history of the code rules for accreditation, certification, and related issues

CHAPTER

33

33.1 INTRODUCTION This chapter provides an overview of the history of the ASME

Boiler and Pressure Vessel (B&PV) Code from its inceptionthrough its incorporation of nuclear components and up to thepresent. Applicable rules for certification and accreditation weredeveloped in parallel with the growth of the Code. From earliercoverage of boilers and pressure vessels, the nuclear initiativerequired coverage for piping, pumps, valves, storage tanks, vesselinternals, and both component and pipe supports.

The earlier Codes not within the scope of the Boiler and PressureVessel Code are described, including their part in the developmentof the rules incorporated into the Nuclear Code in 1971. Codeimplementers, including the Registered Professional Engineer aswell as the Authorized Nuclear Inspector and his or her supervisor,are highlighted, as are the organizations that employ them.Authorized Inspection and Quality Assurance are discussed, as wellas Inservice Inspection and its interfaces with the ConstructionCode. Rules for repairs and replacements of nuclear componentsand the use of newer codes are referenced, including the need forcode reconciliation and commercial grade dedication. Two exam-ples of how Code reconciliation is used are provided.

The development of certification and accreditation are covered,with emphasis on the new requirements for organizations seekingASME accreditation. The globalization of the ASME certificatesand stamps are thoroughly described.

33.2 HISTORICAL BACKGROUND A detailed article titled History of the ASME Boiler Code

was written by Dr. Arthur M. Greene, Jr., and was first publishedin various issues of Mechanical Engineering Magazine in 1952and 1953. Later, it was published as a book by the AmericanSociety of Mechanical Engineers (ASME) in 1955. The followingparagraphs are excerpted from this publication [1].

As a result of the many boiler explosions experienced duringthe nineteenth century, a committee was formed in 1897 underthe American Boiler Manufacturers Association (ABMA) to

develop uniform specification laws. Due to inability by somemembers of the association to look beyond the interests oftheir own companies, the proposed rules were not approved.

On August 30, 1907, the Commonwealth of Massachusettsapproved the first set of rules for construction of boilers proposedby a committee headed by John A. Stevens. By 1909, the originalrules, consisting of three pages, were expanded to three parts andapproved as An Act for the Operation and Inspection of SteamBoilers. The State of Ohio did likewise on October 24, 1911,approving a code essentially identical to the Massachusetts rules.This code went into effect on January 1, 1912, and Part 3 becamemandatory on July 1, 1912.

In June 1911, Col. E. D. Meier, president of the Heine BoilerCompany and a past president of the ABMA, became president ofthe ASME. With the ABMA rejection still in his mind, Col. Meierbelieved that a set of rules formulated by ASME, with its reputa-tion and broad scientific interests which made it commerciallydisinterested, might be accepted. In September, 1911, he askedthe Council of the ASME to approve appointment of a committeeto formulate standard specifications for the construction of steamboilers and other pressure vessels. After four years of hard workand negotiations, better described in the above cited book, thestandard, based on the Massachusetts rules, was approved byCouncil on March 12, 1915, as the Rules for the Construction ofStationary Boilers and for Allowable Working Presssures.

The first meeting of the Boiler Code Committee was held inBuffalo, NY, on June 23, 1915, and among its first considerationswas the question of Code Symbol Stamping. Authorization to usethe Code symbol would be referred to the Boiler Code Committeefor recommendation and report to the Council. The committeealso adopted the statement that the Code Symbol applied to aboiler would indicate that the boiler had been built in full compli-ance with the Code, and that the stamp should be applied by themanufacturer. It is interesting to note that Code Case No. 10asked where the Code stamp specified in par. 332 of the Codemight be obtained, and the reply was through ASME at a priceof $3 each. These were the rules for certification of boiler manu-facturers: compliance with the code rules, and stamping the boilerwith a stamp procured from ASME.

HISTORY OF THE CODE RULESFOR ACCREDITATION, CERTIFICATION,

AND RELATED ISSUESMarcus N. Bressler

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Through the years it only required a letter from the jurisdiction(a State of the United States or a Province of Canada) to permit amanufacturer to secure a Code Symbol Stamp. On March 2, 1922,the Boiler Code Committee directed that a new certificate and aconsecutive number be issued to each company holding a stampfor the purpose of registration. This registry recorded the issuanceof 222 stamps as of April 25, 1924. The minutes of the 1937 to1940 meetings reported, for the first time, the names of the com-panies to which the Code Symbol Stamp had been issued or reis-sued, with an indication of the section or sections for which thesewould be used. There were 108 in 1938, 109 in 1939, and 124 in1940. Certification remained an unstructured program for manyyears.

33.3 NUCLEAR ENERGY Nuclear energy was first harnessed for the production of steam

to drive ship turbines in the early 1950s. The first applicationswere military; specifically, they were for the propulsion of sub-marines. The standards used for the construction of the requiredequipment were necessarily developed by the U.S. Navy and alsoby the contractors involved in designing and manufacturing thenuclear power plants for early prototype submarines. The firstcommercial operating nuclear power plant, Shippingsport, placedinto service a duplicate of a ships reactor near Pittsburgh,Pennsylvania. In 1958, the ASME B&PV Committee created theSpecial Committee to Review Code Stress Basis, charging it withthe responsibility of developing rules for the safe construction ofpressure vessels of superior quality. Admiral Hyman Rickoverpersuaded the committee to first develop a standard for nuclearvessels to permit the Navy to bid competitively among the nationspressure vessel manufacturers. That action marked the first timethat a nuclear Code would be prepared and become the globalStandard for nuclear power plant construction. It became SectionIII, Nuclear Vessels, described in paragraph 33.5.

33.4 PIPING, VESSELS, PUMPS,AND VALVES IN THE 1950s

When the early commercial nuclear power plants were consid-ered in the late 1950s, the available Codes and Standards werethose used for the construction of thermoelectric power plants,refineries, and chemical plants. This in itself was not a problem,as more than 60% of a nuclear power plant is essentially the sameas that of any other thermal or chemical plant in terms of its pip-ing, vessels, pumps, valves, and supports.

The late 1950s was a period of quantum leaps. The develop-ment of calculating equipmentfrom hand-cranked and electrifiedadding machines to gas-tube electronic calculators and gas-tubecomputing machinesreached a high level of sophistication bythe end of the 1950s. Stress analysis techniques and the high-speed calculations permitted by the new computers led to largeincreases in the formulation of piping and vessel analyses. Newstandards and revisions to existing standards took advantage ofthe new tools in time for the needs of nuclear power.

A historical note may help explain the various prefixes to thesestandards. The American Standards Association (ASA) was reor-ganized in 1966 as the U.S.A. Standards Institute. Standardsapproved as American Standards were designated U.S.A.Standards (USAS). In 1969, the Institute was renamed the

American National Standards Institute (ANSI); its approvednational Standards became ANSI Standards. Throughout thoseyears, the ASME had administrative control of its Standardsunder procedures accredited by the ANSI. These are now listed asASME Standards; moreover, the ASME retains the sole responsi-bility for their interpretation.

In 1955, the B31 committee decided to publish the new pipingCode in separate volumes. At the beginning of the 1960s, pipingwas designed, fabricated, and installed in accordance with ASAB31.11955. Vessels were under the scope of the ASME B&PVCode, Section VIII (Unfired Pressure Vessels), 1959 edition, sum-mer and winter 1959 addenda.

Pumps and valves were designed and manufactured under manu-facturers standards, with dimensional parts complying with nationalstandards. For example, flanges for bonnet and casing covers aswell as for flanged pipe connections were manufactured to thedimensional standard ASA B16.51957. (For a history of the devel-opment of the flanges and flange fitting standards, as well as rulesfor nuclear valves, see ref. [2].) The pump and valve bodies andinternals were the responsibility of the manufacturer. The onlyother requirements were hydrostatic tests of the bonnets and bodiesperformed at 1.50 times the rating pressure at room temperature, asdeveloped in B16.5 or in accordance with the ManufacturersStandardization Society of the Valve and Fittings Industry (MSS)SP-61, Hydrostatic Testing of Steel Valves, and SP-66, PressureTemperature Ratings for Steel Butt-Welding End Valves. (Note:SP denotes Standard Practice.) The ratings were based onmaterial specifications, and their allowable stresses at temperature.Minimum wall-thickness for valve bodies were listed in the stan-dard, as well as dimensional standards for seven classes of pressure.(For a history of the development of the pump rules, see ref. [3].)

33.5 ASME SECTION III, NUCLEARVESSELS, IN THE 1960S

Section III, Nuclear Vessels, was first published in 1963, atwhich time it was the only National Standard completely dedicatedto nuclear applications. The second edition was published in 1965;the third, in 1968. Code editions were effective on July 1 of thepublication year and were published triennially. Semiannualaddenda were issued between Code editions and could be usedafter their publication; they became mandatory six months afterissue. The publication dates were on June 30 and December 31 ofthe publication year; they were referred to as Summer and Winteraddenda, respectively. Each edition was intended to include theprevious edition as modified by the six addenda published duringthe triennial period ending in the next edition. The 1963 editiononly included two addenda, because the next edition was pub-lished at the same time as the rest of the 1965 ASME B&PV Code.

The first ASME Section III Code made provisions for the con-struction of three classes of vessels: Class A, Class B, and Class C.Class A rules were intended for the construction of vesselsdesigned to contain nuclear fuel and reactor coolant within thereactor-coolant pressure boundary, as stated in paragraph N-131(a).Class A vessels were designed by analysis using the maximumshear stress theory of failure (Tresca Criterion). Primary stresses(general membrane, local membrane, and bending) were limited bythe stress intensities allowed for design conditions. The summa-tions of primary and secondary stresses were also determined forthe operating conditions, the results of which were compared to theallowable design stress intensities. An extensive discussion of stress


COMPANION GUIDE TO THE ASME BOILER & PRESSURE VESSEL CODE 585

intensity classification in vessels is beyond the scope of this chap-ter. Reference [4] provides such a discussion, as it is the commen-tary that describes the basis for Design by Analysis in Section IIIfor Class A vessels. This reference was reissued in 1969, Reference[5], and included the design criteria used in Section VIII, Division 2,which was published in July 1968.)

Class B rules covered primarily vessels used for containment,as noted in paragraph N-132. Class B vessels could be designedeither to the maximum stress theory used in Section VIII or, alter-natively, to the rules of the new Section VIII, Division 2, andwere similar to Class A vessel design.

Class C rules were intended for vessels used for the auxiliarysystems in the plant not classified as either Class A or Class B(paragraph N-133). Class C vessels could be entirely constructedto the rules of Section VIII, Division 1.

33.6 PIPING, PUMPS, AND VALVES IN THE1960s

While the ASME B&PV Code Section III was developing, theB31 Standards Committee was reorganizing and, in keeping abreastof current technological improvements, issued a new individualstandardUSAS B31.1.0 (in title only)in 1967 until the revisionof B31.11955 could be completed. A separate subcommittee,which included many personnel involved in the development ofSection III, started the development of a code for nuclear pipingand another for nuclear pumps and valves. In 1968, the draft codefor pumps and valves was issued for trial use and comment; in1969, the USAS B31.7 Code for Nuclear Piping was published.

Although the Data Report Form NP-1 referred to the Certificate ofAuthorization Number for the Fabricator, the text of the B31.7 Codemade no reference to requirements for certification of the piping fab-ricator by the ASME. It was not until the piping code was absorbedby Section III in the 1971 edition that requirements for certificationwere included and Code Symbol Stamps were established.

Section III, 1968 edition with summer 1969 addenda, intro-duced paragraph N-153, which stated the following: Piping thatis part of a nuclear energy system and is required to be constructedin accordance with this code, shall meet the requirements forClass I piping of USAS B31.7, Code for Nuclear Piping. Also,the paragraph made the provision that pumps and valves shallmeet the requirements for Class I pumps and valves of the ASMECode for Pumps and Valves for Nuclear Power. In addition tothose references, paragraph N-153 specified that piping, pumps,and valves would need the required inspections to be performedby qualified inspectors, in accordance with paragraph N-612, andalso that they would need to be stamped, in accordance with para-graphs N-811 and N-815 to N-818 (inclusive).

Section III, winter 1969 addenda, published a revision to para-graph N-153. This clarified that the requirements for inspectionand stamping applied to piping, pumps, and valves when theywere required to be constructed in accordance with this code. Inaddition, the paragraph clarified that this requirement appliedwhen the piping was classified as Class I, USAS B31.7, and alsowhen pumps and valves were classified as Class I, ASME Codefor Pumps and Valves.

As early as 1967, the Atomic Energy Commission (AEC) pub-lished criteria for the codes and standards to be used for nuclearplant construction. In January 1975, the AEC was renamed theNuclear Regulatory Commission (NRC); it continued to be sensi-tive to national standards as it always had been.

The NRC, defined by the Section III Code as the RegulatoryAuthority, is responsible for administering the applicable federallaws that are listed in Title 10, Energy of the Code of FederalRegulations (CFR), PART 50: DOMESTIC LICENSING OF PRO-DUCTION AND UTILIZATION FACILITIES, Section 50.55,Conditions of Construction Permits, Section a, Codes and Standards.

The AEC stated in 10CFR50.55a that plants licensed before1967 could use the ASA B31.11955 Code, addenda, and applic-able Code Cases for piping within the reactor-coolant pressureboundary. Later revisions of 10CFR50.55a permitted use of theUSAS B31.1.01967 Code, addenda, and applicable Code Cases,as well as the Class I section of USAS B31.71969.

33.7 ACCREDITATION FOR NUCLEARCONSTRUCTION

When Section III, Nuclear Vessels, was first published in 1963,the ASME was not ready for a formal approach to certification ofmanufacturers of nuclear vessels. In a book by Wilbur Cross [6],Melvin R. Green, Managing Director of the ASME Codes andStandards Department, is quoted as follows:

In 1965, ASME included nuclear vessels in its CertificationProgram. A certificate was issued based on a favorable reportfrom the authorized inspection agency and the jurisdictionalauthority.

Section III introduced Mandatory Appendix IX, QualityControl and NDE Methods, in the winter 1967 addenda. With therevisions presented in the appendix, the ASME was now in posi-tion to establish procedures for certification: On July 1, 1968, theconcept of nuclear survey teams became mandatory.

Prior to 1968, ASME had depended on the jurisdictional bod-ies or inspection agencies for recommendations to use CodeSymbol Stamps. Then, in July 1968, more comprehensiveCode requirements were put into effect regarding applicantsfor nuclear accreditation. These requirements introducedQuality Assurance on a more formal basis and also initiatedthe use of nuclear survey teams. Since then, requirements forother sections of the Code evolved to require a review team;the revisions have maintained the principle that an authorizedinspection agency must have a potential legal or insuranceinterest in the finished product to be stamped with the ASMECode symbol stamp. The inspector must assure himself thatthe manufacturer conformed to the Code rules. The NationalBoard of Boiler and Pressure Vessel Inspectors (NationalBoard) acknowledged the value of this organizational proce-dure for reviewing Code Stamp applicants and began to par-ticipate in the survey teams in July 1968.

Audit teams began to visit manufacturers and their reports werereviewed by the Subcommittee on Code Symbol Stamps.

33.8 DEVELOPMENTS OF THE 1970sIn 1971, ASME Code Section III was renamed Nuclear Power

Plant Components and incorporated the USAS B31.71969 Codefor Nuclear Power Piping and its three addenda: ANSIB31.7a1971 (February 16, 1971); ANSI B31.7b1971 (March 10,1971); and ANSI B31.7c1971 (October 21, 1971). The draftASME Code for Pumps and Valves for Nuclear Power (dated


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November 1968) and its March 1970 addenda were also included.This Draft Code was never published as a separate ASME stan-dard. The renamed Code provided rules for the construction ofpressure vessels, pumps, valves, piping systems, and metal con-tainment vessels. The winter 1971 addenda added rules for atmos-pheric storage tanks based on the American Petroleum Institute(API) Standard 650, and 015 psig storage tanks based on APIStandard 620.

By 1973, 10CFR50.55a, paragraph (c), specified that for con-struction permits issued before January 1, 1971, for reactors notlicensed for operation, pressure vessels which are part of thereactor-coolant pressure boundary shall meet the requirementsfor Class A vessels set forth in Section III . . . applicable on thedate of order of the vessel. Paragraph (d)(i) equally permittedUSAS B31.1.0 or USAS B31.7 editions, addenda, and applicableCode Cases in effect on the date of order of the piping; paragraph(d)(ii) specified B31 Code Cases N7, N9, and N10; and para-graph (f) permitted valves to meet the requirements of B31.1.0 orClass I rules of the draft ASME Code for Pumps and Valves forNuclear Power.

Federal law limited the early codes of the 1960s to plants withconstruction permits that had been issued before January 1, 1971.The permitted codes were as follows: For piping, the rules ofASA B31 and USAS B31.1.0 were permitted, of which the earli-est code of record was the one in effect 6 months before theissuance of the construction permit. For pumps and valves, therules of ASA B31 and USAS B31.1.0 could be used, as couldthose of the draft ASME Code for Pumps and Valves for NuclearPower and addenda in effect on the date of order; the earliest per-mitted code of record was the one in effect 12 months before theissuance of the construction permit. For vessels that were part ofthe reactor-coolant pressure boundary, the requirements thatapplied were those for Class A vessels set forth in the Section IIIand applicable Code Cases and addenda in effect on the date oforder of the vessels. The earliest permitted code of record was theone in effect 18 months before the issuance of the constructionpermit. No guidance was given for Classes 2 and 3 constructionuntil Safety Guide 26 (later Regulatory Guide 1.26) was pub-lished with such rules.

For construction permits issued on or after January 1, 1971, butbefore July 1, 1974, 10CFR50.55a specified that the foregoingrules still applied; however, reference was made to ASME SectionIII because it now covered all the foregoing components. For con-struction permits issued on or after July 1, 1974, the requirementswere revised to specify that Codes applied to the componentsneeded to be of an edition no earlier than the 1971 edition,Section III, winter 1972 addenda. This revision required materialsfor Class 1 construction to meet the new fracture toughnessrequirements introduced in the summer 1972 addenda.

Since March 15, 1984, the federal rules require nuclear compo-nents to be certified and stamped in accordance with the ASMEB&PV Code, as exemplified by most utilities specifying ASMEcomponents for their plants while they were under construction,as well as for replacement equipment and parts for operatingplants. These requirements are described in paragraphs (c), (d),and (e) of 10CFR50.55a.

In 1973, the B31.7 committee issued B31 Case 115, which stip-ulated that piping designed and constructed in accordance withASME Section III of the B&PV Code, including addenda andapplicable Code Cases, may be accepted as complying with therequirements of USAS B31.71969 and applicable addenda forthe particular class of construction. ASME Section III, winter

1973 addenda, added rules for component supports (SubsectionNF), for core-support structures (Subsection NG), and for materi-als (Subarticle NA-3700).

To make the requirements for the different classes of compo-nents more easily understood, Section III, Division 1, was splitinto seven separate volumes in the 1974 edition. Each volume wascalled a subsection, and Division 1 included all the rules for themetal components of the nuclear system. Section III, Division 2,Code for Concrete Reactor Vessels and Containments, was pub-lished in 1975 as a separate volume. It contained the rules of con-struction for concrete reactors and containment vessels as well asthe requirements for containment metallic liners; it referencedDivision 1 and its appendices as required. Note that Division 2includes its own appendices, so one must be careful in using thisCode. A referenced Division 1 appendix will state that it is part ofDivision 1; otherwise, a referenced appendix will be to theDivision 2 volume.

In the 1977 edition, the General Requirements for Division 2(identified as CA) were put into the same volume as the GeneralRequirements for Division 1 (identified as NA). In the summer1977 addenda, the two sets of General Requirements were com-bined into one set identified as NCA, and NA-3700 was renum-bered as NCA-3800. The purpose of this change was to consolidatethe Code and, more important, to show the close relationshipbetween various parties in the construction of Code components fora nuclear plant. Every plant with concrete reactor vessels or con-crete containments has other components that are made of steel. Infact, all Division 2 concrete reactors and containments have steelparts that must meet some of the requirements of Division 1.

By 1976, the Code Cases that were referenced as applicable toSections III or XI numbered well over 175. The Nuclear CodeCases were removed from the B&PV Code Cases, and both werepublished in separate volumes in 1977. The Nuclear Code Caseswere listed with both the 4-digit number from the 1974 edition (inparentheses), as well as the new N-XXX number, but only duringthe three years of the 1977 edition. Recently, the ASME B&PVMain Committee ordered that cases be incorporated into the Codeor otherwise annulled or permitted to expire.

In 1977, the Committee authorized the publication of Inter-pretations in separate volumes. These books were published everysix months and included all of the Interpretations issued for eachsection of the Code. Initially, they were sold separately from theCode (eleven volumes were issued altogether), but sales were dis-appointing. To ensure that all users of the Code had access tothese Interpretations, they were published with each addendabeginning with volume 12 in the summer 1983 addenda. The1986 edition was the first in which the entire Code had addendapublished only once every year. It was decided to continue pub-lishing the Interpretations every six months. (Although manyorganizations involved in Code activities consult theInterpretations to resolve problems, they cannot be used to changethe Code: inquiries to do so generally result in a Code revision ora new Code Case.) Volume 57 was published in December 2004;it includes three indexes: a subject index for the new volumeproper; a key word-type subject index for volumes 12 through 57;and a cumulative numerical index. The Interpretations of Section III, Divisions 1 and 2, are included with the updateservice to Subsection NCA.

Beginning with the 2004 Edition, Interpretations of the Code will be distributed annually in July with the issuance of the edition and subsequent addenda. Interpretations previ-ously distributed in January will be posted in January at



www.cstools.asme.org/interpretations and included in the Julydistribution. The next Volume will be number 58.

Appendices to Section III were included from its inception asa Vessel Code. In the 1971 edition, all sections and appendicesof the Nuclear Code were included in one volume, and in the1974 edition, Subsection NA and Appendices were printed inthe same book. Because of their significant size increase, theAppendices were published in the 1977 edition in a volume sep-arate from NCA.

Appendices are classified as either Mandatory or Non-Mandatory. Mandatory Appendices are titled by Roman numeralsand include topics such as allowable design stress intensities,allowable stresses, material properties, and design charts forexternal pressure (these were moved in the 1992 edition toSection II, Part D), as well as stress analysis techniques; designrules for bolted flange connections and flat heads with large open-ings; and other Code design and administrative rules. Non-Mandatory Appendices are titled by capital letters and includestress analysis and dynamic analysis methods; rules for suchspecific design details as flat-face flanges, bolt cross-sectionalarea, and clamp connections); guidelines for the preparation ofDesign Specifications and Reports; lowest service temperaturedetermination and protection against nonductile failure; evalua-tion of service loadings in the faulted condition; and guidelines tofabrication controls. When they are specified in the DesignSpecifications, Non-Mandatory Appendices become mandatory.

33.9 THE REGISTERED PROFESSIONALENGINEER

The Registered Professional Engineer (RPE) who certifies aDesign Report as accurate and complete is responsible for recon-ciling the design drawings and any changes thereto with his or herDesign Report. The Design Report must be provided to theAuthorized Nuclear Inspector (ANI), although the Inspector is notresponsible for its accuracy or even for reviewing it. The RPEwho certifies an Owners Design Specification is probably theperson best qualified to perform the required Owners Review ofthe Design Report (NCA-3260); however, he or she is notrequired to certify the performance of the review. Revisions toDesign Specifications and Reports must be recertified. The RPEcertifying either document does not need to be from a differentdesign organization, but he or she must be independent from thespecific activity.

Initially, the Code did not address how competency in a field ofdesign could be verified. The Board on Nuclear Codes andStandards organized the Main Committee on Qualifications andDuties for Authorized Nuclear Inspection and SpecializedProfessional Engineers (N626). The first standard coveringqualification of RPEs was issued in 1978; in the 1980 edition,summer 1980 addenda, Section III adopted N626.31979, requir-ing review by Owners or N-Certificate Holders of the credentialsof RPEs as well as the maintenance of his or her expertise onceevery three years. Survey teams expect to see an RPEs self-evaluation of qualification at least once annually.

Section III, 1995 edition with A95 addenda, refers to N626.31993. It provides guidance to the RPE by listing the paragraphnumbers of each subsection for which he or she should haveworking, hands-on knowledge and those for which he or sheshould have general knowledge. This Standard became AppendixXXIII in the 1996 addenda and has since been annulled.

33.10 AUTHORIZED INSPECTION The ASME system uses an independent third party to inspect

work performed to ascertain that nuclear parts, appurtenances,and components meet the requirements of the ASME B&PVCode. This party is represented by the Authorized NuclearInspector (ANI), who is the key to success of the ASME Code.He or she is truly independent because he or she cannot be anemployee of the two parties most interested in the construction ofthe component: the Owner and the Certificate Holder. He or sheworks for an Authorized Inspection Agency (AIA) designated byor acceptable to the appropriate Enforcement Authority.

The ANI has experience and background in the inspection ofnuclear components and, in addition, is knowledgeable of boththe Nuclear Code and Quality Assurance. The duties and respon-sibilities of the ANI are detailed in NCA-5220. The qualificationrequirements for the ANI and the Authorized Nuclear InspectorSupervisor (ANIS) were originally listed in ANSI N626.01974for Division 1 and ANSI N626.21976 for Division 2. The threeStandards, including N626.11975 for Inservice Inspection, werecombined in 1985 into one Standard: N626 Qualifications andDuties for Authorized Nuclear Inspection Agencies andPersonnel. The ANI and the ANIS are tested and commissionedby the National Board and must be qualified according to ASMEN6261990 and addenda N626a1991. For Division 2, the ANIand ANIS must meet the requirements of Part N626.2 and haveexperience in the activities required for the placement of structur-al concrete. The AIA is required to meet ANSI N6261990 andaddenda N626a1991, Parts N626.0 and N626.2, of that Standard.

The ANI monitors the Quality Assurance Program and verifiescompliance with the Code by the Certificate Holder. He or sheverifies that the Certificate Holder has the necessary and up-to-dateCodes and addenda, that the Design Specifications and the DesignReport are available and properly certified, and that the OwnersReview of the Design Report has been received by the manufac-turer. He or she verifies that the materials used com-ply with theCode requirements, that proper welding procedures are used, andthat welders are properly qualified. Another requirement is toensure that Non-Destructive Examination (NDE) procedures areacceptable and that the NDE personnel are qualified. He or sheverifies that design calculations have been prepared when DesignReports are not required and, in addition, verifies and certifies that(to the best of his or her knowledge) the component or part is infull compliance with the Code. The integrity of the ASME pro-gram using Authorized Inspectors has been demonstrated for over85 years.

If any questions regarding Code compliance are encountered,they can be answered by the ANI. A Code decision by an ANImust be accepted; otherwise, the same question must be asked ofhis or her supervisor (the ANIS) or of his or her employer (theAIA). The question may also be posed to the ASME; it should fol-low the provisions of Appendix XX. Interpretations cannot beobtained from individual committee members or from any otherparty; only the B&PV Committee can issue official Interpretations.

Before the initial application, and during the three yearsbetween ASME surveys, the ANIS is responsible to review andaccept all changes to the Quality Assurance Manual. No changesto the program are to be implemented until acceptance by theANIS. For certificates having a scope that includes the manufac-ture and supply of material, the ANIS is required to audit this por-tion of the program annually. Also, he or she is required to per-form semiannual audits of both the ANI performance by ASME


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N626 and the status of the Certificate Holders Quality AssuranceProgram. The A92 addenda approved the N626a1991 addenda toASME N6261990. This revision introduced N626.4, requiringnuclear AIAs to be accredited by the ASME (NCA-5121); thisrequirement became mandatory on July 1, 1993, and nuclearAIAs are currently accredited.

This Standard was replaced by QAI-11995, Qualifications forAuthorized Inspection. This document introduced Part 5, whichrequires accreditation of AIAs who provide boiler and pressurevessel insurance and inspection services. An addenda was issued inApril 30, 1996, QAI-1-1a1996. It was followed by another adden-da, QAI-1-1b1999, issued on April 10, 2000. The N626Committee was also renamed the QAI Main Committee, whichnow serves under the supervisory Board on Conformity Assessment.The latest Edition was published in April 15, 2003, as QAI-12003.

Three Code Cases were issued in the 1995 edition, and onewith the 2003 edition. Interpretations are included with the edi-tion and addenda. With the publication of the 2003 Edition therewill no longer be addenda issued, only Editions.

33.11 QUALITY ASSURANCE The rules of Section III are very comprehensive regarding the

duties and responsibilities of all parties involved in the construc-tion of each component. Once the component has been classified,it is the duty of all parties involved to ensure that all the Coderules for design and construction have been met.

These parties are the following:

The Owner (generally the licensing utility). The N-Type Certificate Holders: N, NV, NA, NPT and NS. The Quality System Certificate (Materials) Organizations: MOs. The AIA and its employees: the ANSI and ANSIs. The Enforcement Authority: municipality, U.S. State, or

Canadian Province. The Regulatory Authority: NRC.

The Quality Assurance (QA) Program is an integral part of theNuclear Code. All parties are required to have such a system, forit is used for verification of Code compliance. In addition, it isextremely helpful in the manufacturing process because it consti-tutes a detailed program for doing work properly. Thus a goodQuality Assurance Program saves both time and money; it enablesthe Certificate Holder to do the work correctly.

Section III introduced Mandatory Appendix IX, QualityControl and NDE Methods, in the winter 1967 addenda. On April17, 1969, the AEC published a proposed amendment to 10CFR50that would add an Appendix B, Quality Assurance Requirementsfor Nuclear Power Plants. Appendix B was issued officially onJune 27, 1970.

In May 1969, the N45 Committee, Reactor Plants and TheirMaintenance, established an ad hoc Committee on QualityAssurance Program Requirements, composed of AEC representa-tives and key segments of the nuclear industry. In 1970, a newN45-2 Subcommittee on Nuclear Quality Assurance Standardswas formed to provide for the preparation, coordination, andapproval of the N45.2 series Standards, which resulted in the ini-tial issue of ANSI N45.21971.

In the 1971 edition of ASME Section III, the Quality AssuranceProgram requirements were included in NA-4000. The N45.2Committee indicated in its Foreword that these requirements were

consistent with the requirements of their Standard. There were manyproblems in interpreting the coverage, for the criteria were not asnumerous as the 18 criteria of 10CFR50 Appendix B, QualityAssurance Criteria for Nuclear Power Plants and Fuel ReprocessingPlants. Finally, in the 1977 edition, the winter 1978 addenda revisedNCA-4000 to match the 18 criteria of 10CFR50 Appendix B.During that time, another Main Committee under the Board onNuclear Codes and Standardsthe Committee on Nuclear QualityAssurancecompleted review of the ANSI N45.2 series of docu-ments and issued the first edition of ANSI/ASME NQA-11978,Quality Assurance Program Requirements for Nuclear Power Plants.

NQA-1 incorporated some of the ANSI/ASME N45.21977daughter Standards:

N45.2.6, Qualification of Inspection, Examination, andTesting Personnel for Nuclear Power Plants.

N45.2.9, Requirements for the Collection, Storage, andMaintenance of Quality Assurance Records for NuclearPower Plants.

N45.2.10, Quality Assurance Terms and Definitions. N45.2.11, Quality Assurance Requirements for the Design of

Nuclear Power Plants. N45.2.12, Requirements for Auditing of Quality Assurance

Programs for Nuclear Power Plants. N45.2.13, Quality Assurance Requirements for Control of

Procurement of Items and Services for Nuclear Power Plants. N45.2.23, Qualification of Quality Assurance Program Audit

Personnel for Nuclear Power Plants.

NQA-1 was reviewed by Section III for some time. After threeyears, the 1979 edition of NQA-1 was adopted in the winter 1982addendawith many exclusions, primarily Supplement 2S-2, andinstead retained NX-5500 for qualification of NDE personnel.Several editions of NQA-1 have been adopted by Section III.

The A91 addenda approved ASME NQA-11989 edition and1a1989 addenda, Quality Assurance Program Requirements forNuclear Facilities, and the A92 addenda added the 1b1991addenda, as modified and supplemented in NCA-4134. The 2006Addenda approved NQA-11994, which has also been acceptedby the NRC.

NQA-12008 has been reviewed by both ASME and NRC, andmay be approved in the near future, with some exclusions.

33.12 INSERVICE INSPECTION Lawrence Sage provided a history of the development of Section

XI in ref. [7]. He recounted that there were no inservice inspectionrequirements for nuclear equipment beyond that mandated by stateboiler and pressure vessel laws and insurance requirements. In1967, the ANSI and ASME established the American StandardsCommittee N45, Reactor Plants and Their Maintenance, to developthese rules. In 1969, the AEC announced its intention to requireinservice inspection at all new nuclear power plants.

The results of this work were first published in 1970 as ASMEB&PV Code Section XI, Inservice Inspection of Nuclear ReactorCooling Systems. Section XI includes rules for inservice inspec-tion, examinations, repairs, alterations, and replacements for thosenuclear components until the nuclear power system is taken out ofservice. It began as the nuclear equivalent of Sections VI and VIIand the National Board Inspection Code for boilers and pressurevessels.



In June 1971, the AEC made the requirements of Section XImandatory for all nuclear power plants with construction permitsissued on or after January 1, 1971. In February 1976, the NRCmade Section XI mandatory for all nuclear power plants. Sagesarticle provides a good summary of the changes that haveoccurred in this Code from 1970 to 1992.

Section XI now provides inspection rules for three differentnuclear cooling systems in three divisions. Rules for inservicetesting of pumps and valves have been moved to the Operationsand Maintenance (O&M) Standards; they are no longer containedin Section XI. In service testing of snubbers has also been moved.

The rules of Section III are applicable during the constructionof components by their Manufacturers, including their installationat the nuclear plant site. Section XI rules are applicable after thecompletion and stamping of the component or after its installationat the nuclear plant site. The two Codes operate in parallel duringthe later years of the construction cycle (as partial and completedsystems are turned over to the start-up organizations at the con-struction site for calibration of instruments), to provide water-coolant flow for pumps and rotating machinery, and also to famil-iarize the operating personnel with the equipment.

There are no accreditation requirements for organizationsworking under the provisions of Section XI. However, when theISI work plan calls for replacement of components with CodeSymbolStamped components or the addition of a complete newsystem, the rules for Section III and its accreditation requirementsapply. The work plan can specify the original Construction Codeor any other Code of record.

33.13 CONSTRUCTION CODESECTION XIINTERFACES

Owen F. Hedden described ASME Section XI as providing therules for examinations, test, analyses, and repairs to ensure thatstructural integrity of the primary coolant pressure boundary ismaintained [8]. Whereas all other sections of the Code are direct-ed toward the activities of equipment manufacturers, Section XI isthe only mandatory section of the Code that is directed toward thepostconstruction activities of the power plant Owner/Operator.

Deardorff et al. state the following about Section XI [9]:

[It] contains rules and requirements for inservice inspection,testing, evaluation, and repairs to operating nuclear plants.Section XI defines the inspection interval and inspection loca-tions such that fatigue cracking (or other material degradationmechanisms) would be detected in a timely manner. If crackingis detected, then evaluation criteria are provided for determiningif continued operation is acceptable, or if repairs/replacementsare required.

33.14 REPAIRS AND REPLACEMENTS Metrow, in ref. [10], describes the development of the

Enforcement Authoritys involvement in Section XI activities,including review of repair plans. The 1970 edition containedrequirements for a repair program, but details on content weremissing, and the summer 1973 addenda introduced repair pro-grams for reactor vessels only. Section XI combined replacementswith repairs until well into the 1974 edition. The summer 1976addenda separated replacements into a separate article: IWA-7000.

The winter 1985 addenda established the requirement for a doc-umented replacement program. Repair programs were described inthe various articles, but it was not until the 1989 addenda thatdetails of the content, scope, and other aspects intended for inclu-sion into the repair program were written into one central location.

The A91 addenda consolidated the replacement requirementsfrom IWA-7000 into IWA-4000. The rules for repair and replace-ment are located in Section XI, IWA-4000, 1992 edition. This wasdone as a result of a feasibility study begun in 1986. Gimple pro-vided an excellent description of the current changes and plansfor future revisions in ref. [11].

Significant changes were introduced in the 1992 addenda thatprovided alternative requirements in paragraph IWA-4122 forNPS 1 Class 1 piping, tubing (except heat-exchanger tubing,sleeves, and welded plugs), valves, fittings, and associated sup-ports and other criteria described in IWA-4121.

The alternative requirements include exclusions from NCA-3800, Certificate of Authorizations (and, obviously, Code SymbolStamping), and agreements with AIAs. The Owners QA Programprovides measures for assurance that material is furnished inaccordance with the Material Specifications and applicablerequirements of Section III.

Pressure testing, AIA participation, and completion of the NIS-2Data Report Forms are not required for the installation of theseitems.

33.15 BACKFITTING OF NEW CODEREQUIREMENTS ON OPERATINGPLANTS

As a result of the lessons learned from the Three Mile Island(Middletown, Pennsylvania) accident, many items have beenadded to operating plants to prevent recurrence of similar events.Unless specified by the NRC, retrofitting is not a code require-ment. In fact, many codes and standards clearly state that thisstandard is not retroactive. It also should be understood that aplant license is pegged to an effective Code, and changes intro-duced in later Codes are not required by the Construction Code orthe plant operating license.

As described in the preceding paragraph, the NRC permits theuse of older Codes and Standards. Even today, the RegulatoryAuthority accepts repair and replacement plans that refer to theoriginal Construction Code. It is therefore imperative that theCode specified for the repairs, replacements, and mandatory back-fits be carefully reviewed for acceptability and for impact on thefabrication and installation schedules.

33.16 CAREFUL PREPARATION OF WORKPLANS

It must by now be self-evident that knowledge of the operatingplants effective Codes is of paramount importance in properlyselecting original Codes for repair or replacements. Availabilityof replacement items (material, parts, or components) depends onthe continued presence of the original manufacturer or fabricator.

The brief summations of original Codes in this chapter do notprovide enough information on specific requirements that wouldpermit the use of reduced requirements on organization certifica-tion, material quality assurance, or procedure qualification. Doingso can be achieved only through careful review of the original


590 Chapter 33

Construction Codes or by general knowledge of the requirements ofcurrent Codes to permit use of later editions and addenda.

The two examples shown in Sections 33.18 and 33.19 involve aplant representative for most of the operating plants with con-struction permits awarded before January 1, 1971. The require-ments of the Construction Code were used to justify questionsand findings raised by Inspectors during review of the workprogress against the work plans.

33.17 CODE RECONCILIATION In maintaining the operating status of a power plant, it is often

necessary to procure material, parts, and components as part ofplant repairs and replacements. If like-for-like replacement mater-ial, parts, and components can be procured, reconciliation maynot be required. However, because nuclear power plants encom-pass a period of over 30 years, it may no longer be possible tofind like-for-like replacements. Both Sections III and XI makeprovisions for the use of later editions and addenda of the Codesbecause the Codes are revised every year (before the 1986 editionfor Section III and the 1983 edition for Section XI, they wererevised twice each year). Owners are often forced to buy replace-ment items to the Code currently in effect.

Section III, Divisions 1 and 2, Subsubarticle NCA-1140, per-mits the use of specific provisions of later editions and addenda ifall related requirements are met. It also permits the use of existingmaterial procured before the Code effective date for the compo-nent with certain provisos. These provisos are: the mutual consentof the Owner or Owners designee and the N-Certificate Holder,and acceptance of their use must be made by the Enforcement andRegulatory Authorities having jurisdiction at the site. NCA-3554also requires that the Owner reconcile the Design Report whenany modification of a document affects the design.

Since the summer 1976 addenda, Section XI, IWA-7210(c),requires reconciliation by the Owner of the use of any provisionsof later editions or addenda not specified in the DesignSpecifications for design, fabrication, and examination of areplacement. Mechanical interfaces, fits, and tolerances that pro-vide satisfactory performance are not to be changed; moreover,new materials must be compatible with the installation and sys-tem requirements. These requirements are now in IWA-4170beginning with the 1992 edition. Future revisions to this subsub-article will clarify the requirements for reconciliation.

Reconciliation has frequently consisted of a detailed and onerouscomparison of every revision, however minor, that has occurredbetween the Code of Record and the Code being reconciled. Eachutility is compelled to repeat the reconciliation process for eachitem. Commercial programs are available to simplify this process.

To the best of this authors knowledge, no U.S. state or anyCanadian province has yet objected to the use of any Code editionor addenda. Except for the time in which the NRC did not acceptthe ASME Section XI flaw size, all editions and addenda throughthe 1998 edition and including the 1998, 1999 and 2000 addendahave been accepted by the NRC; in fact, the 2001, 2002 and 2003Addenda edition were incorporated by reference in a final amend-ment to 10 CFR 50.55a which was published on October 1, 2004(69 FR 58804). This rule became effective on November 1, 2004.NRC staff is completing the technical bases for the amendment to10 CFR 50.55a to endorse the 2004 Edition. The proposed rule isscheduled to be published for public comment in the last quarterof 2005.

33.18 EXAMPLE A: REPAIR OF STEAM-GENERATOR FEEDWATER-NOZZLECRACKS AT A NUCLEAR PLANT INTHE 1970s1

33.18.1 Statement of the Problem Feedwater nozzles in Westinghouse steam generators were

examined under the Southeastern Electric Cooperative (SEC)Section XI program and were determined to have sustained severecracking that required weld repairs. An IWA4000 program waswritten and submitted to the Authorized Nuclear InserviceInspector (ANII) for approval; preparations for weld repairs werethen initiated. A foreign repair organization, not holding anyASME accreditation, was engaged to do the repairs. The weld fillermetal was not procured to the requirements of NCA-3800, nor hadit been qualified in accordance to Section IX. The ANII, reviewingthe progress of the repairs, took the position that the work packagedid not meet SEC commitments and that the welding did not meetCode. He indicated that a nonconformance should be identified andresolved before he would sign off on the NIS-2 Data Report Form.

33.18.2 Background Example Nuclear Plant (ENP) is a Codes and Standardstransition

plant because of its licensing chronology. Criteria for materialprocurement at ENP are dependent on whether the material isintended for components procured by the SEC or intended forcomponents supplied by the Nuclear Steam Supply System(NSSS) vendor under his scope. This position paper will establishthe basis for material procurement at ENP for NSSS-suppliedcomponents by using the Westinghouse steam generators to repre-sent typical ASME Section III components. The bid specificationfor ENP was completed in late 1967; the NSSS contract wasawarded to Westinghouse on April 18, 1968. Most of the nuclearequipment procured by Westinghouse had a Code of Record ofASME Section III Nuclear Vessels, 1968 edition.

The construction permit for ENP was granted on October 1,1969. At that time of nuclear plant construction, the AEC wasforced to accept a commercial Code such as USAS B31.1.01967as the Construction Code for power piping. In addition, it accept-ed B31 Code Cases specifically intended for nuclear applications.The various Code requirements were specified in the FederalRegulations.

The component contracts that Westinghouse awarded resultedin the ordering of items for Westinghouses scope of supply inmid-1968. This established the 1968 edition, ASME Section III,Nuclear Vessels as the Code of record for the reactor pressurevessels, pressurizers, and steam generators. The certification sheetof the steam generator stress report states the 1968 edition ofSection III, which verifies the foregoing assumption.

As for piping, pumps, and valves, the SEC continued its estab-lished practice of placing a contract with a piping fabricator. Itexpanded this contract to cover the procurement of pumps, valves,component and piping supports, and loose material for field fabri-cation.

The principal piping systems and appurtenances contract atENP was awarded to National Valve & Manufacturing Company(NAVCO) on August 26, 1970 (SEC Reference No. 71C-37-92615).The USAS B31.7 Code had been approved on August 24, 1969.For main steam piping, the SEC selected B31.1.01967 and

1In these examples, both the name of the utility and the affected nuclear powerplant have been disguised, as well as the names of personnel listed.



labeled it as non-QA but specified B31.7 for fabrication and test-ing for all other piping.

The rest of this discussion will consider the Code of Record forthe steam generator as ASME Section III, 1968 edition. For pip-ing, it will consider USAS B31.71969 without addenda becauseall B31.7 addenda were issued after August 26, 1970.

33.18.3.1 Steam Generators: Code Requirements ParagraphN-141, Design Specification only covered functions, loadings,environmental conditions, and classification; no material require-ments were addressed. Paragraph N-142, Stress Report also failedto address materials. The Authorized Inspector duties listed inparagraph N-143 included ascertaining that the vessel or part wasconstructed in accordance with approved drawings and theDesign Specification, using materials complying with the relevantMaterial Specifications. In paragraph N-144, the manufacturersresponsibility included that related to the identification of correctmaterials and valid material certification.

Paragraph N-151 established that the jurisdiction of Section IIIterminated at the limit of reinforcement given by N-454(b) but notcloser to the main shell than the first circumferential joint, exclu-sive of the connecting weld in welded connections. A furtherrequirement stated that where connected piping differs from thevessel in nominal thickness or coefficient of thermal expansion,the joint was to be analyzed as a vessel joint.

The connecting weld at the feedwater nozzle bridges a verylarge difference in nominal thickness between the nozzle and thepiping because of the difference in material specified tensile andyield strengths. The nozzle is made of quenched and temperedSA-508 Class 2 material, with an allowable design stress intensityof 26.7 ksi at the design temperature of 600F. The pipe materialis made of SA-333 Grade 6, ASTM A 333 Grade 6, or ASTM A106 Grade B material, all of which have an allowable designstress intensity of 17.3 ksi. The inverse ratio of stresses results ina calculated pipe thickness 1.5 times the nozzle thickness at thedesign pressure of 1,085 psig.

The joint needed a transition dutchman or a special com-pound-bore nozzle, wherein the vesselside ID narrowed to thepipe ID using a tapered transition to reduce discontinuity stresses.The SEC chose to transition using an intermediate piece of SA-508 Class 2 material that was shop-welded to the first pipe spoolby means of a qualified P3-to-P1 weld procedure specification.The field weld then became an essentially uniform wall-thicknessP3-to-P3 weld. If the field weld was within the reinforcementlimit, then the weld had to meet Section III requirements. If it waslocated outside of the reinforcement limits, it was a piping weldand had to meet either B31.7 or B31.1.0 requirementsbased onthe Code Class listed in the NAVCO contract.

The material requirements in ASME Section III, 1968 edition,are listed in Article 3, Materials. Paragraph N-310 required thatpressure-boundary material or material welded thereto usedunder the rules of Subsection A shall conform to the requirementsof one of the specifications for materials in Tables N-421, N-422,and N-423. . . . All special requirements of Article 3 applicableto the product form (plate, forging, tube, pipe, etc.) had to be met.

The Material Manufacturer was required to satisfy all therequirements of the Material Specifications and those of Article 3in addition to certifying the completion of all activities that wereperformed. The Manufacturer had to include certified reports ofany required tests, inspections, and repairs made on the materials(paragraph N-312). Paragraph N-511 required that the vessel man-ufacturer certify compliance with the special requirements of

Article 3 for any activity that was performed, and to certify reportsof all tests and examinations that were made on the material.

The requirements for welding material were described in para-graph N-511.3, which specified that tests shall be conducted foreach lot of covered or flux-cored electrodes, for each heat of bareelectrode and for each combination of heat of bare electrode andbatch of flux mix to be used for vessel welding. The tests wererequired to be made in accordance with paragraph N-511.4 or N-511.5; the results had to conform to the minimum requirementsof those paragraphs. Paragraph N-512 (Material Identification)required that pressure-part material carry identification markingsuntil the vessel was completed; therefore, an as-built sketch or atabulation of materials had to be prepared to identify each pieceof material with the mill test report (MTR) or coded marking (if such was used).

Although Section III, paragraph N-512, did not specificallyaddress weld metal identification, paragraph N-523(b) specifiedthat the manufacturer is responsible for control of the weldingelectrodes . . . suitable identification, storage, and handling ofelectrodes, flux, and other welding materials shall be maintained.It was not until the summer 1970 addenda that any quality pro-grams for materials were specified. The requirements of the newparagraph N-335 were later incorporated in NX-2600 in the 1971edition, where they remained until NA-3700 was introduced inthe winter 1973 addenda.

To summarize, it can be stated that for Section III (NuclearVessels), Material Manufacturers did not have to be accredited,material did not have to be manufactured to a written QA program,and the only requirements were provision of the MTR (certifyingthat the material complied with the requirements of the MaterialSpecifications) and also certification by the material and/or vesselmanufacturer that all the special requirements of Article 3 for pres-sure-boundary material and Article 5 for welding material had beenmet. Another provision was providing identification of base metaland welding material traceable to the MTR or other test results.

33.18.3.2 Piping Systems: Code Requirements If the firstcircumferential weld was a piping weld, the applicable Code wasUSAS B31.71969. The requirements for piping were very simi-lar to those of Section III, 1968 edition, for vessels. Paragraph 700required a Design Specification; 700(d) required the manufacturer,fabricator, and erector to provide materials complying with therequirements of this Code and Design Specification. Paragraph700(e) required MTRs to be retained by the Owner for the life ofthe plant, and paragraph 700.1.4 included the first circumferentialweld joint external to the vessel under B31.7s jurisdiction.

Chapter 1-III, B31.71969, listed all the requirements for materi-als for this Code. Paragraph 1-723.1.1 stated that material shallconform to the requirements of Table 1-724 and Division 1-724 thatapply to the material product form. Paragraph 1-723.1.2 required acertification from the material manufacturer that all the requirementsof the Material Specifications were complied with and all specialrequirements of the chapter fulfilled. Certifying that the material metthese requirements had to be done, and a certified report of theresults of all required tests, examinations, and repairs performed onthe materials and their identification had to be included.

Paragraph 1-723.1.3 required each piece of pipe, each fitting,and any component part to be clearly identified, as described in(a), (b), or (c) of the paragraph. Paragraph 1-725.5 described theidentification requirements for welding material.

The only references to brittle fracture in B31.71969 were foundin paragraph 1-723.2.3, which warned about low temperatures,


592 Chapter 33

and in paragraph 1-737.3(g), where the standard indicated a warn-ing that hydrostatic pressure tests should be conducted at a fluidtemperature of 60F and higher for material whose resistance tobrittle fracture at low temperature had not been enhanced. Thisrequirement was similar to what was traditionally listed in SectionI, Power Boilers, and Section VIII, Unfired Pressure Vessels70Fand 60F, respectivelyfor minimum hydrostatic test temperature.

The fabricator and erector were required to certify that thematerials used complied with all the requirements of Chapter 1-III, as shown in paragraph 1-727.2. Paragraph 1-727.2 permittedfiller metal not incorporated in Section IX to be used if a proce-dure qualification test was first successfully made in accordancewith Section IX.

It was not until the ANSI B31.7b1971 addenda, dated April 1,1971, that Appendix I was introduced. Paragraph I-104.4described the requirements for Cv-Notch impact testing of welds.It was also in the B31.7b1971 Addenda that a paragraph wasadded: 1-727.5.7, which required additional welding qualificationsfor weld procedure qualification tests for materials with impacttest requirements.

USAS B31.71969 did not have any requirements forqualification of Material Manufacturers, nor did it have any forwritten quality control programs. To summarize, the materialrequirements for USAS B31.71969 were less stringent thanthose for Section III, Nuclear Vessels, 1968 edition. Theyrequired certified MTRs to certify compliance with the require-ments of the Material Specifications, and certification by the fab-ricator or erector that the materials comply with all requirementsof Chapter 1-III and of Section IX.

33.18.4 Conclusions The filler metal used in the connection welds to the steam-

generator feedwater nozzle is acceptable under the provisions ofSection XI, which permit the use of the original ConstructionCode for repairs.

The filler metal should be procured in identified containers orspools; the weld deposit should be tested as required by Section III,1968 edition, and/or by Section IX; and a welding procedurequalification test should be performed to show successful compli-ance with Code requirements.

The filler Material Manufacturers MTR, the certification by therepair organization of the results of the chemical and mechanicaltests of the deposited weld metal as meeting the requirements ofthe Code, and the records of the weld procedure qualification testand welder performance tests for the welders making the feedwater-nozzle welds, should all be presented to the ANII for reviewbefore the NIS-2 Data Report Form is accepted and certified.

If the steam generator is not to be subjected to another full-pres-sure test, the hydrostatic pressure test requirements of Section XImay be used in lieu of the 1.25 times design pressure of paragraph1-737.4. Alternatively, the helium mass spectrometer test or halideleak test of paragraph 1-737.1(b) (more fully described in para-graph 1-737.1.3) may be used if it is acceptable to the ANII. Itwould then be followed by a Section XI system leak test.

33.19 EXAMPLE B: ENP-SUPPORTMATERIAL REQUIREMENTS

33.19.1 Statement of the Problem In reviewing material documentation at the ENP, the ANII

identified a problem with piping supports material documentation.

His interpretation of the SECs material procurement commit-ments was that material for load-bearing supports should be pro-cured with at least the Material Manufacturers ASME Section IIICertificate of Compliance (COC). With the plant attempting toreturn on line after an extended outage, any delay in establishing amaterial verification program would have resulted in an unaccept-able time delay in the return schedule. This investigation is intendedto identify the Code requirements and the applicable SEC commit-ments, as well as to justify the use of commercial grade material ifit is permitted for pipe supports.

33.19.2 Background Before addressing the resolution of the problems presented by

various ENP Problem Evaluation Reports (PERs) on componentsupport materials, it is important to establish the requirements ofthe Codes and Standards involved, as well as list the commit-ments made by the SECs Division of Engineering Design (DED)during the early years of the ENPs construction.

The Codes and Standards first employed in the design and con-struction of the ENP were ASME Section III, Nuclear Vessels,1968 edition, for vessels and pumps, and also USAS B31.1.01967(B31.1.0) for piping and valves. With the publication of USASB31.71969, Nuclear Power Piping, the ENP piping design con-tinued in accordance with B31.1.0. Fabrication and testing inaccordance with B31.7 were selected for systems associated withthe reactor-coolant pressure boundary. Supports for piping andcomponents were not covered in a separate standard as they aretoday in ASME Section III, Division 1, Subsection NF, ComponentSupports. Vessel supports were covered by the vessel DesignSpecification and were constructed with the same rules used forthe pressure-retaining component. Large pumps and valves wereindividually supported, using the design rules associated with ves-sel supports. Valves and pumps installed in the piping systemtransferred their weight to the piping and its supports; thus theywere included in the design of the piping.

Piping was described in paragraph 100.1.1 as also includinghangers, supports, and other equipment items considered neces-sary to prevent overstressing the pressure-containing parts. Pipe-supporting elements, such as hangers, supports, and structuralattachments, were defined in paragraph 100.2 in B31.1.0 and ingreater detail in paragraph 120.1. Design rules for pipe-support-ing elements were discussed in paragraph 121; this wording wasused for the rules of piping supports in B31.7 and was included inASME Section III, Subsection NF, 1971 edition, winter 1973addenda. The definition of piping moved to paragraph 700.2 inB31.7, which also included the definition for pipe-supporting ele-ments. These definitions are very important to the requirementsfor material certification and identification (which vary significantlyin these two Piping Codes).

In the early years of piping design, the loadings to be consid-ered did not vary significantly and were related to sustainedloads from pressure and temperature; from impact forces causedby external and internal causes, such as the effects of water andsteam hammer; from natural events such as wind, earthquakes,and vibration; and weight effects from piping, fittings, insulation,snow, ice, and the transported fluid. Unexpected conditionsoperator error, equipment malfunction, and limited variationsfrom normal operation (e.g., transients in pressure and tempera-ture)were evaluated and accounted for by allowances onstresses for various operating periods that experience indicatedwould not cause damage to the piping requiring repair orreplacement.



Because these loadings did not cause major changes of the mag-nitude of loads, items covered by MSS SP-58 (which were loadcapacityrated by the Standard) and mass-produced catalog items(which were load capacityrated by their manufacturers) wereemployed to design a load-carrying string or path. These standardsand catalogs specified what today are referred to as commercialgrade materials; they relied on the integrity of MaterialManufacturers and suppliers to guarantee that materials met theminimum properties specified in the Material Specifications.Reporting the results of specified tests to verify the compliance ofthe material with the Specifications was not required unless it wasrequested by the purchaser.

Most specifications did not require material identification withthe ladle-analysis of the heat from which the product was manu-factured. At best, some specifications required marking the materialwith the specification number and the material type or grade, orelse a specific symbol that accomplished the same result as todaysCOC. The material manufacturers COC was introduced in theANSI B31.7c1971 addenda, paragraph 1-723.1.2, and the ASMESection III, winter 1973 addenda, paragraph NA-3767.4(a).

As nuclear plant design became subjected to failure-preventionconsiderations, many postulated events of low and intermediateprobability were added to piping design by the AEC/NRC staffpositions (later to the Standard Review Plans). These eventsrequired design consideration of loadings of a magnitude neverbefore envisioned by piping analysts: single- and double-endedpipe-guillotine failures, jet-impingement loads, pipe-whip, cold-legbreaks, and so forth. The resulting loads were so large that theStandard supports of MSS SP-58 were no longer adequate foruse in piping supports, so new, engineered support assemblieswere designed using large, thick, structural shapes and, withincreasing frequency, cold-drawn ASTM A 500 Grade B materialand hot-rolled ASTM A 501 square and rectangular pipes. Materialrequirements changed from medium carbon steels to quenched andtempered low-alloy steels in the 110160 ksi specified minimumtensile strength levels. Eventually the ASME Code published CodeCase 1644, which was divided into two cases in the 1977 editionand reissued as Nuclear Code Cases N-71 (for welded construction)and N-249 (for nonwelded construction). These Code Cases pro-vided many more high-strength, quenched, and tempered materialsthan were listed in Tables I-11.1, I-12.1, I-13.1, and I-13.3, all ofwhich are now listed in Section II, Part D, Subpart 1, Tables 1A,1B, 2A, 2B, 3, and 4, 2004 edition.

Much of the change in design philosophy was introduced intoB31.7, which adopted the ASME maximum shear-stress designcriteria for Class A Nuclear Vessels and Section VIII, Division 2Nuclear Vessels (Class B) Pressure Vessels for Class I NuclearPiping. In addition, it adopted more restrictive requirements formaterials. In B31.7, paragraph 1-723.1.2, Certification of Materialsby Manufacturer and 1-723.1.3, Identification of Materials, nodistinction was made between pressure-boundary and structuralmaterial, which resulted in significant constraints on support man-ufacturers. The ANSI B31.7c1971 addenda solved the problem bydefining a new termpressure-retaining materialin paragraph1-723.1.1(b), as well as by addressing nonpressure-retainingmaterial in paragraph 1-723.1.1(c), which were both published in acompletely rewritten Chapter 1-III. Paragraph 1-723.1.2 definedthe COC as the Material Manufacturers certification that the mate-rial complies with the applicable material certification. It also statedthat COCs may be supplied in lieu of Certified Material TestReports (CMTRs) for pipes, tubes, and fittings of in. nominalsize and less, as well as for all material to be used as attachments,

34

hangers, supports, and fasteners. Paragraph 1-723.1.3 requiredidentification for pressure-retaining material, but did not addresscomponent standard supports or support material. Paragraph 1-723.1.4 provided specific requirements for materials of supportand hangers, and exempted other materials from CMTRs andCOCs. No such requirements were specified for materials used forpiping in USAS B31.1.0, thereby deferring material certificationand identification to whatever requirements were listed in theapplicable Material Specifications or standards.

33.19.3 SEC Commitments The first recorded SEC commitment for design and installation of

piping systems is in the ENP Preliminary Safety Analysis Report(PSAR), Table 3.2.2-2, and later revisions in the Final Safety AnalysisReport (FSAR). This table makes no reference to pipe-supportingelements; therefore, the requirements of the referenced Codes for thepiping systems become the initial SEC commitment for the design of,installation of, and materials used for piping supports.

The National Valve and Manufacturing Companys (NAVCO)principal piping contract incorporated the SEC Specification 9923for principal piping systems and appurtenances for the ENP.Paragraph 2(a) required NAVCO to furnish, fabricate, test, anddeliver hangers required for the work. Paragraphs 4(a) and 4(b)listed the piping systems that would be fabricated and tested toB31.1.0 and B31.7, respectively.

Paragraph 6(a) makes reference to the SEC Standard HangerDrawingsprefixed G-M-4 in Drawing Series 30W615-1 through -5(Pipe Supports and Anchors)whereas paragraph 6(d) describestheir purpose. Paragraph 7(c)(2)C addresses other materials andappears to indicate that heat traceability and MTRs are required;however, it does not clearly specify that the term other includesmaterials for component supports and component standard sup-ports. The General Notes section of Table 3.2.2-2 (Sheet 1, secondparagraph) does identify the reference to B31.7 as includingAddenda (a), (b), and (c), so therefore the certification require-ments of paragraph 1-723, Addenda (c) for hangers and supportspermit the use of COCs when certification is required.

Further clarification on SEC commitments was provided by theDED in a series of memorandums (or letters) issued in 1975 and1976. One memorandum from D. R. Punter to R. M. Kicker, datedSeptember 5, 1975, addressed certification requirements for integralattachments and component standard supports for ENP Classes A,B, C, and D piping. In this memorandum, the MechanicalEngineering Branch indicated that integral attachments to Class Apipe and Classes B, C, and D impact-tested pipe would requireCMTRs and heat-code traceability. On the other hand, for Classes B,C, and D pipe with no specified impact testing, only COCs would berequired. Component standard supports would require COCs toMSS SP-58 and the SEC specification. Items excluded by paragraphNF-2121 would not require any Material Manufacturers COCs.

A memorandum from R. G. Center to R. H. Passer, dated April20, 1976, established requirements for traceability of material andexamination requirements for later nuclear plants. The provisionsof the September 5, 1975, letter were repeated; for all other rawsupport material, traceability in accordance with the MaterialSpecifications was required from the project-segregated ware-house to the mill heat number, but traceability from the ware-house to the installed location was not required. The April 20,1976, letter also made provision for testing support material notmeeting these requirements in accordance with the materialsrelation to nuclear safety. The ENP was specifically excludedfrom the provisions of this memorandum.


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A later memorandum from R. G. Center to R. H. Passer, datedAugust 17, 1976, and of the same subject, permitted the use of theprovisions of the April 20, 1976, memorandum at ENP. Therefore,the requirements of the original Code of Record still apply.

Finally, the PERs mention load-bearing members. This termhas been used in documents of the SEC (and possibly that of otherlicensees) documents. The earliest reference available to the authoris a copy of the NRC Information Notice No. 88-95, InadequateProcurement Requirements Imposed by Licensees on Vendors. Areference to a manufacturers QA manual addresses the require-ments for load-bearing parts of a snubber. However, this termdoes not exist as a defined term in ASME Section III, SubsectionNF, nor is it by itself a basis for review or lack of compliance.

There are many load-bearing parts of component and pipingsupport load paths for which no code requirements have beenspecified. As described in the aforementioned NRC notice, many ofthese parts are listed in paragraph NF-2121 and are excluded fromthe requirements of Article NF-2000; as such, they do not requireCOCs pursuant to the provisions of NF-2130. These parts are andhave always been exempted from the requirements of ASME NCA-3800. The exemptions clearly indicate that these parts can be pro-cured as commercial grade items. Component and piping supportmaterials not exempted by NF-2121 may be exempted from NCA-3800, Metallic Material Manufacturers and Material SuppliersQuality System Program, as permitted by NF-2610 for small prod-ucts and materials that NF-2130 permits to be supplied with COCs.All these considerations are for information only, for ENP is notrequired to meet the requirements of Subsection NF.

33.19.4 Section XI Requirements The requirements of IWA-7220 must also be considered in the

replacement of material in existing supports or for installation ofnew supports. The reason for the installation becomes an impor-tant issue in this evaluation.

Although components, parts, and appurtenances are called itemsin ASME Sections III and XI, so are material and component sup-ports (IWA-9000). Therefore, the provision of IWA-7220 thatbefore authorizing the installation of an item to be used for replace-ment, the Owner shall conduct an evaluation to determine the suit-ability of that item. By addressing failure of an item that necessi-tates replacement, the cause of the failure must be considered.

When the replacement item is like-for-like and caused only byroutine maintenance (such as replacing a fastener during the dis-assembly and reassembly of a pipe clamp), the evaluation for suit-ability would only consist of Material Specifications year-datereconciliation.

When the need for like-for-like replacement is caused by erosion,corrosion, or fatigue failure, for example, replacement without elim-inating the cause of failure or making appropriate corrective provi-sions will only result in a need for future replacement. The correc-tive action must be consistent with the original Construction Code orthe Section III Code in effect at the time of specification revision.The evaluation report becomes part of the Form NIS-2 Data Report.

Supports for piping systems NPS 1 and smaller are exemptfrom the foregoing requirements.

33.19.5 Compliance with Original Construction CodeRequirements

The following position is based on the assumption that theSection XI replacement program work package specifies the orig-inal Construction Code and Code Cases as applicable.

(1) For ENP, the certification requirements for replacementmaterial for piping system supports that NAVCO was per-mitted to fabricate and test to B31.1.0 only need to meet therequirements of the items Material Specifications. Sincethe original Construction Code permitted ASTM material,commercial grade material is acceptable as-is if it is orderedto the Material Specifications year-date originally used.The only evaluation required is a reconciliation to ensurethat material manufactured and supplied to later editions ofCodes and Standards meets or exceeds the requirements ofthe original Material Specifications.

(2) The certification requirements for replacement material ofcomponent supports for piping systems that NAVCO wasrequired to fabricate and test to B31.7 are, as a minimum, aCOC with the Material Specifications. However, this is notthe same COC defined in NA-3767.4(a), for B31.7 onlyrequired a statement of compliance with the MaterialSpecifications. Because qualification of Material Suppliersand Manufacturers was not a B31.7 Code requirement,qualification and QA procedures for approved vendor listsonly apply if ENP committed to them in their QA Program.If the material is ordered to the same specifications used asthe original Material Specifications, the only evaluationrequired is a reconciliation to ensure that material manufac-tured to later Codes and Standards meets or exceeds therequirements of the base Material Specifications. This rec-onciliation can be accomplished by the provision of a ven-dor-supplied COC or by performing a commercial gradededication process and, if the material is determined to beacceptable, issuing a SEC COC. A proposed example ofsuch a SEC COC is shown in Fig. 33.1.

33.19.6 Commercial Grade Dedication (CGD) CGD evaluates the critical characteristics and specifies the

required inspection and acceptance criteria to ensure that items ded-icated after receipt are acceptable for use as replacement parts. Itprovides more assurance of the capabilities of the material thanwhat was required by the original Construction Code. For ENP, inwhich the material could have been originally procured as what istoday called commercial grade, in addition to being supplied toB31.7 with a simplistic COC for meeting the requirements ofMaterial Specifications, the testing and inspection of the CGDclearly can be considered equal to or better than the specified COC.

In any of the preceding cases, if replacement material is pro-cured to different Material Specifications than those of the origi-nal support material, a more thorough evaluation is required. Thisevaluation must prove the functional adequacy of the material,including equal or higher specified tensile requirements, ductility,weldability, and fracture toughness (if specified). To do this task,the SEC evaluation program must include material testing toestablish actual mechanical properties. For attachment material,chemical testing is required to establish the chemical compositionof the replacement material.

33.19.7 Recommended Disposition of PERs Procurement of new supports required for a modification will

necessitate design, materials, fabrication, and testing commensu-rate with the original Construction Code or a specified later Code,as permitted by ASME Section XI. Certification requirements formaterials shall be specified in accordance with the requirementsof that Code.



Procurement of replacement material for existing supportsshould, as a minimum, meet the requirements of the originalConstruction Code, the SEC commitments, and the Section XIrequirements as described previously.

33.20 MATERIALS PROCUREMENT Material can be procured to the current Code. Contrary to the

perception of many Quality Assurance organizations, it hasnever been the intent of the NRC to limit material certificationto only those specifications listed in Section II, Parts A, B, andC of the edition and addenda accepted in 10CFR50.55a. Thatparticular edition (currently the 2004 edition) includes NCA-1140(b) and (f). This inclusion permits the use of later editionsand addenda as well as earlier Code material. Reconciliation isthe key.

Earlier or existing material must be certified to the samespecification, grade, type, or class as the material it will replace.In addition, it must meet the minimum specified tensile and yieldstrengths of the material described in the componentsConstruction Code. Although ductility (evidenced by reduction ofarea and elongation properties) does not affect design, it doesaffect toughness, so it merits evaluation. Changes in chemicalcomposition should be reviewed and evaluated as well.

Material procured to certification year-dates later than thosespecified in the components Construction Code must also be rec-onciled. Technical changes made to subsequent revisions to thespecifications should be documented and evaluated.

On some occasions, Material Specifications are annulled; forexample, SA-155, welded-with-filler-metal carbon and alloy steelpipe, was replaced by SA-671, SA-672, and SA-691. Due care

must be exercised in procuring this material to ensure that the cor-rect grade and class have been specified.

Material initially used on the basis of a Code Case can bereplaced by material listed in a later edition or addenda by the useof the Code Case index and documentation of the date in whichthe Case was incorporated into the Code. Material whose specifiedstrengths have been changed in the Material Specifications can stillbe used. In the case of a reduction in strength, documentation ofthe mechanical property test results can be shown to meet theeffective Codes specified strength levels. If the specified tensileand yield strengths have increased, the new material is inherentlyqualified.

33.21 MATERIALS DOCUMENTATION Before the winter 1973 addenda, the terms CMTR and COC

were not used in the Code. Material Specifications required, orelse provided, that when requested by the purchaser, the results ofall tests and examinations would be supplied with the material.This record was (and still is) referred to as the MTR.

There are two basic documents specified by the Code for mate-rials: the CMTR and the COC (materials include pipe madefrom sheet or strip rolled and welded without the addition of fillermetal, as in autogenous welding). The contents of a typicalCMTR are described in Section III, Division 1, Appendix P.Welded-with-filler metal-type pipe requires fabrication under anNCA-4134 QA Program for an N-Type Certificate Holder; theapplication of an NPT-Code Symbol Stamp; and the completion,certification, and submission of the NM-1 Data Report Form.Nameplates, however, are not required.

FIG. 33.1 EXAMPLE OF A PROPOSED SEC CERTIFICATE OF COMPLIANCE


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Not all Material Specifications require heat traceability. Forreplacement items and for upgrading stock material, thisauthor believes that traceability is necessary to prevent material-identification problems later. The Owners QA Program assumes theresponsibility for acceptability. Without traceability, a significantcost will be incurred in qualifying unidentified material.

33.22 COMPONENT PROCUREMENTFOR REPLACEMENTS

Anyone can order a component. The reason is that a componentis a stand-alone item. Its data package must include the following:

The Owners Certified Design Specification. The Manufacturers Certified (if required) Design Report. The documentation of the Owners Review of the Design

Report (if applicable). The appropriate Data Report Forms. All required permanent and nonpermanent records. Certifications specified in NCA-4134.17.

The component can be ordered to any Code, as described inNCA-1140 and IWA-4140. Because of the rapidly diminishingmanufacturers still retaining their Certificates of Authorization,the NRC has issued guidelines for use of commercial grade dedi-cation items, or CGD. CGD programs are becoming increasinglyimportant to operating utilities, but discussion is beyond the scopeof this chapter. Some aspects of a CGD program are included inExample B (Section 33.19).

33.23 USE OF SPECIFIC PROVISIONS OFLATER EDITIONS AND ADDENDA

The nature of the rules and the definitions of nuclear poweritems are included in subsubarticles NCA-1110 and NCA-1120.NCA-1130(a) limits the scope of the rules to new construction. Itrequires consideration of mechanical and thermal stresses causedby cyclic operation. NCA-1130(b) lists one of the important setsof exclusions from the rules of Section III. Exempted are valveoperators, controllers, position indicators, pump impellers, pumpdrivers, and other accessories and devices that are not pressure-retaining. Also exempted are intervening elements used as com-ponent or piping supports; instruments and permanently sealedfluid-filled tubing systems furnished with instruments; and instru-ment, control, and sampling piping unless they are specified asCode items in the Design Specifications.

Before the winter 1977 addenda, the Code of Record for acomponent was the Code in effect on its order date. The summer1977 addenda made provisions for Division 2 Components. Thewinter 1977 revision permitted the Owner to specify one editionand addenda of the Code to be used for all components at a plantsite. The date that the Owner chooses for his or her plant cannotbe any earlier than three years before the docket date of the con-struction permit application. The change allows duplicate nuclearplants at the same site (or replicate plants at different sitesevenwith different Owners) to be built to the same Code within athree-year umbrella period between nuclear plant PSAR filings.The NRC has increased the umbrella to five years, but the Codehas not. NCA-1140 still allows updating by means of Code Cas

history of the code rules for accreditation, certification, and related issues

Documents

nuclear code

set of rules

proposed rules

massachusetts rules

applicable rules

pressurevessel code

asme boiler codewas

president ofthe asme