee with minimal change.

In August 1970, a new 45-3.7 subcommittee on Nuclear Quality Assurance
Standards was formed to provide for consolidation of this and other N45
quality assurance standards. In July 1971, this subcommittee delivered
to the American National Standards Committee N45 its initial draft of
the quality assurance program requirements standard to be issued for
public comment. This subcommittee performed final review of the standard
produced by N45-3.7. Prior to publication, the ANSI standard number was
first changed to N45.3.0 and then to N45.2. ANSI N45.2 also included
definitions and supplemental requirements for design, document and
records control, and audits. ANSI N45.2-1971 was approved by the
American National Standards Committee N45 and the subsequently by the
ANSI Board of Standards Review in October 1971. ANSI N45.2-1971[^43] was
published in February 1972.

AEC and ASME regulations and codes, as well as other American National
Standards, were considered in the development of ANSI N45. The structure
and content of the standard were as follows:

+-----------------------------------------------------------------------+
| > **ANSI N45.2-1971 STRUCTURE**                                       |
| >                                                                     |
| > INTRODUCTION                                                        |
|                                                                       |
| 11. Purpose                                                           |
|                                                                       |
| 12. Scope                                                             |
|                                                                       |
| 13. Responsibility                                                    |
|                                                                       |
| 14. Definitions                                                       |
|                                                                       |
| 15. Referenced Documents                                              |
|                                                                       |
| ```{=html}                                                            |
| <!-- -->                                                              |
| ```                                                                   |
| 1.  QUALITY ASSURANCE PROGRAM                                         |
|                                                                       |
| 2.  ORGANIZATION                                                      |
|                                                                       |
| 3.  DESIGN CONTROL                                                    |
|                                                                       |
|     11. General                                                       |
|                                                                       |
|     12. Interface Control                                             |
|                                                                       |
|     13. Design Verification                                           |
|                                                                       |
|     14. Change Control                                                |
|                                                                       |
| 4.  PROCIREMENT DOCUMENT CONTROL                                      |
|                                                                       |
| 5.  INSTRUCTIONS, PROCEDURES AND DRAWINGS                             |
|                                                                       |
| 6.  DOCUMENT CONTROL                                                  |
|                                                                       |
| 7.  CONTROL OF PURCHASED MATERIAL, EQUIPMENT, AND SERVICES            |
|                                                                       |
| 8.  IDENTIFICATION AND CONTROL OF MATERIALS, PARTS AND COMPONENTS     |
|                                                                       |
| 9.  CONTROL OF SPECIAL PROCESSES                                      |
|                                                                       |
| 10. INSPECTION                                                        |
|                                                                       |
| 11. TEST CONTROL                                                      |
|                                                                       |
| 12. CONTROL OF MEASURING AND TEST EQUIPMENT                           |
|                                                                       |
| 13. HANDLING STORAGE AND SHIPPING                                     |
|                                                                       |
| 14. INSPECTION, TEST, AND OPERATING STATUS                            |
|                                                                       |
| 15. NONCONFORMING ITEMS                                               |
|                                                                       |
| 16. CORREECTIVE ACTION                                                |
|                                                                       |
| 17. QUALITY ASSURANCE RECORDS                                         |
|                                                                       |
| 18. AUDITS                                                            |
+-----------------------------------------------------------------------+

The ANSI N45.2-1971 set fort the requirements and guidance for planning,
managing and implementing overall quality assurance programs for nuclear
power plants. These general requirements were intended to be applied to
all phases of the quality assurance program and to the total power plant
while other codes and standards applied to specific structures, systems
and components of the plant, or to specific activities related to the
plant design, construction or operation.

The principal difference between the ANSI N45.2-1971 standard
requirements and 10 CFR50 Appendix B criteria was in their degree of
specificity, which fundamentally supports the reason why N45.2 was
developed: the ANSI N45-2-1971 standard contained supplemental
requirements and guidance on quality assurance in its subparagraphs.
Appendix B contained only basic criteria, which some industry reviewers
found to be appropriate for use as a regulatory requirement and
licensing commitment. Appendix B applied directly, using a graded
approach, to the applicant (plant owner), whereas ANSI N45-2-1971
applied to any individual organization participating in the nuclear
power plant quality assurance program, such as the nuclear reactor
system designer and supplier, the plant designer; the plant constructor;
and equipment suppliers, as well as the plant owner. Hence the N45.2
standard was to be included or referenced in procurement documents for
items and services essential not only to the safe and reliable operation
of the plant but also to mission success. Some subcontractors and
suppliers have used Appendix B for their quality assurance programs in
non-regulatory situations.

Concurrent with the development and publication of the ANSI N45-2-1971
standard, other Ad Hoc Committees of N45 were developing a series of
standards that set forth more detailed requirements for certain
activities to assure quality of nuclear power plants. These requirements
were to be coordinated with the requirements of ANSI N45.2-1971 as they
were being developed. In September 1971, these ad hoc committees were
changed to working groups. In November 1971, these working groups had
the following work practice standards in preparation:

[Working Groups]{.underline} [Standards in Preparation]{.underline}

N45-3.1 N45.2.1, Cleaning of Fluid Systems and Associated Components for
Nuclear Power Plants

N45-3.2 N45.2.2, Packaging, Shipping, Receiving, Storage and Handling of
Items for Nuclear Power Plants...

N45-3.3 N45.2.3, Housekeeping During the Construction Phase of Nuclear
Power Plants

N45-3.4 N45.2.4, Installation, Inspection and Testing Requirements for
Instrumentation and Electrical Equipment During the Construction of
Nuclear Generating Stations

N45-3.5 N45.2.5, Construction Phase Quality Assurance Requirements for
Installation, Inspection and Testing of Structural Concrete and
Structural Steel for Nuclear Power Plants

N45-3.6 N45.2.6, Qualification of Inspection, Examination and Testing
Personnel for the Construction Phase of Nuclear Power Plants

N45-3.8 N45.2.8, Quality Assurance Requirements During Installation,
Inspection and Testing of Mechanical Equipment and Piping for the
Construction Phase of Nuclear Power Plants

N45-3.9 N45.2.9, Requirements for Quality Assurance Records for Nuclear
Power Plants

N45-3.10 N45-2-10, Quality Assurance Terms and Definitions

**[ASQC Matrix]{.underline}**

In October 1973, the Interface Committee of the American Society for
Quality Control (ASQC) Nuclear Power Technical Committee met in Groton,
Connecticut, to draft a nuclear quality assurance requirements matrix
for release in early November. This matrix presented a side-by-side
comparison of five quality system standards: 10 CFR 50 Appendix B; ANSI
N45.2-1971; ASME SECTION III, NA-4000; RDT F2-2T; and MIL-Q-9858A.
Members of the ASQC Interface Committee attending the meeting included:

Tom Colandrea

Fred Hannon

Gene Langston

The express purpose of this Committee, which met for the first time in
May 1971, was to respond to the needs of the nuclear power industry by
providing education and training in quality standards and regulations
and interfacing with technical societies.

#  {#section-11}

# **AEC Safety Guide 28**

In 1970, the AEC developed and published a series of Safety Guides to
inform applicants of acceptable solutions to specific safety issues.
Consistent with AEC policy to use national consensus standards in the
regulatory process, in June 1972 the AEC staff endorsed the ASME
N45.2-1971 standard in Safety Guide 28 in June 1972 with only a few
regulatory positions.

Later in 1972, the AEC established a new series of Regulatory Guides to
cover a broader scope of regulatory interests. Regulatory Guides were
issued to describe acceptable ways of implementing federal regulations.
According to the AEC, Regulatory Guides were not intended as substitutes
for regulations and compliance with them was not mandatory; other
methods or solutions were acceptable provided that they permitted
positive findings relative to the licensing process.

**[N45.2 Daughter Standards]{.underline}**[^44] [^45] [^46]

After sponsoring the development of the basic N45.2 quality assurance
program requirements standard and the initial drafts of work practice
standards, N45.2.1 -- 2.10, by the working groups, the ANSI Committee
N45 established a permanent N45-3 Subcommittee. The subcommittee was
given broader responsibilities and representation to serve as a focal
point for the completion, coordination, and review of other quality
assurance standards under the cognizance of the ANSI Committee N45 and
for the development of other quality assurance-related standards.

The ANSI N45-3 Subcommittee embodied the members of the previous
subcommittee and ad hoc work groups under N45 sponsorship, and enlisted
the support of other technical societies, including the Institute of
Electrical and Electronic Engineers (IEEE), the American Institute of
Chemical Engineers (AIChE)), the American Society of Civil Engineers
(ASCE), and the American Society for Quality Control (ASQC). For
example, the IEEE Joint Committee of Nuclear Power Standards (JCNPS)
prepared the standard IEEE 336 covering the installation, inspection and
testing of Class 1E electrical equipment. The IEEE JCNPS reviewed and
approved this standard, which was also recognized as the N45.2.4
standard. Furthermore, the ASCE developed a standard covering
installation, and inspection of civil/structural items that was approved
by the N45 committee and became N45.2.5.

In addition to the quality assurance standards initiated in 1969, the
N45-3 subcommittee recognized that supplemental requirements and
non-mandatory guidance would be necessary to effectively implement the
basic quality assurance program requirements of the N45 standard in
certain key quality-affecting areas. Some of the background for these
identified areas is discussed in the following paragraphs:

-   **[Design:]{.underline}** Although Requirement 3 of N45.2 included
    basic requirements for the assurance of quality in the design phase,
    the AEC and industry recognized that more specificity was needed to
    assure quality for the design of nuclear power plants. The N45-2
    Subcommittee formed a work group composed primarily of senior design
    managers to describe realistic design control practices for nuclear
    power plants. This effort resulted in the N45.2.11 daughter
    standard.

-   **[Procurement:]{.underline}** A similar need was identified to
    effectively implement quality assurance program requirements in the
    procurement phase. The N45-2 Subcommittee formed another work group
    consisting of individual with extensive experience in procurement
    and shop inspection practices. This group developed the N45.2.13
    standard.

-   **[Auditing and Auditor Qualification:]{.underline}** Industries
    desire to share audits was a principal driver for the development of
    additional requirement and guidance on the performance of audits and
    the qualification of auditors. Once again, the N45-2 Subcommittee
    selected highly experienced individuals from industry, the national
    laboratories and the government to develop two documents: N45.2.12
    for auditing quality assurance programs for nuclear power plants and
    N45.2.23 for qualification of auditors.

-   **[Records:]{.underline}** Substantive concern was expressed about
    the types of quality assurance records to be generated and retained
    for nuclear power plants and their retention periods. Additionally,
    there was considerable uncertainty about the methods for
    safeguarding records, particularly during the construction phase as
    well as during plant operation when rapid access to plant
    configuration records became vital in an emergency situation. Senior
    individuals familiar with records management and technical needs for
    records during all phases of the plant design, construction,
    operation, maintenance, modification and repair were assembled by
    the Subcommittee for a work group to develop the N45.2.9 standard.

Throughout development of these diverse quality assurance standards, the
N45-2 Subcommittee demonstrated the unique ability to manage, complete
and achieve consensus on a suite of needed standards on an expedited
schedule. This approach received recognition of the value of having a
single organization to manage, oversee and integrate the development of
related standards.

Based on this recognition, in 1973, ANSI, through the Nuclear Technical
Advisory Board (NTAB), made changes in the organizational setup for
producing nuclear standards. These changes included the use of Area
Managers for nuclear standards development under general requirements
suggested by ANSI. The ASME N45-2 Subcommittee took the position that it
would the Area Manager for nuclear quality assurance standards and so a
smaller advisory group within the subcommittee was formed.

The supplemental quality assurance standards became known as "daughter"
standards to the overall N45.2 programmatic standard. ANSI N45.2
daughter standards consisted of two basic types:

1.  Those standards that amplified the programmatic aspects of the
    parent N45.2 quality assurance program standard, and

2.  Those standards that focused on quality-related work practices.

The seven ANSI N45.2 programmatic daughter standards and the
corresponding Regulatory Guides that endorsed them were:

> N45.2.6-1978 Qualifications of Inspection, Examination, and Testing
> Personnel for Nuclear Power Plants (Reg. Guide 1.58)
>
> N45.2.9-1979 Requirements for Collection, Storage, and Maintenance of
> Quality Assurance Records for Nuclear Power Plants (Reg. Guide 1.88)
>
> N45.2.10-1973 Quality Assurance Terms and Definitions (Reg. Guide
> 1.74)
>
> N45.2.11-1974 Quality Assurance Requirements for the Design of Nuclear
> Power Plants Reg. Guide 1.64)
>
> N45.2.12-1977 Requirements for Auditing of Quality Assurance Programs
> for Nuclear Power Plants (Reg. Guide 1.144)
>
> N45.2.13-1976 Quality Assurance Requirements for Control of
> Procurement of Items and Services for Nuclear Power Plants (Reg. Guide
> 1.23)
>
> N45.2.23-1978 Qualification of Quality Assurance Program Audit
> Personnel for Nuclear Power Plants (Reg. Guide 1.146)

Each of the standards issued by the N45-2

Subcommittee was subjected to an intensive preparation and review
process to ensure that they contained precise statement of acceptable
current practices for commercial nuclear power plants -- practices that
were currently available and considered necessary to achieve the
required level of quality, consistent with 10 CFR 50 Appendix B.

**[AEC Rainbow Series]{.underline}**

In the early 1970\'s, the nuclear industry\'s implementation of the
quality assurance criteria of 10 CFR Part 50 Appendix B was variable and
inconsistent. Many of the ANSI N45.2 quality assurance daughter
standards were still under development. Guidance was lacking on
acceptable quality assurance practices for the relatively few nuclear
power plants that were in operation and for the large number of plants
nearing completion. More dialogue between nuclear industry and
regulatory management was needed to better understand what was required
to implement an effective N45.2 quality assurance program in compliance
with 10 CFR Part 50 Appendix B.

In June 1973, in conjunction with a new AEC regulatory review process,
AEC Commissioner L. Manning Muntzing directed the AEC DRS regulatory
staff to issued a series of guidance documents, the first referred to as
the "Gray Book" (WASH 1283) to provide guidance on quality assurance
requirements during the design and procurement phase of nuclear power
plant construction. With the concurrence of the N45 Standards Committee,
the "Gray Book" included the pertinent N45.2 quality assurance standards
that had been issued or were nearing completion. In July 1973, the AEC
Director of Regulation and staff sponsored a series of regional
conferences across the U.S. to discuss the contents of the "Gray Book".

In October 1973, the AEC regulatory staff issued the second in the
series, the "Orange Book" (WASH 1284), to provide additional guidance on
quality assurance during the operating phase of nuclear power plants. A
second series of regional conferences was held in November 1973 to
discuss the "Orange Book".

In May 1974, the "Gray Book" was redesigned as a compilation of federal
regulations, Regulatory Guides, American National Standards, and
conference comments pertinent to nuclear power plant quality assurance
during design and construction.

Also in May 1974, the third in the Rainbow Series of regulatory
document, known as the "Green Book" (WASH 1309), was issued and a final
series of regional conferences were held to discuss quality assurance
during the construction phase of nuclear power plants.

It is significant to note that an underlying theme for the Rainbow
Series of documents was the encouragement by the AEC to use applicable
national consensus standards, particularly N45.2 and its daughter
standards some of which were recognized as being appropriate even in
draft form, to meet the criteria of 10 CFR Part 50 Appendix B. The ANSI
N45-3 Subcommittee was pleased with the sustained participation of Bill
Morrison of the NRC staff and timely support of the AEC management from
the Commissioners, and the directors of reactor standards and licensing,
DRS and DRL, and their staffs.

**[ASME N45-2-1977]{.underline}**

The N45 Committee was an American National Standards Committee operation
under the ANSI Procedures for Standards Development Committees. Its
Secretariat was ASME. In 1975, this changed when ASME became an
accredited standards development committee and took over the N45.2
standards. ANSI then terminated the N45 Committee and the ANSI N45.2
standards became ASME N45.2 standards.

Responding to user experience and feedback and the perceived need to
expand quality assurance program requirements to encompass other
regulated nuclear facilities, a revision of the ANSI N45.2 standard was
published in 1977. Accordingly, the title and scope of the standard was
changed to \"ASME N45-2-1977, Quality Assurance Program Requirements for
Nuclear Facilities\". This change provided for the application of the
ASME N45.2 standard to nuclear facilities for power generation, spent
nuclear fuel storage, fuel reprocessing, and plutonium processing and
fuel fabrication. This change was consistent with the extended
applicability of 10 CFR Part 50 Appendix B to Parts 70, which applies to
fuel reprocessing facilities.

Revisions 1 (1978) and 2 (1979) to Regulatory Guide 1.28 (formerly
Safety Guide 28) endorsed the N45.2-1977 edition with only a few
supplemental regulatory positions. Revision 3 of NRC Regulatory Guide
1.28 permitted applicants to follow either the appropriate ANSI/ASME
N45.2-series of standards or endorsed ANSI/ASME NQA-1-1983 and ANSI/ASME
NQA-1a-1983 addenda with Regulatory Positions on:

-   Qualifications of Inspection and Test Personnel

-   Quality assurance records retention times

-   Internal and external quality assurance program audits

# **SECTION IV.** 

#  {#section-12}

# **ASME AND RELATED NQA STANDARDS**

#  {#section-13}

*This section describes the evolution of the ASME NQA-1and other "N"
committees quality assurance program requirements standards; the
consolidation of the ASME N45.2 work practice standards and ASME NQA-1
standards, and their restructuring of the ASME NQA standards from 1979
to 2004.*

# **"N" Committees Standards**

Under what had begun as a generally harmonious cover of ANSI "N"
Committees for nuclear standards development, jurisdictional and
redundancy concerns arose in the 1970's among the various sponsoring
technical societies, including ASME (N45), IEEE (N41), ANS (N18), and
AIChE (N46).

**[ANS N18.7( 3.2) Standards]{.underline}**[^47]

Preparation of the first edition of the ANSI/ANS 3.2 standard commenced
in 1969. Historically, the administrative controls section of the
Facility Operating License Technical Specifications contained provisions
for meeting many of the requirements that subsequently became associated
with the quality assurance requirements for nuclear power plant
operation. During the same period, the ANSI N45-3 Subcommittee was
developing N45.2 quality assurance standards.

In 1972, the AEC issued Safety Guide 33, endorsing Draft 8 of ANS 3.2
(which later became ANSI N18.7-1972) and ANSI/ANSI N45.2-1971. Because
of this dual NRC endorsement, the ANS 3.2 and N45-2 standards-writing
groups undertook a cooperative effort to incorporate the appropriate
quality assurance requirements for operation into a single standard. The
result was ANSI N18.7-1976 (ANS 3.2), which was endorsed by NRC
Regulatory Guide 1.33, Revision 2, in February 1978.

Following the Three Mile Island Unit 2 accident in 1979, ANS revised
N18.7-1976 to incorporate administrative "lessons learned" into the
standard, which was subsequently published as ANSI/ANS 3.2-1982. This
revision also reflected the requirements of ASME NQA-1-1979 discussed
below, which had superseded several of the N45.2 daughter standard that
had been referenced in N18.7-1976. NRC Regulatory Guide 1.33, Revision
2, dated February 1978, endorsed this revision.

The 1988 and 1994 versions of ANS 3.2 continued strong emphasis on
operational aspects and performance-based quality assurance practices.

**[ANSI N46.2 Standards]{.underline}**

Paralleling the development of the ANSI N45.2 standards, the N46
Committee sponsored by AIChE drafted a quality assurance program
standard for fuel cycle facilities that was similar to ANSI N45.2 in its
format and requirements, In 1978, the ANSI N46-2 Committee issued
Revision 1 of its ANSI N46.2-1978 standard for Post-Reactor Nuclear Fuel
Cycle Facilities as an American National Standard. By joint agreement
between ASME and AIChE to use N45.2, N46.2 was withdrawn.

**[ASME NQA Committee]{.underline}**

**\\**

Recognizing the need to minimize redundancy in similar requirements and
for clearer definition of responsibilities for quality assurance program
standards development and maintenance for nuclear facility applications,
early in 1975, the ANSI Nuclear Standards Management Board (NSMB) issued
a policy bulletin stating that:

**"There should be a single quality assurance standard for nuclear
activities\"**

Consequently, the NSMB under ANSI assigned overall responsibility for
development, coordination among other technical societies, and
maintenance of quality assurance program standards for nuclear facility
applications, which included the N45.2 standards, to the N45
Subcommittee. Because of this NSMB policy pronouncement, the chairs of
the N45 and N46 efforts agreed to merge their committees and develop a
single standard covering both scopes. Subsequently, ANSI modified its
policy to allow standards-writing organizations to develop and approve
standards using their own procedures, provided these procedures met ANSI
consensus requirements. The organizations' standards then could be
submitted to ANSI for their approval. Such standards could be designated
as products of the organization and carry the statement that they were
American National Standards. In response to this policy, the ASME Board
of Nuclear Codes and Standards was formed to manage ASME nuclear
standards efforts. In October 1975, the N45 Committee transferred
quality assurance standards responsibility to a newly constituted ASME
Committee on Nuclear Quality Assurance (NQA).

In a special report[^48] on "QA Standards -- Application and Problems",
John Landis, President of Gulf General Atomic Company and chairman of
ANSI referred to the new ASME NQA Committee that would carry on and
expand the work of the former ANSI N45-3 Subcommittee.

#  {#section-14}

# **ASME NQA-1-1979** 

Because different N45-3 working groups had developed numerous
interrelated N45.2 daughter standards at different times, these various
standards contained some redundant and conflicting quality assurance
program requirements, causing users and regulators confusion in their
application, endorsement and enforcement. The N45-3 Subcommittee had
decided early on that it was more important to issue these urgently
needed standards and obtain feedback from industry on their use rather
than delay their development and issuance process to attempt
harmonization of redundant requirements. It was always the
Subcommittee's intent to consolidate them at some point into a single
document.

In 1979, ANSI N46.2 Committee merged with the ASME NQA Committee to
jointly produced ASME NQA-1-1979, which integrated ASME N45-2-1977 and
ANSI N46.2-1978.[^49]

The need to consolidate the N45.2 series of standards was also firmly
supported by the ASME as a means for amplifying the quality assurance
provisions of the ASME Boiler and Pressure Vessel Code to make them more
compatible with regulatory requirements.

The task of the new ASME NQA Committee was to consolidate the quality
assurance program requirements of the ASME N45.2-1977 standard and the
seven N45.2.6 - 2.23 programmatic daughter standards listed above into a
single standard. Toward this objective, the new NQA Committee adopted
the following approach to consolidation:

-   The 18 criteria structure of Appendix B would be preserved as basic
    requirements.

-   These basic requirements would provide an overview of the quality
    assurance program logic and would be sufficiently general to have
    wide applicability.

-   More specific, detailed requirements would be contained in
    supplements.

-   Requirements would be clearly separated from guidance, the latter
    being contained in non-mandatory appendices.

-   The full expertise of ASME and other standards-writing societies
    would be employed in developing, coordinating, and maintaining the
    standard.

-   The standard would provide for flexibility in its application as
    well as growth or reduction of supplementary requirements and
    guidance.

-   Redundancy and conflicts in programmatic requirements would be
    minimized.

The standard would not be as limited as 10 CFR 50Appendix B is to
safety-related structures, systems components, and associated activities
but would be applicable also to those items and activities that were
essential to the achievement of project objectives and assurance of
reliable operation.

As illustrated in Table 1I in this document below, the newly
consolidated and restructured ASME NQA-1-1979 standard was restructured
as basic requirements, supplements and appendices, as follows:

-   Retained the 18 Basic Requirements structure of 10 CFR Part 50
    Appendix B.

-   Incorporated detailed requirements and guidance from the seven
    programmatic daughter standards separately in Supplements and
    non-mandatory Appendices.

-   Reworded Basic Requirements using clear and concise language, such
    as \"The design shall be defined, controlled, and verified\...\"
    instead of the obtuse language of Appendix B, i.e., \"Measures shall
    be established to..."

    Three members of the ASME NQA Committee, Gordon Beer, Bud Crevasse
    and the Gene Langston met in Consolidated Edison's New York City
    office and literally cut and pasted the reworded draft of the ASME-
    NQA-1-1979 standard. The revised draft standard was further reviewed
    and edited by the ASME NQA Committee and approved through the
    consensus balloting process.

    **[ASME NQA-1-1983]{.underline}**

    The ASME NQA Committee made minor changes in wording of the
    requirements and guidance in the 1983 edition and two addenda to
    ASME NQA-1.[^50]

    In August 1984, the NRC endorsed ASME NQA-1-1983 in Revision 3 to
    Regulatory Guide 1.28 in August 1985 with only three regulatory
    positions. These regulatory positions were concerned with the
    qualification of inspection, test, and nondestructive examination
    personnel; quality assurance records; and audit frequency. While the
    NRC has not consistently endorsed successive revisions after the
    1983 edition, several licensees have had their quality assurance
    program approved to a version a more recent version of the standard.

    **[ASME NQA-1-1986]{.underline}**

    The ASME NQA Committee issued three addenda to the ASME --NQA-1-1986
    edition and addenda with relatively minor editorial changes.[^51]

# **ASME NQA-2 STANDARDS** {#asme-nqa-2-standards .unnumbered}

The NQA Committees developed the ASME NQA-2-1983 standard [^52] to
incorporate the following seven ASME N45.2 daughter standards as parts.
ASME NQA.2-1983 standard was amended eight times from 1983 to 1991:

N45.2.1-1980 Cleaning of Fluid Systems and Associated Components for
Nuclear Power Plants (NQA-2 Part 2.1)

N45.2.2-1978 Packaging, Shipping, Receiving, Storage, and Handling of
Items for Nuclear Power Plants (NQA-2 Part 2.2)

N45.2.3-1973 Housekeeping During the Construction Phase of Nuclear Power
Plants (NQA-2 Part 2.3)

N45.2.5-1978 Supplementary Quality Assurance Requirements for
Installation, Inspection, and Testing of Structural Concrete, Structural
Steel, Soils, and Foundations During the Construction Phase of Nuclear
Power Plants (NQA-2 Part 2.5)

N45.2.8-1975 Supplementary Quality Assurance Requirements for
Installation, Inspection and Testing of Mechanical Equipment and Systems
for the Construction Phase of Nuclear Power Plants NQA-2 Part 2.8)

N45.2.15-1981 Hoisting, Rigging, and Transporting of Items for Nuclear
Power Plants (NQA-2 Part 2.15)

N45.2.20-1979 Supplementary Quality Assurance Requirements for
Subsurface Investigations for Nuclear Power Plants (NQA-2 Part 2.20)

After issuing the ASME NQA-2-1983 edition, the following additional
NQA-2 quality assurance work practice standards were incorporated as
additional sub-parts of ASME NQA-2-1986 and NQA-2-1989 editions and
addenda:

Part 2.4 Installation, Inspection, and Testing Requirements for Power,
Instrumentation, and Control Equipment at Nuclear Facilities (IEEE 336)

Part 2.7 Quality Assurance Requirements of Computer Software for Nuclear
Facility Applications -- Incorporated by reference to IEEE 336

Part 2.16 Requirements for Calibration and Control of Measuring and Test
Equipment Used in Nuclear Facilities -- Incorporated by reference to
IEEE 498, subsequently canceled.

Part 2.18 Quality Assurance Requirements for Maintenance of Nuclear
Facilities

**Referencing parts of NQA-2 standards to another society's standards
where the expertise existed seemed like a good idea at the time.
However, when these standards were not maintained or were canceled by
the other society, (e.g., IEEE 498), ASME N45.2 was left without a valid
reference.**

# **ASME NQA-1-1989**  {#asme-nqa-1-1989 .unnumbered}

In the 1989 edition of ASME NQA-1[^53], the ASME NQA Committee
incorporated the changes of the 1983 and 1986 editions and their
addenda. The scope of ASME NQA-1-1989 was extended to include siting and
decommissioning of nuclear facilities. Certain supplementary
requirements, such as design control and document control, were
clarified or amplified. Three addenda to the 1989 edition were issued in
1990, 1991 and 1992.

# **ASME NQA-3-1989**  {#asme-nqa-3-1989 .unnumbered}

Responding to an identified need, in 1984, the ASME NQA Main Committee
established a Subcommittee on Nuclear Waste Management. This
subcommittee was assigned the task of developing a standard for assuring
quality during site characterization of high-level nuclear waste
repositories. With the assistance of geotechnical experts from the U.S.
Geological Survey and DOE national laboratories and the tacit support of
the DOE and NRC, this Subcommittee prepared ANSI/ASME NQA-3-1989[^54].

In addition to those activities affecting quality in the ASME NQA-1
standard, ASME NQA-3 contained basic requirements, supplements and
non-mandatory appendices on:

-   Readiness reviews

-   Peer reviews

-   Data and sample management

-   Data collection and analysis

-   Coring

-   Sampling

-   In-situ testing

-   Scientific investigation

-   Design data process control

    In restructuring and consolidating the three ASME NQA standards,
    NQA-1, NQA-2 and NQA-3, the ASME NQA Committee was undecided about
    what to do with NQA-3. The reason was that NQA-3 was an application
    standard for which Part IV of NQA-1 had not yet been established.
    Thus, according to its Foreword, ASME NQA-3-1989 guidance on the
    application of NQA-3-type quality assurance programs was expected to
    be included in future revisions of Part III or Part IV of ASME
    NQA-1. In the current ASME NQA-1-2004 edition, ASME NQA-3 is
    embedded in Part III, Subpart 3.3, as a non-mandatory appendix. It
    is further noted that salient requirements (of NQA-3) will
    subsequently be integrated into Parts I and/or Part II in a future
    edition of the ASME NQA-1 standard.

    ASME NQA-3-1989 was listed as a source document in Revision 10 of
    the DOE Office of Civilian Radioactive Waste Management Quality
    Assurance Requirements and Description Document for site
    characterization work conducted for the DOE Yucca Mountain Project.

#  {#section-15 .unnumbered}

# **ASME NQA-1-1994**  {#asme-nqa-1-1994 .unnumbered}

In the early 1990\'s, the ASME NQA Committee leadership perceived that
the NQA-1, NQA-2 and NQA-3 standards were not structured in a way that
enabled users to understand and apply these three standards. Thus, the
NQA Committee decided to consolidate the NQA-1 and NQA-2 standards into
a single multi-part document that was intended to allow a more rapid
response to varied applications of the NQA standards quality assurance
requirements and guidance. The NQA Committee restructured ASME
NQA-1-1994[^55]. The title was changed to nuclear facility applications"
to accommodate the inclusion of NQA-2. The 1994 edition consisted of the
following three parts:

-   PART I contained an introduction and basic quality assurance program
    requirements followed by supplementary requirements for nuclear
    facilities from the former NQA-1 standard.

-   PART II included the quality assurance requirements for nuclear
    facility applications work practices requirements for facility
    applications as subparts from the former NQA-2 standard. A new
    subpart 2.21 was added in the 1995 addenda to address quality
    assurance guidelines for decommissioning nuclear facilities.

-   PART III consisted of non-mandatory guidance as subparts in
    appendices from the former NQA-1 and NQA-2 standards.

    ANSI/ASME NQA-3 was not integrated into the restructured NQA-1
    standard at this time.

    An addendum to the 1994 edition, ASME NQA-1a-1995, was issued on
    December 7, 1995. It contained a definition of surveillance and a
    number of editorial changes.

    **ASME NQA-1-1997**

    Quality assurance program requirements and guidance included in the
    NQA-1 standard were derived from and focused primarily on
    compliance-based practices and after-the-fact verification. Both
    industry and NRC leadership expressed an interest in revisiting some
    of the fundamental quality assurance program requirements, including
    those of Appendix B to 10 CFR Part 50, on performance-based
    practices. Consequently, a major review and revision effort was
    initiated by the NQA Committee to improve and update the
    requirements and guidance in a performance-based NQA-1 standard.
    Criteria were developed and the standard was revised to complement
    proven quality assurance program principles and practices that had
    consistently provided a high level of quality during the past three
    decades. Performance-based criteria were retained or introduced to
    ensure safe and reliable nuclear facility operation. Factors related
    to the current technology and maturity of the nuclear industry were
    used in evaluating whether the benefits were commensurate with
    implementation costs and whether the level of detail was appropriate
    for the desired results.

    The specific criteria used for the review and revision of the
    standard were:[^56]

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-   Is it performance based?

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-   Does it contribute to safe and reliable nuclear facility operation?

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-   Are benefits commensurate with implementation costs?

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-   Is it consistent with current technology and maturity of the nuclear
    industry?

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-   Is the level of detail adequate to achieve the desired results?

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-   Is the level of detail more than the minimum required to produce the
    desired results?

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```
-   For the intended activity, is this the minimum requirement that
    applies to all applications?

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-   Is the requirement stated once and not duplicated?

    With the expansion of the NQA-1 scope from nuclear power plants to
    nuclear facilities, minimum requirements that were commonly
    applicable to nuclear facilities were presented. The NQA Committee
    also did some housecleaning to further remove redundant phraseology.
    The most common example of redundancy was to remove \"written\" or
    \"documented\" for quality assurance program-related activities that
    are generally required or implied to be documented by Requirement 5,
    "Instructions, Procedures and Drawings" and records maintained in
    accordance with Requirement 17, "Quality Assurance Records".
    However, the NQA Committee was firm in its intention not to delete
    any basic quality principles or practices that have contributed to
    the achievement and maintenance of safe and reliable operation in
    U.S. nuclear facilities. In those cases where there were
    prescriptive details for compliance with a requirement, this type of
    information was deliberately moved to guidance as an appendix in
    Part III.

    Seeking to further improve the usability of the NQA -1 standard, the
    NQA Committee once again restructured the consolidated ASME NQA-1
    standard in the 1997 edition[^57], as follows:

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```
-   PART I included the Introduction and combined the basic and
    supplementary requirements for quality assurance programs from the
    former NQA-1 standard.

-   PART II contained the quality assurance requirements for nuclear
    facility applications as work practices from the former NQA-2
    standard.

-   PART III consisted of the non-mandatory guidance as appendices and
    subparts from NQA-1 and NQA-2 and NQA-3.

-   PART IV was reserved for non-mandatory guidance as appendices on
    positions and application matrices, such as Subpart 4.1, Application
    Appendix - Guidance on Quality Assurance Requirements for Computer
    Software.

    As noted above, changes to the NQA-1-1997 edition and Part I of the
    1999 addenda[^58] deleted or modified requirements of earlier
    editions and addenda to further minimize redundancy and eliminate
    unduly restrictive requirements. Some requirements were changed to
    useful guidance and included in Part III.

    An addendum to the 1997 edition, ASME NQA-1a-1999, was issued on May
    25, 1999. It contained a definition of surveillance and a number of
    editorial changes. A significant change to NQA-1 was made by the
    complete revision of Subpart 2.7 in Part II, Quality Assurance
    Requirements for Computer Software for Nuclear Facility
    Applications, to make it compatible with the software quality
    assurance guidance in Part IV. Requirements 3 and 11 of NQA-1 were
    also modified to incorporate basic requirements for software
    development and testing.

> **[ASME NQA-1-2000]{.underline}**

This edition[^59] continued the NQA Committee practice of providing
additional guidance with only minimum change to requirements.
Definitions of document and quality assurance record were revised.
Requirements 3, Design Control, were minimally revised in the Addenda.

Several non-mandatory appendices were added to PART IV: Subpart 4.1,
Guide on Quality Assurance Requirements for Software; and Subpart 4.2,
Guidance on Graded Application of Quality Assurance (QA) for
Nuclear-Related Research and Development.

The NRC issued a limited endorsement of ASME NQA-1-2000 in Exelon's
licensing application for its nuclear facilities.

**[ASME NQA-1-2004]{.underline}**

ASME NQA-1-2004[^60] revised the ASME NQA-1-2000 edition by retaining
the four-part format that began with the restructuring of the ASME
NQA-1-1994 edition. This standard includes the revisions, corrections
and editorial changes introduced in the NQA-1a-2002 addenda.

**[ASME NQA Survey]{.underline}**

An ASME NQA survey was conducted in 2004 of the quality assurance
program basis for U.S. nuclear power reactors. The survey indicated an
encouraging shift of numerous nuclear utilities toward the adoption of
the ASME NQA-1 1994 and later standards.

**[ASME NQA Standards Benefits]{.underline}**[^61]

The potential benefits of using the ASME NQA standards are:

-   Standards are developed by a broad and balanced representation of
    experts and stockholders with many years of experience in a wide
    variety of licensed and unlicensed nuclear facilities.

-   Standards development is an open process that encourages free
    expression of views.

-   Technical issues are discussed and resolved in a non-confrontational
    and non-threatening environment.

-   Open process provides for public review.

-   Standards clearly separate basic and supplementary requirements from
    non-mandatory guidance.

-   Standards are performance based.

-   Standards requirements can be applied selectively using a graded
    approach.

-   Standards approval is achieved through a consensus process with no
    single dominating interest.

-   Standards are reviewed and updated in addenda and revised
    periodically to keep abreast of current industry and regulatory
    practices.

-   Process provides for an opportunity for interpretation and revision
    of requirements and guidance.

-   Standards can achieve greater public acceptance and credibility that
    similar industry practices or regulatory requirements.

    **SECTION V.**

    *This section describes the early history of the ASME Boiler and
    Pressure Vessel (B&PV) Code Section III.*

    **ASME B&PV CODE SECTION III** [^62]

    Although not yet required by federal regulations, ASME, as a joint
    industry-regulatory effort, was capturing those elements critical to
    pressure-retaining components used in nuclear power plants. In 1963,
    ASME introduced Section III of the B&PV Code. It contained no
    quality assurance requirement until the 1967 Amendment. Then Section
    III contained 15 quality assurance criteria based largely on MIL-Q
    9858A, as follows:

+-----------------------------------------------------------------------+
| **ASME B&PV Code Section III Appendix IX**                            |
|                                                                       |
| **Quality Assurance Criteria**                                        |
+-----------------------------------------------------------------------+
| -   Organization                                                      |
|                                                                       |
| -   Quality Assurance Control Program                                 |
|                                                                       |
| -   Description of Procedures                                         |
|                                                                       |
| -   Drawings and Changes                                              |
|                                                                       |
| -   Receiving Examination                                             |
|                                                                       |
| -   Control, Identification and Marking                               |
|                                                                       |
| -   Manufacturing Fabrication Procedures                              |
|                                                                       |
| -   Examination                                                       |
|                                                                       |
| -   Testing                                                           |
|                                                                       |
| -   Calibration                                                       |
|                                                                       |
| -   Handling, Storage and Delivery                                    |
|                                                                       |
| -   Component and Material Repair                                     |
|                                                                       |
| -   Quality Control Records                                           |
|                                                                       |
| -   Quality Audits                                                    |
+-----------------------------------------------------------------------+

ASME Section III Appendix IX required ASME to review and approve quality
assurance programs. It also required a firm to hire a third-party
inspection agency, usually an insurance company.

In 1989, ASME Section III dropped Appendix IX and adopted ASME
NQA-1-1989 with Addenda through 1992 as its quality its quality
assurance basis. ASME Section III added and modified basic and
supplemental requirements of NQA-1 with some modification of definitions
in NCA 4120 and requirements in NCA 4134 for Division 1 Code activities
and in WA 4120 and WA 4134 for Division 3 Code activities.

##### SECTION VI. {#section-vi. .unnumbered}

##### NQA MANAGEMENT ISSUES[^63] {#nqa-management-issues .unnumbered}

> Since 1985, when Revision 3 to Regulatory Guide 1.28 endorsed ASME
> NQA-1-1983, the NRC staff has been reluctant to develop regulatory
> positions based on the evolution of nuclear quality assurance program
> requirement, and to endorse current versions of ASME NQA-1. This
> position was considered by the NRC staff to be reasonable because of
> the resistance of plant owners and operators to change from previously
> approved regulatory commitments due to cost and licensing
> uncertainties. An ASME NQA survey indicates that progress is being
> made by the nuclear utilities in adopting more current editions of
> ASME NQA-1. The ASME NQA Committee continues to work toward the
> broader NRC endorsement of current NQA-1 editions in keeping with a
> longstanding regulatory policy to endorse national consensus
> standards. Other ASME NQA Committee management issues are:

-   To improve the timeliness of ASME NQA Committee responses to
    enquires and changing needs and new technologies.

-   To take more positive action on NQA Committee initiatives to include
    performance-based requirements and guidance to complement
    traditional programmatic compliance elements.

# **SECTION VII.**  {#section-vii. .unnumbered}

#  {#section-16 .unnumbered}

# **ASME NQA VISION**[^64] {#asme-nqa-vision .unnumbered}

Looking to the future of NQA requirements and the ASME NQA standard, one
must never forget the need to protect the health and safety of the
public. The viability of the nuclear industry is directly linked to its
safety record of success. This relationship was validated by the events
that occurred at the Three Mile Island nuclear facility in March 1979.
However, since then, the record for nuclear utility operational
performance has improved dramatically, and facility safety has not been
compromised. The industry must continue to focus on those activities and
programs that are necessary to maintain safety and realize that this
effort must never be sacrificed to achieve performance of financial
objectives.

The nuclear industry must again be capable of building new nuclear power
generation in the United States. There are many complex issues related
to new nuclear power generation, including the application of new
quality assurance regulations and the associated consensus standards,
including ASME NQA-1, and the experience and technical capability of
vendors to supply quality products to complex plants.

New nuclear power generation must embrace the latest technologies and
regulations in an efficient and cost effective manner. New technologies
must utilize proven quality control concepts based on past performance
but tailored to the demands of the new technologies. Regulations for new
nuclear power generation must also move out of the past and into today's
changing environment without compromising the commitment to safety.

In today's environment of limited resources, standardization is one
method of achieving a cost-effective product. This can also apply to the
regulation of new nuclear power generation. There needs to be one set of
quality assurance requirements related to nuclear power generation
because the health and safety of a citizen in New York is just as
important to a citizen in Georgia, California or Illinois. Support of
the ASME NQA-1 standard with regulatory endorsement is the best method
of achieving this standardization. The NQA Committee has been
continuously evaluating the NQA-1 standard to address new technologies
and regulatory position changes and is ready to be a part of the effort
to make new nuclear power generation a viable solution to our nation's
energy programs.

Since its inception, the NQA Committee has continued to focus on
providing a timely and value-added product to the industry. The
Committee vision is for ASME NQA-1 to be recognized as the world-class
nuclear quality assurance standard of choice adopted by organization to
achieve safe performance and reliable cost-effective operation in all
activities while maintaining an acceptable compliance position relative
to their respective regulatory requirements. To achieve this, the vision
of the NQA Committee will be to continue to pursue the following:

-   Develop timely changes to address new and improved technologies,
    changes in regulations and industry experience.

-   Promote the utilization of the latest edition of ASME NQA-1 for all
    nuclear facilities, including both licensed and unlicensed
    facilities.

-   Seek support, interaction and cooperation of the NRC and other
    regulatory bodies in the utilization of the latest edition of the
    ASME NQA-1 standard through a consensus process that minimizes
    differences between the standard and the accepted industry and
    regulatory positions.

# TABLE I: DOD MIL-Q-9858A, NASA NHB 5300.4(1B) AND NRC 10 CFR 50 APPENDIX B COMPARISON {#table-i-dod-mil-q-9858a-nasa-nhb-5300.41b-and-nrc-10-cfr-50-appendix-b-comparison .unnumbered}

+-----------------------+-----------------------+-----------------------+
| # **DOD MI            | # **NASA NHB 530      | # *                   |
| L-Q-9858A** {#dod-mil | 0.4(1B)** {#nasa-nhb- | *AEC/NRC 10 CFR Part  |
| -q-9858a .unnumbered} | 5300.41b .unnumbered} | 50 Appendix B** {#aec |
|                       |                       | nrc-10-cfr-part-50-ap |
| ### QUALITY P         | **Quality Program     | pendix-b .unnumbered} |
| ROGRAM REQUIREMENTS { | Provisions for        |                       |
| #quality-program-requ | Aeronautical and      | ### Quality Assur     |
| irements .unnumbered} | Space System          | ance Criteria for Nuc |
|                       | Contractors**         | lear Power Plants {#q |
|                       |                       | uality-assurance-crit |
|                       |                       | eria-for-nuclear-powe |
|                       |                       | r-plants .unnumbered} |
+-----------------------+-----------------------+-----------------------+
| # 1.1                 | # Chapter 1: INTRODUC | # Introduction {#intr |
| Applicability {#appli | TION {#chapter-1-intr | oduction .unnumbered} |
| cability .unnumbered} | oduction .unnumbered} |                       |
+-----------------------+-----------------------+-----------------------+
| 1.2 Summary           | #                     | # II. Quality         |
|                       |  1B100 General {#b100 |  Assurance Program {# |
|                       | -general .unnumbered} | ii.-quality-assurance |
|                       |                       | -program .unnumbered} |
+-----------------------+-----------------------+-----------------------+
| 3.1 Organization      | 1B201 ORGANIZATION    | # I. O                |
|                       |                       | rganization {#i.-orga |
|                       |                       | nization .unnumbered} |
+-----------------------+-----------------------+-----------------------+
| # 3.2 I               | Chapter 2: QUALITY    | # II. Quality A       |
| nitial Quality Planni | PROGRAM MANAGEMENT    | ssurance Program {#ii |
| ng {#initial-quality- | AND PLANNING          | .-quality-assurance-p |
| planning .unnumbered} |                       | rogram-1 .unnumbered} |
|                       | 1B 200 GENERAL        |                       |
|                       |                       |                       |
|                       | 1B202 TRAINING        |                       |
+-----------------------+-----------------------+-----------------------+
| # 3.3 Work Ins        | # 1B 2                | # V. Instructi        |
| tructions {#work-inst | 03 QUALITY INFORMATIO | ons, Procedures, and  |
| ructions .unnumbered} | N {#b-203-quality-inf | Drawings {#v.-instruc |
|                       | ormation .unnumbered} | tions-procedures-and- |
|                       |                       | drawings .unnumbered} |
|                       | 1B 300 TECHNICAL      |                       |
|                       | DOCUMENT              | VI\. Document Control |
+-----------------------+-----------------------+-----------------------+
| #  {#se               | # CHAPTER 3: DESI     | # III. Design         |
| ction-18 .unnumbered} | GN AND DEVELOPMENT CO | Control {#iii.-design |
|                       | NTROLS {#chapter-3-de | -control .unnumbered} |
|                       | sign-and-development- |                       |
|                       | controls .unnumbered} |                       |
|                       |                       |                       |
|                       | 1B 300 TECHNICAL      |                       |
|                       | DOCUMENTS             |                       |
|                       |                       |                       |
|                       | 1B 302 CHANGE CONTROL |                       |
+-----------------------+-----------------------+-----------------------+
| # 3.4 Records {       | # 1B405 DATA RET      | # XVII. Quality A     |
| #records .unnumbered} | RIEBAL OF RECORDS {#b | ssurance Records {#xv |
|                       | 405-data-retriebal-of | ii.-quality-assurance |
|                       | -records .unnumbered} | -records .unnumbered} |
+-----------------------+-----------------------+-----------------------+
| # 3.5 Correcti        | # 1b802/1B907 RE      | # XVI. Corrective Ac  |
| ve Action {#correctiv | MEDIAL AND PREVENTIVE | tion {#xvi.-correctiv |
| e-action .unnumbered} |  ACTION {#b8021b907-r | e-action .unnumbered} |
|                       | emedial-and-preventiv |                       |
|                       | e-action .unnumbered} |                       |
+-----------------------+-----------------------+-----------------------+
| # 4.1 Drawings {#     | #  {#se               | # V. Instruction      |
| drawings .unnumbered} | ction-19 .unnumbered} | s, Procedures, and Dr |
|                       |                       | awings {#v.-instructi |
|                       |                       | ons-procedures-and-dr |
|                       |                       | awings-1 .unnumbered} |
+-----------------------+-----------------------+-----------------------+
| # 4.2 Measuring       | # CHAPTER 9:          | #                     |
|  and Test Equipment { |  METROLOGY CONTROLS { |  XII. Control of Meas |
| #measuring-and-test-e | #chapter-9-metrology- | uring and Test Equipm |
| quipment .unnumbered} | controls .unnumbered} | ent {#xii.-control-of |
|                       |                       | -measuring-and-test-e |
|                       |                       | quipment .unnumbered} |
+-----------------------+-----------------------+-----------------------+
| # 5.0 Control of Pur  | # CHAPTER 5: P        | # IV. Procurement D   |
| chases {#control-of-p | ROCUREMENT CONTROL {# | ocument Control {#iv. |
| urchases .unnumbered} | chapter-5-procurement | -procurement-document |
|                       | -control .unnumbered} | -control .unnumbered} |
| 5.1 Responsibility    |                       |                       |
|                       | 1B501 SELECTION OF    | VII\. Control of      |
| 5.2 Purchasing Data   | CONTRACTOR            | Purchased Equipment   |
|                       | PROCUREMENT SOURCES   | and Services          |
|                       |                       |                       |
|                       | 1B502 PROCUREMENT     |                       |
|                       | DOCUMENTS             |                       |
+-----------------------+-----------------------+-----------------------+
| # 6.1 Materials and   | # CHAPTER 4: IDENTI   | # VII                 |
|  Material Control {#m | FICATION AND DATA RET | I. Identification and |
| aterials-and-material | RIVAL {#chapter-4-ide |  Control of Materials |
| -control .unnumbered} | ntification-and-data- | , Parts, and Componen |
|                       | retrival .unnumbered} | ts {#viii.-identifica |
|                       |                       | tion-and-control-of-m |
|                       |                       | aterials-parts-and-co |
|                       |                       | mponents .unnumbered} |
+-----------------------+-----------------------+-----------------------+
| # 6.2 Produc          | # CHAPTER 6: FAB      | # IX. Control of Sp   |
| tion Processing and F | RICATION CONTROLS {#c | ecial Processes {#ix. |
| abrication {#producti | hapter-6-fabrication- | -control-of-special-p |
| on-processing-and-fab | controls .unnumbered} | rocesses .unnumbered} |
| rication .unnumbered} |                       |                       |
+-----------------------+-----------------------+-----------------------+
| # 6.3                 | # 1b704 END-ITEM IN   | #                     |
| Completed Item Inspec | SPECTION AND TEST SPE | X. Inspection {#x.-in |
| tion and Test {#compl | CIFICATIONS AND PROCE | spection .unnumbered} |
| eted-item-inspection- | DURES {#b704-end-item |                       |
| and-test .unnumbered} | -inspection-and-test- | XI\. Test Control     |
|                       | specifications-and-pr |                       |
|                       | ocedures .unnumbered} |                       |
+-----------------------+-----------------------+-----------------------+
| # 6.4 Handling, St    | # CHAPTER 11: HANDLI  | # XII                 |
| orage and Delivery {# | NG, STORAGE, ETC {#ch | I. Handling, Storage  |
| handling-storage-and- | apter-11-handling-sto | and Shipping {#xiii.- |
| delivery .unnumbered} | rage-etc .unnumbered} | handling-storage-and- |
|                       |                       | shipping .unnumbered} |
+-----------------------+-----------------------+-----------------------+
| # 6                   | # CHAPER 8: NONC      | # XV. Non             |
| .5 Nonconforming Mate | ONFORMING ARTICLE AND | conforming Materials, |
| rial {#nonconforming- |  MATERIAL CONTROL {#c |  Parts, or Components |
| material .unnumbered} | haper-8-nonconforming |  {#xv.-nonconforming- |
|                       | -article-and-material | materials-parts-or-co |
|                       | -control .unnumbered} | mponents .unnumbered} |
+-----------------------+-----------------------+-----------------------+
| # 6.6 Statistical     | # CHAPTER 12          | #  {#se               |
| Quality Control and A | : SAMPLING PLANS, STA | ction-20 .unnumbered} |
| nalysis {#statistical | TISTICAL PLANNING AND |                       |
| -quality-control-and- |  ANALYSIS {#chapter-1 |                       |
| analysis .unnumbered} | 2-sampling-plans-stat |                       |
|                       | istical-planning-and- |                       |
|                       | analysis .unnumbered} |                       |
+-----------------------+-----------------------+-----------------------+
| # 6.7 Indication of I | # CHAP                | # XIV. Inspection,    |
| nspection Status {#in | TER 10: STAMP CONTROL |  Test, and Operating  |
| dication-of-inspectio | S {#chapter-10-stamp- | Status {#xiv.-inspect |
| n-status .unnumbered} | controls .unnumbered} | ion-test-and-operatin |
|                       |                       | g-status .unnumbered} |
+-----------------------+-----------------------+-----------------------+
| # 7.2 Government P    | # CHAPTER 1           | #  {#se               |
| roperty {#government- | 3: GOVERNMENT PROPERT | ction-21 .unnumbered} |
| property .unnumbered} | Y CONTROL {#chapter-1 |                       |
|                       | 3-government-property |                       |
|                       | -control .unnumbered} |                       |
+-----------------------+-----------------------+-----------------------+
| #  {#se               | # 1B205               | XVIII\. Audits        |
| ction-22 .unnumbered} | QUALITY ROGRAM AUDITS |                       |
|                       |  {#b205-quality-rogra |                       |
|                       | m-audits .unnumbered} |                       |
+-----------------------+-----------------------+-----------------------+

Table 1I ANSI/ASME NQA-1-1979 STRUCTURE

+---+-----------------+--------------------------+---------------------+
|   | **BASIC         | **SUPPLEMENTS**          | **APPENDICES**      |
|   | REQUIREMENTS**  |                          |                     |
+---+-----------------+--------------------------+---------------------+
| 1 | Organization    | 1S-1 Organization        | 1A-1 Organization   |
+---+-----------------+--------------------------+---------------------+
| 2 | Quality         | 2S-1 Qualification of    | 2A-1 Qualifications |
|   | Assurance       | Inspection and Test      | of Inspection and   |
|   | Program         | Personnel                | Test Personnel      |
|   |                 |                          |                     |
|   |                 | 2S-2 Qualification of    | 2A-2 Quality        |
|   |                 | Nondestructive           | Assurance Programs  |
|   |                 | Examination Personnel    |                     |
|   |                 |                          | 2A-3 Education and  |
|   |                 | 2S-3 Qualification of    | Experience of Lead  |
|   |                 | Audit Personnel          | Auditors            |
|   |                 |                          |                     |
|   |                 | 2S-4 Personnel           |                     |
|   |                 | Indoctrination and       |                     |
|   |                 | Training                 |                     |
+---+-----------------+--------------------------+---------------------+
| 3 | Design Control  | 3S-1 Design Control      | 3A-1 Design Control |
+---+-----------------+--------------------------+---------------------+
| 4 | Procurement     | 4S-1 Procurement         | 4A-1 Procurement    |
|   | Document        | Document Control         | Document Control    |
|   | Control         |                          |                     |
+---+-----------------+--------------------------+---------------------+
| 5 | Instructions,   |                          |                     |
|   | Procedures, and |                          |                     |
|   | Drawings        |                          |                     |
+---+-----------------+--------------------------+---------------------+
| 6 | Document        | 6S-1 Document Control    |                     |
|   | Control         |                          |                     |
+---+-----------------+--------------------------+---------------------+
| 7 | Control of      | 7S-1 Control of          | 7A-1 Control of     |
|   | Purchased Items | Purchased Items and      | Purchased Items and |
|   | and Services    | Services                 | Services            |
+---+-----------------+--------------------------+---------------------+
| 8 | Identification  | 8S-1 Identification and  |                     |
|   | and Control of  | Control of Items         |                     |
|   | Items           |                          |                     |
+---+-----------------+--------------------------+---------------------+
| 9 | Control of      | 9S-1 Control of          |                     |
|   | Processes       | Processes                |                     |
+---+-----------------+--------------------------+---------------------+
| 1 | Inspection      | 10S-1 Inspection         |                     |
| 0 |                 |                          |                     |
+---+-----------------+--------------------------+---------------------+
| 1 | Test Control    | 11S-1 Test Control       |                     |
| 1 |                 |                          |                     |
+---+-----------------+--------------------------+---------------------+
| 1 | Control of      | 12S-1 Control of         |                     |
| 2 | Measuring and   | Measuring and Test       |                     |
|   | Test Equipment  | Equipment                |                     |
+---+-----------------+--------------------------+---------------------+
| 1 | Handling,       | 13S-1 Handling, Storage, |                     |
| 3 | Storage, and    | and Shipping             |                     |
|   | Shipping        |                          |                     |
+---+-----------------+--------------------------+---------------------+
| 1 | Inspection,     |                          |                     |
| 4 | Test, and       |                          |                     |
|   | Operating       |                          |                     |
|   | Status          |                          |                     |
+---+-----------------+--------------------------+---------------------+

  ---- ------------------ --------------------------- ----------------------
  15   Control of         15S-1 Control of            
       Nonconforming      Nonconforming Items         
       Items                                          

  16   Corrective Action                              16A-1 Corrective
                                                      action

  17   Quality Assurance  17S-1 Quality Assurance     17A-1 QA records
       Records            Records                     

  18   Audits             18S-1 Audits                18A-1 Audits
  ---- ------------------ --------------------------- ----------------------

**[ATTACHMENT: ASME NQA-1-2004 BASIC REQUIREMENTS AND
PRACTICES]{.underline}**[^65]

**BASIC REQUIREMENT 1: ORGANIZATION**

**Responsibilities for the establishment and implementation of the
quality assurance program shall be defined. The organizational
structure, functional responsibilities, levels of authority and lines of
communication for activities affecting quality shall be documented.**

.

-   Management is responsible for establishing and ensuring the adequacy
    and effectiveness of the organization's quality assurance program.

-   The quality assurance staff assists management in establishing the
    quality assurance program and ensuring its effectiveness but top or
    senior managers are ultimately responsible for quality.

-   The quality assurance staff is provided an open line to management
    to resolve quality problems.

-   Organizational independence means that the quality assurance staff
    is allowed to assess and report on the quality assurance program
    compliance to top and senior managers but does not preclude the
    quality assurance staff from contributing to mission success and
    continual quality improvement.

**BASIC REQUIREMENT 2: QUALITY ASSURANCE PROGRAM**

**(a) A documented quality assurance program shall be planned,
implemented and maintained in accordance with this Part (Part 1), or
portions thereof. The program shall identify the activities and items to
which it applies. The program shall provide for control over activities
affecting quality to an extent consistent with their importance. The
program shall include monitoring activities against acceptance criteria
in a manner sufficient to provide assurance that the activities
affecting quality are performed satisfactorily. The program shall be
established at the earliest time consistent with the schedule for
accomplishing the activities.**

**The program shall provide for the planning and accomplishment of
activities affecting quality under suitably controlled conditions.
Controlled conditions include the use of appropriate equipment, suitable
environmental conditions for accomplishing the activity, and assurance
that prerequisites for the given activity have been satisfied. The
program shall provide for any special controls, test equipment, tools,
and skills to attain the required quality of activities and items and
for verification of that quality. The organization shall establish and
implement processes to detect and correct quality problems.**

\(b\) The program shall provide for the indoctrination, training and
qualification as necessary of personnel performing or managing
activities affecting quality to assure that suitable proficiency is
achieved and maintained.

\(c\) Management shall regularly assess the adequacy and effectives of
the quality assurance program.

-   The quality assurance program is the composite of plans and actions
    that define a management system, not just a document.

-   The quality assurance program considers public and worker safety,
    environmental protection, and mission success.

-   The management organization having overall responsibility for the
    quality assurance program decides what parts it will implement and
    what parts it will assign to others to implement.

-   For expected results, those who are to achieve these results must
    know what to do, have the resources and be competent to do it.

-   Realistic quality objectives are established and measured to provide
    managers sufficient confidence that the quality assurance program
    policies and requirements will be achieved as planned**.**

    **BASIC REQUIREMENT 3: DESIGN CONROL**

    **The design shall be defined, controlled, and verified. Design
    inputs shall be specified on a timely basis and translated into
    design documents. Design interfaces shall be identified and
    controlled. Design adequacy shall be verified by individuals other
    than those who designed the item or computer program. Design changes
    shall be governed by control measures commensurate with those
    applied to the original design.**

-   The design is the solution to a defined need, not just the
    documents.

-   Design is accomplished through a controlled process.

-   The objective of design control is to provide confidence that what
    is required or desired will result by implementing the design.

-   Design interface control encompasses both the design process and
    design organizations.

-   In most cases, there is significant time and many actions that are
    required between design completion and full implementation of the
    design for construction and operation.

-   Peer review is one way to verify design adequacy

-   Design changes are controlled to preserve confidence in the design.

    **BASIC REQUIREMENT 4: PROCUREMENT DOCUMENT CONTROL**

    **Applicable design bases and other requirements necessary to ensure
    adequate quality shall be included or referenced in documents for
    procurement of item and services. To the extent necessary,
    procurement documents shall require Suppliers to have a quality
    assurance program consistent with the applicable requirements of
    this Standard.**

-   Procurement of items and services involves a legally enforceable
    contract between the buyer and supplier.

-   To ensure adequate quality, the contract documents is clearly stated
    and understood by the Supplier complete in their content, and
    legally enforceable.

-   Procurement documents define both the technical requirements,
    including the quality characteristics of the item or service being
    procured, as well as the quality assurance activities that have been
    delegated to the Supplier to implement under the contract.

-   Procurement documents and their changes are controlled.

    **BASIC REQUIREMENT 5: INSTRUCTIONS, PROCEDURES, AND DRAWINGS**

    **Activities affecting quality and services shall be prescribed by
    and performed in accordance with documented instructions,
    procedures, and drawings that include or reference appropriate
    quantitative or qualitative acceptance criteria for determining that
    prescribed results have been satisfactorily attained. The activity
    shall be described to a level commensurate with the complexity of
    the activity and the need to assure consistent and acceptable
    results. The need for and level of detail in written procedures or
    instructions shall be determined based on complexity of the task,
    work environment, and worker proficiency and capability (education,
    training, experience).**

-   Most failures to achieve quality in items and services can be traced
    to poor communications.

-   NQA-1 requirements underscore the importance of conformance to
    instructions, procedures, and drawings to prevent or overcome
    failures due to inadequate communications.

-   Inadequate communications can results in failure to be aware of and
    follow properly updated and maintained procedures and instructions.

-   Instructional material is in a form and contains substance that is
    readily understood by users.

    **BASIC REQUIREMENT 6: DOCUMENT CONTROL**

    The preparation, issue, and change of documents that specify quality
    requirements of prescribed activities affecting quality such as
    instructions, proce3dures, and drawings shall be controlled to
    assure that correct documents are being employed. Such documents,
    including changes thereto, shall be reviewed for adequacy and
    approved for release by authorized personnel.

+-----------------------------------------------------------------------+
| -   The right documents must be in the right places, in the right     |
|     hands, and recognizable as being the correct authorized version   |
|     for the work being performed.                                     |
|                                                                       |
| -   Obsolete versions of documents are controlled to prevented misuse |
|     on items and activities important to quality.                     |
+-----------------------------------------------------------------------+
|                                                                       |
+-----------------------------------------------------------------------+

**BASIC REQUIREMENT 7: CONTROL OF PURCHASED ITEMS AND SERVICES**

**The procurement of items and services shall be controlled to assure
conformance with specified requirements. Control shall provide for the
following, as appropriate: source evaluation and selection, evaluation
of objective evidence of quality furnished by the Supplier, source
inspection, audit, and examination of items and services upon delivery
or completion.**

-   The buyer make sure by monitoring that the supplier knows what is
    important and why before there is an opportunity for supplier
    produces an unacceptable item.

-   The buyer plans and performs actions that direct and control
    supplier performance to ensure satisfaction of contracted
    requirements.

-   Evidence of suppler performance may only be captured during the
    event necessitating pre-arrangements to gather such evidence.

-   Cooperation between the buyer and supplier is mutually beneficial
    for contract oversight.

    **BASIC REQUIREMENT 8: IDENTIFICATION AND CONTROL OF ITEMS**

    **Controls shall be established to ensure that only correct and
    accepted items are used or installed.**

    **Identification shall be maintained on the item or in documents
    traceable to the items, or in a manner that assures that
    identification is established and maintained.**

-   Any item that is important to quality must be identifiable and
    traceable in the event of a problem.

-   Item identification is necessary to ensure the correct item is
    installed and used.

-   Item identification is essential to effective configuration
    management and maintenance of limited shelf life materials.

> **BASIC REQUIREMENT 9: CONTROL OF SPECIAL PROCESSES**
>
> **Special processes that control or verify quality, such as those used
> in welding, heat treating, and nondestructive examination, shall be
> performed by qualified personnel using qualified procedures in
> accordance with specified requirements.**

-   Quality assurance programs, as defined by NQA-1, generally apply to
    the control classical special processes, such as welding and
    nondestructive examination.

-   There may be other special processes that meet the intent of NQA-1
    definition and that also need special skills, training, procedures,
    and environmental controls to assure quality.

    **BASIC REQUIREMENT 10: INSPECTION**

    **Inspection required to verify conformance of an item or activity
    to specified requirements or continued acceptability of items in
    service shall be planned and executed. Characteristics subject to
    inspection and inspection methods shall be specified. Inspection
    results shall be documented. Inspection for acceptance shall be
    performed by qualified persons other than those who performed or
    directly supervised the work being inspected.**

-   To maximize value, inspections are planned, performed in accordance
    with procedures or instructions, results analyzed, and documented.

-   Inspections may be performed for various reasons beyond simply
    measuring conformance to requirements such as operational or
    in-service preventive maintenance.

-   Inspections for acceptance are performed by those independent to
    those who performed the work.

    .

    **BASIC REQUIREMENT 11: TEST CONTROL**

    Tests required to collect data such as for siting or design input,
    to verify conformance of an item, or to demonstrate satisfactory
    performance for service shall be planned and executed.
    Characteristics to be tested and test methods to be employed shall
    be specified. Test results shall be documented and their conformance
    with test requirements and acceptance criteria shall be evaluated.

-   Testing is performed for a multitude of purposes besides verifying
    conformance to specified requirements. Examples include development
    testing where acceptance criteria have not yet been established.

-   Regardless of purpose, testing is planned, performed in accordance
    with procedures, results analyzed, and documented.

    **BASIC REQUIREMENT 12: CONTROL OF MEASURING AND TEST EQUIPMENT**

    **Tools, gages, instruments, and other measuring and test equipment
    used for activities affecting quality shall be controlled,
    calibrated at specified intervals, adjusted, and maintained to
    required accuracy limits.**

-   It is important to understand the risks associated with using
    unreliable or uncelebrated M&TE

-   M&TE precision and accuracy must be understood when selecting M&TE.

    **BASIC REQUIREMENT 13: HANDLING, STORAGE, AND SHIPPING**

    **Handling, storage, cleaning, packaging, shipping and preservation
    of items shall be controlled to prevent damage or loss and to
    minimize deterioration. These activities shall be controlled in
    accordance with established work and inspection instructions,
    drawings, specifications, shipment instructions, or other pertinent
    procedures specified for conducting the activity.**

-   Improper protection of high-cost or sensitive items can results in
    avoidable costs.

-   The credibility of expensive items that have been qualified and
    certified and damaged subsequently is compromised.

-   Improper handling, lifting and rigging of items has resulted in
    numerous incidents of personal injury.

    **BASIC REQUIREMENT 14: INSPECTION, TEST, AND OPERATING STATUS**

    **The status of inspection and test activities shall be identified
    either on the items or in documents traceable to the items where it
    is necessary to ensure that required inspections and tests are
    performed and to ensure that items that have not passed the required
    inspections and tests are not installed, used or operated.**

-   Methods for identifying and maintaining inspection and test status
    are identified, maintained and readily available to workers as work
    progresses.

-   Methods for identifying operating status, including temporary
    hookups changes, are essential to avoid serious injury, accidents or
    accidents.

    **BASIC REQUIREMENT 15: CONTROL OF NONCONFORMING ITEMS**

    **Items that do not conform to specified requirements shall be
    controlled to prevent inadvertent installation or use. Controls
    shall provide for identification, documentation, evaluation,
    segregation when practical, and disposition of nonconforming items,
    and for notification to affected organizations.**

-   Controlling nonconforming items is essential to assure the
    acceptability of the completed item, its readiness to perform an
    intended function, and maintaining contractual and regulatory
    compliance.

-   Improper dispositioning of nonconforming items, such as
    suspect/counterfeit items, may have hazardous consequences.

-   Nonconformances are evaluated for lessons learned.

    **BASIC REQUIREMENT 16: CORRECTIVE ACTION**

    **Conditions adverse to quality shall be identified promptly and
    corrected as soon as practicable. In the case of significant
    conditions adverse to quality, the cause of the condition shall be
    determined and corrective action taken to prevent recurrence. The
    identification, cause, and corrective action for significant
    conditions adverse to quality shall be documented and reported to
    appropriate levels of management. Completion of corrective actions
    shall be verified.**

-   Corrective action goes beyond the correction of a deviation from a
    specification. Depending upon the significance, corrective action
    includes:

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-   Identification of the causes of the nonconforming conditions adverse
    to quality.

-   Determination of similarity to previous nonconformances

-   Deciding on preventive action to preclude reoccurrence

-   Communicating occurrence to other organizations that may benefit
    from the corrective and preventive actions taken.

    **BASIC REQUIREMENT 17: QUALITY ASSURANCE RECORDS**

    **Quality assurance records shall furnish documentary evidence that
    items or services meet specified quality requirements. Quality
    assurance records shall be identified, generated, authenticated, and
    maintained, and their disposition specified. Requirements and
    responsibilities for these activities shall be documented. The term
    records, used throughout this section, is to be interpreted as
    quality assurance records.**

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-   To have confidence in achieved quality and enable future work and
    investigations, the results must be recorded and readily available.

-   Some nuclear power plants were not completed and were prematurely
    decommissioned, due in part to the lack of recorded evidence of work
    performed.

-   Adequate quality assurance records reflect well planned and managed
    work

-   Adequate quality assurance records may play a pivotal role in
    decommissioning plants and dispositioning their components.

    **BASIC REQUIREMENT 18: AUDITS**

    **Audits shall be performed to verify that performance criteria are
    met and to determine the effectiveness of the program. These audits
    shall be performed in accordance with written procedures or
    checklists by personnel who do not have direct responsibility for
    performing the activities being audited. Audit results shall be
    reported to and reviewed by responsible management. Follow-up action
    shall be taken where indicated.**

-   A quality audit is a disciplined process for assessing performance
    of quality assurance activities in support of judgments of
    responsible managers regarding adequacy and effectiveness.

-   Such audits are not to be used to determine product acceptance but
    may influence this determination.

-   An audit is a sampling process and, therefore, to be effective, is
    conducted by trained, experienced and qualified auditors who are
    skilled in drawing conclusions concerning significance.

-   From these conclusions, responsible management determines how to
    correct an observed inadequacy and improve upon the program.

-   An audit may also identify best or exemplary practices and
    recommended actions for improvement.

-   Audits may be supplemented by real-time surveillance of work in
    progress or other evaluations.

-   Audits provide input to management reviews and assessments

[^1]: Merritt E. (Gene) Langston of SAIC is the principle author of and
    ASME NQA Project Manager for PARTS I and II of this document. Gene
    co-authored 10 CFR 50 Appendix B, RDT F2-2T and DOE Order 5700.6. He
    is a charter member of the former ASME N45-3 Subcommittee; and a
    current member of the ASME NQA Main Committee. Contributions to this
    document by other ASME Main Committee members are recognized in the
    text and footnotes throughout the document.

[^2]:
    #######  Contributed by Joe Anderson, former ASME N45-2 Subcommittee Chairman and currently NQA Main Committee and Applications Subcommittee member. 

    #######  {#section-1}

[^3]: Address by VADM H.G. Rickover, "The Never-Ending Challenger",
    44^th^ Annual Metals Congress, New York, NY, October 29, 1962.

[^4]: Address by AEC Commissioner James T. Ramey, "Quality Assurance as
    a Matter of Public Policy in the Safety 0f Atomic Power Plants",
    1966 Winter Meeting of the American Nuclear Society, Pittsburgh, PA,
    November 2, 1966.

[^5]: MIL-STD-109, Quality Assurance Terms and Definitions, (Source:
    MIL-STD-490. SPECIFICATION PRACTICES, 30 October 1968)

[^6]: Reference to MIL-STD-109 definition of quality assurance
    contributed by Robert Hartstern, member, ASME Nuclear Quality
    Assurance Committee.

[^7]: Address by AEC Commissioner James T. Ramey, Quality Assurance --
    An Essential for Safe and Economic Nuclear Power", American Power
    Conference, Chicago, IL, April 23, 1968.

[^8]: Department of Defense (DOD) Military Specification MIL-Q 9858^?^,
    Quality Program Requirements, issued in April 1959, superseded by
    MIL-Q-9858A in December 1963.

[^9]: Depart of Energy Assistant Secretary for Military Applications,
    Quality Assurance Policy, July 7, 1989.

[^10]: Depart of Energy Assistant Secretary for Military Applications,
    Quality Assurance Policy, November 20, 1989.

[^11]: DOE Defense Programs Memorandum, Implementation of Department of
    Energy Order 5700.6C, "Quality Assurance", February 27, 1992.

[^12]: DOE/NNSA Weapons Quality Policy (QC-1), 02/10/2004

[^13]: AEC QRC-82C, Quality Program Requirements, June 7, 1964.

[^14]: NASA NHB 5300.4(1B), Quality Program Provision for Aeronautical
    and Space System Contractors, April 1969 Edition (formerly NPC
    200.2, April 1962.

[^15]: MIL-I-45208A, Military Specification, Inspection System
    Requirement. 16 December 1963

[^16]: Address by Merritt E. Langston, "Quality Assurance Requirements
    for Reactor Development Programs", 26^th^ Annual Technical
    Conference of the American Society for Quality Control, Washington,
    D.C, May 10, 1972.

[^17]: Assistant Secretary for Energy Technology, Director of Nuclear
    Energy, Quality Assurance Policy for Nuclear Energy Program, 9-1-78

[^18]: U.S. Department of Energy Order, DOE 5700.6, Quality Assurance,
    1-16-81.

[^19]: U.S. Department of Energy Order, DOE 5700.6C, Quality
    Assurance,8-21-91.

[^20]: The remainder of this section on DOE Orders , Rules and guides
    was contributed by Bud Danielson, ASME NQA Main Committee member;
    and Chairman, NQA Subcommittee on Applications.

[^21]: U.S. Department of Energy Order 414.1, Quality Assurance,
    11-24-98

[^22]: U.S. Department of Energy Order 414.1A, Quality Assurance,
    5-19-99.

[^23]: U.S. Department of Energy Order 414.1B, Quality Assurance,
    4-29-04.

[^24]: DOE G 414.1-3, Suspect Counterfeit Items Guide for use with 10
    CFR 830 subpart A, Quality Assurance Requirements, and DOE O 414.1B,
    Quality Assurance.

[^25]: U.S. Department of Energy, 10 CFR PART 830, December 9, 1991.

[^26]: DOE Order 414.1C, Quality Assurance, 6-17-05

[^27]: DOE G 414.1-4, Safety System Software

[^28]: DOE G 414.1-1, Implementation Guide for Use with Independent and
    Management Assessment Requirements of 10 CFR Part 830.120 and DOE
    5700.6C, Quality Assurance, August 1996.

[^29]: DOE G 414.1-2, Quality Assurance Guide for Use with 10 CFR
    830.120 and DOE O 414.1, DRAFT June 1998.

[^30]: DOE Action Plan, Lessons learned from the Columbia Space Shuttle
    Accident and Davis-Besse Reactor Pressure Vessel Head Corrosion
    Event, July 2005.

[^31]: Address by Adm H.G. Rickover, "Advice to Prospective Purchasers
    of Central Station Nuclear Power Plants", AEC Authorizing
    Legislation Hearings before the Congressional Joint Committee on
    Atomic Energy, Washington, D.C., March 14 & 15, 1967.

[^32]: Contributed by Doug Brown, former ASME NQA MC Chair and currently
    MC member.

[^33]: W.M. (Bill) Morrison was the first AEC/NRC representative to the
    ASME N45-3 and ASME NQA Committees

[^34]: Minutes of Commonwealth Edison Company's (CE) Zion Reactors
    Pre-hearing Inspection Meeting Held on November 26, 1968.

[^35]: "AEC Regulatory Staff Evaluation of Commonwealth Edison's Quality
    Assurance Program for the Zion Station", December 17, 1968.

[^36]: Appendix B to Quality Assurance Criteria for Nuclear Power
    Plants, Federal Register, vol 34, no 73, April 17, 1969.

[^37]: Principal changes to Title 10, Chapter 1, Atomic Energy
    Commission, Part 50, Licensing of Production and Utilization
    Facilities, Quality Assurance Criteria for Nuclear Power Plants,
    undated.

[^38]: Remarks by E.G. Case at the ASME Winter Annual Meeting, "Quality
    Assurance for Design of Nuclear Power Plants", November 18, 1969.

[^39]: Title 10 -- Atomic Energy Commission, Part 50 -- Licensing of
    Production and Utilization Facilities, Quality Assurance Criteria
    for Nuclear Power Plants, Federal Register, vol 35, no. 125, June
    27, 1970.

[^40]: Atomic Energy Commission, "Proposed Amendment to 10 CFR Part 50,
    Quality Assurance Criteria for Fuel Reprocessing Plants" Marc 8,
    1971

[^41]: Appendix B to Part 50, Quality Assurance Criteria for Nuclear
    Power Plants and Fuel Reprocessing Plants, September 11, 1971.

[^42]: "Improving Quality and the Assurance of Quality in the Design and
    Construction", NUREG 1055. 1984

[^43]: American National Standard, ANSI N45.2-1971, \"Quality Assurance
    Program Requirements for Nuclear Power Plants\", October 20, 1971.

[^44]: Contributed by S. Bernsen, former ASME N45-2 Subcommittee
    chairman and ASME NQA Committee member.

[^45]: S.A. Bernsen and S.K. Hellman, Nuclear Power Plant Quality
    Assurance Standards, The Status and Application of ANSI N45-2
    Standards, July 1973.

[^46]: M.E. Langston and A.W. Crevasse, Nuclear Standards and Federal
    Regulations -- Where Are We and Where Are We Going? Presented at the
    Third Annual National Conference on Nuclear Power.

[^47]: Contributed by Charles Moseley, NQA Committee member and ANS 3-2
    Committee liaison

[^48]: Introductory Remarks by J. W. Landis extracted from the
    Proceedings of the Executive Conference on Quality Assurance held in
    San Diego in April 1976.

[^49]: ASME NQA-1-1979. Quality Assurance Program Requirements for
    Nuclear Facilities, August 31, 1979.

[^50]: ASME NQA-1-1983, issued July 11, 1983

    ASME NQA-1a-1983, issued December 31, 1983

    ASME NQA-1b-1883, issued March 15, 1985

[^51]: ASME NQA-1-1986. issued July 11, 1986

    ASME NQA-1a-1986, issued February 15, 1997

    ASME NQA-1b-1987, issued March 15, 1988

    ASME NQA-1c-1988, issued February 28, 1989

[^52]: ASME NQA-2-1983, Quality Assurance Requirements for Nuclear
    Facility Applications

[^53]: ASME NQA-1-1989, "Quality Assurance Program Requirements for
    Nuclear Facilities, September 15, 1989

[^54]: ASME NQA-1-1983, "Quality Assurance Program Requirements for the
    collection of Scientific and Technical Information for Site
    Characterization of High-Level Nuclear Waste Repositories" March 23,
    1990.

[^55]: ASME NQA-1-1994, \"Quality Assurance Requirements for Nuclear
    Facility Applications"

[^56]: Contributed by Sid Bernsen, op. cit.

[^57]: ASME NQA-1-1997, :Quality Assurance Program Requirements for
    Nuclear Facility Applications", December 31, 1997

[^58]: ASME NQA-1a-1999, May 25, 1999

[^59]: ASME NQA-1-2000, ", May 21, 2001Quality Assurance Program
    Requirements for Nuclear Facility Application"

[^60]: ASME NQA-1-2004, Quality Assurance Requirements for Nuclear
    Facility Applications, December 22, 2004

[^61]: Contribute by Sid Bernsen.

[^62]: Contributed by Doug Brown.

[^63]: Contributed by T.E. Dunn, Vice Chairman, ASME NQA Committee and
    Executive Committee member.

[^64]:
    #  Contributed by John Adkins, Chairman, ASME NQA Committee {#contributed-by-john-adkins-chairman-asme-nqa-committee .unnumbered}

    #  {#section-17 .unnumbered}

[^65]: Contributed by Joe. Anderson