52730 Federal Register / Vol. 73, No. 176 / Wednesday, September 10, 2008 / Rules and Regulations 

NUCLEAR REGULATORY 

COMMISSION 

10 CFR Part 50 

RIN 3150–AH76 

[NRC–2007–0003] 

Industry Codes and Standards; 

Amended Requirements 

AGENCY: Nuclear Regulatory 

Commission. 

ACTION: Final rule. 

SUMMARY: The U.S. Nuclear Regulatory 

Commission (NRC) is amending its 

regulations to incorporate by reference 

the 2004 Edition of Section III, Division 

1, and Section XI, Division 1, of the 

American Society of Mechanical 

Engineers (ASME) Boiler and Pressure 

Vessel Code (BPV Code), and the 2004 

Edition of the ASME Code for Operation 

and Maintenance of Nuclear Power 

Plants (OM Code) to provide updated 

rules for constructing and inspecting 

components and testing pumps, valves, 

and dynamic restraints (snubbers) in 

light-water nuclear power plants. The 

NRC also is incorporating by reference 

ASME Code Cases N–722, ‘‘Additional 

Examinations for PWR [pressurized 

water reactor (PWR)] Pressure Retaining 

Welds in Class 1 Components 

Fabricated with Alloy 600/82/182 

Materials, Section XI, Division 1,’’ and 

N–729–1, ‘‘Alternative Examination 

Requirements for PWR Reactor Vessel 

Upper Heads With Nozzles Having 

Pressure-Retaining Partial-Penetration 

Welds, Section XI, Division 1,’’ both 

with conditions. The amendment also 

removes certain obsolete requirements 

specified in the NRC’s regulations. This 

action is in accordance with the NRC’s 

policy to periodically update the 

regulations to incorporate by reference 

new editions and addenda of the ASME 

Codes and is intended to maintain the 

safety of nuclear reactors and make NRC 

activities more effective and efficient. 

DATES: Effective Date: October 10, 2008. 

The incorporation by reference of 

certain publications listed in the 

regulation is approved by the Director of 

the Office of the Federal Register as of 

October 10, 2008. 

ADDRESSES: You can access publicly 

available documents related to this 

document using the following methods: 

Federal e-Rulemaking Portal: Go to 

http://www.regulations.gov and search 

for documents filed under Docket ID 

[NRC–2007–0003]. Address questions 

about NRC dockets to Carol Gallagher 

301–415–5905; e-mail 

Carol.Gallagher@nrc.gov. 

NRC’s Public Document Room (PDR): 

The public may examine and have 

copied for a fee publicly available 

documents at the NRC’s PDR, Public 

File Area O1F21, One White Flint 

North, 11555 Rockville Pike, Rockville, 

Maryland. 

NRC’s Agencywide Documents Access 

and Management System (ADAMS): 

Publicly available documents created or 

received at the NRC are available 

electronically at the NRC’s electronic 

Reading Room at http://www.nrc.gov/ 

reading-rm/adams.html. From this page, 

the public can gain entry into ADAMS, 

which provides text and image files of 

NRC’s public documents. If you do not 

have access to ADAMS or if there are 

problems in accessing the documents 

located in ADAMS, contact the NRC’s 

PDR reference staff at 1–800–397–4209, 

301–415–4737, or by e-mail to 

pdr.resource@nrc.gov. 

FOR FURTHER INFORMATION CONTACT: L. 

Mark Padovan, Office of Nuclear 

Reactor Regulation, U.S. Nuclear 

Regulatory Commission, Washington, 

DC 20555–0001, telephone 301–415– 

1423, e-mail Mark.Padovan@nrc.gov. 

SUPPLEMENTARY INFORMATION: 

I. Background 

II. Analysis of Public Comments 

III. Section-by-Section Analysis 

IV. Generic Aging Lessons Learned Report 

V. Availability of Documents 

VI. Voluntary Consensus Standards 

VII. Finding of No Significant Environmental 

Impact: Environmental Assessment 

VIII. Paperwork Reduction Act Statement 

IX. Regulatory Analysis 

X. Regulatory Flexibility Certification 

XI. Backfit Analysis 

XII. Congressional Review Act 

I. Background 

The NRC is amending 10 CFR 50.55a 

to incorporate by reference the 2004 

Edition of Section III, Division 1 and 

Section XI, Division 1 of the ASME BPV 

Code and the 2004 Edition of the ASME 

OM Code. Section 50.55a requires the 

use of Section III, Division 1 of the 

ASME BPV Code for the construction of 

nuclear power plant components; 

Section XI, Division 1 of the ASME BPV 

Code for the inservice inspection (ISI) of 

nuclear power plant components; and 

the ASME OM Code for the inservice 

testing (IST) of pumps and valves. The 

NRC published a proposed rulemaking 

on this subject in the Federal Register 

on April 5, 2007 (72 FR 16731). The 75- 

day public comment period for the 

proposed rule closed on June 19, 2007. 

The introductory paragraph of 

§ 50.55a establishes the applicability of 

the conditions therein to licenses and 

approvals issued under Part 52. 

Specifically, that rule states the 

following: 

• ‘‘Each combined license for a 

utilization facility is subject to the 

following conditions in addition to 

those specified in § 50.55, except that 

each combined license for a boiling or 

pressurized water-cooled nuclear power 

facility is subject to the conditions in 

paragraphs (f) and (g) of this section, but 

only after the Commission makes the 

finding under § 52.103(g) of this 

chapter.’’ 

• ‘‘Each manufacturing license, 

standard design approval, and standard 

design certification application under 

part 52 of this chapter is subject to the 

conditions in paragraphs (a), (b)(1), 

(b)(4), (c), (d), (e), (f)(3), and (g)(3) of this 

section.’’ 

Accordingly, combined licenses, 

manufacturing licenses, standard design 

approvals, and standard design 

certifications are subject to these 

requirements. 

The ASME BPV Code and OM Code 

are national, voluntary consensus 

standards, and are required by the 

National Technology Transfer and 

Advancement Act of 1995, Public Law 

104–113, to be used by government 

agencies unless the use of such a 

standard is inconsistent with applicable 

law or is otherwise impractical. The 

NRC reviews new editions and addenda 

of the ASME BPV and OM Codes, and 

periodically updates § 50.55a to 

incorporate by reference newer editions 

and addenda. New editions of the 

subject codes are issued every 3 years; 

addenda to the editions are issued 

yearly except in years when a new 

edition is issued. The editions and 

addenda of the ASME BPV and OM 

Codes were last incorporated by 

reference into the regulations in a final 

rule dated October 1, 2004 (69 FR 

58804). In that rule, § 50.55a was 

revised to incorporate by reference the 

2001 Edition, and 2002 and 2003 

Addenda, of Sections III and XI, 

Division 1, of the ASME BPV Code and 

the 2001 Edition, and 2002 and 2003 

Addenda, of the ASME OM Code. 

The NRC is now incorporating by 

reference Section III, Division 1, of the 

2004 Edition of the ASME BPV Code; 

Section XI, Division 1, of the 2004 

Edition of the ASME BPV Code subject 

to modifications and limitations; and 

the 2004 Edition of the ASME OM Code. 

II. Analysis of Public Comments 

The NRC received 23 letters and emails from the public that provided 

about 87 comments on the proposed 

rule. These comments were submitted 

by individuals, nuclear utilities, and 

nuclear industry organizations 


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consisting of the Nuclear Energy 

Institute (NEI), the Performance 

Demonstration Initiative, and the 

Strategic Teaming and Resource Sharing 

(STARS) organization. The NRC 

reviewed and considered the comments 

in its final rulemaking, as discussed in 

the following sections: 

1. 10 CFR 50.55a(b)(1) 

Public Comment: 

In a letter dated June 12, 2007, G.C. 

Slagis Associates commented that the 

reversing dynamic load rules of the 

ASME BPV Code, Section III, should not 

be approved for new construction. The 

commenter stated that the draft rule 

language incorporated the 2004 Edition 

of the Section III piping rules (NB/NC/ 

ND–3600) for evaluation of ‘‘reversing 

dynamic loads,’’ whereas the NRC had 

taken exception to these rules in the 

past. The commenter also stated that 

these piping rules should not be 

approved for new construction. 

NRC Response: 

The NRC has not approved the 

reversing dynamic load rules in the 

piping rules for the ASME BPV Code, 

Section III for new construction or 

existing nuclear plants. The NRC 

believes that the commenter’s 

interpretation of the proposed rule was 

based on the wording contained in the 

summary of the proposed revisions to 

10 CFR 50.55a (on the bottom of page 

72 FR 16732 and top of page 72 FR 

16733; April 5, 2007) that said ‘‘The 

proposed rule would revise 

§ 50.55a(b)(1) to incorporate by 

reference the 2004 Edition of Section III 

of the ASME Boiler and Pressure Vessel 

(BPV) Code. The NRC does not propose 

to adopt any limitations with respect to 

the 2004 Edition of Section III.’’ The 

wording in the second sentence 

contained an editorial error. The 

sentence should have read ‘‘The NRC 

does not propose to adopt any 

additional limitations with respect to 

the 2004 Edition of Section III.’’ The 

proposed rule language on page 72 FR 

16740 retained the previous restriction 

regarding the piping rules. The 

restriction applies to the 1994 Edition 

through the 2004 Edition. To clarify 

this, the NRC revised the subject 

sentences in Section III, Section-by 

Section Analysis, of this document as 

follows: 

The final rule revises § 50.55a(b)(1) in the 

current regulation to incorporate by reference 

the 2004 Edition of Section III of the ASME 

BPV Code into 10 CFR 50.55a. The NRC is 

not adopting any additional limitations with 

respect to the 2004 Edition of Section III. 

2. 10 CFR 50.55a(b)(1)(iii)—Seismic 

Design of Piping 

Public Comment: 

In a letter dated June 19, 2007, 

Westinghouse Electric Company 

requested that the NRC clarify the 

current limitation specified in 

§ 50.55a(b)(1)(iii) regarding seismic 

design. The commenter stated that the 

limitations are related to the treatment 

of piping. However, as is stated in 

§ 50.55a(b)(1)(iii), the rules in Article 

NB–3200 of Section III of the ASME 

BPV Code contain criteria applicable to 

the seismic design of components other 

than piping systems. The commenter 

recommended that the wording in 

§ 50.55a(b)(1)(iii) be revised to clarify 

that the limitation only applies to the 

seismic design of piping. 

NRC Response: 

The NRC agrees with the commenter, 

and has revised § 50.55a(b)(1)(iii) in this 

final rule as follows: 

Seismic design of piping. Applicants and 

licensees may use Articles NB–3200, NB–

3600, NC–3600, and ND–3600 for seismic 

design of piping up to and including the 

1993 Addenda, subject to the limitation 

specified in paragraph (b)(1)(ii) of this 

section. Applicants and licensees may not 

use these Articles for seismic design of 

piping in the 1994 addenda through the latest 

edition and addenda incorporated by 

reference in paragraph (b)(1) of this section. 

3. 10 CFR 50.55a(b)(2)(xv)—Appendix 

VIII Specimen Set and Qualification 

Requirements 

Public Comment: 

Conflicts between §§ 50.55a(b)(2)(xv) 

and 50.55a(b)(2)(xxiv) were identified 

by the Performance Demonstration 

Initiative (letter dated May 11, 2007), 

Nuclear Management Company (letter 

dated June 19, 2007), and Mr. Michael 

Gothard (comment received on the 

NRC’s public Web site on May 11, 

2007). The proposed rule extends the 

application of § 50.55a(b)(2)(xv) from 

the 1995 Edition through the 2001 

Edition to the 1995 Edition through the 

2004 Edition. 10 CFR 50.55a(b)(2)(xxiv) 

prohibits the use of Appendix VIII of 

Section XI, 1995 Edition through the 

2001 Edition, and the supplements of 

Appendix VIII and Article I–3000 of the 

2002 Addenda through the latest edition 

and addenda incorporated by reference 

in § 50.55a(b). The proposed change in 

§ 50.55a(b)(2)(vx) creates confusion, 

unnecessary burden, and conflicting 

requirements. The commentors 

proposed leaving § 50.55a(b)(2)(xv) 

unchanged. 

NRC Response: 

The NRC agrees with the commentors 

that the requirements in 

§§ 50.55a(b)(2)(xv) and 

50.55a(b)(2)(xxiv) conflict. The intent of 

the proposed rule was to minimize the 

burden associated with reconciling an 

existing Appendix VIII of Section XI, 

1995 Edition through the 2001 Edition, 

program with changes that occurred in 

the 2002 Addenda and later edition and 

addenda. In keeping with the NRC’s 

intent, § 50.55a(b)(2)(xv) will reference 

up to, and including, the 2001 Edition 

of Appendix VIII as follows: 

Appendix VIII specimen set and 

qualification requirements. The following 

provisions may be used to modify 

implementation of Appendix VIII of Section 

XI, 1995 Edition through the 2001 Edition. 

Licensees choosing to apply these provisions 

shall apply all of the following provisions 

under this paragraph except for those in 

§ 50.55a(b)(2)(xv)(F) which are optional. 

Licensees who use later editions and 

addenda than the 2001 Edition of Section XI 

of the ASME Code shall use the 2001 Edition 

of Appendix VIII. 

4. 10 CFR 50.55a(b)(2)(xx)—System 

Leakage Tests 

Public Comment: 

In a letter dated June 19, 2007, 

Progress Energy stated that the 

construction code requirement for a 

hydrostatic pressure test is not 

performed at a pressure that constitutes 

a challenge to the material. A 

hydrostatic test at this pressure does not 

contribute to safety any more than a 

pressure test at operating pressure, since 

both are conducted below the yield 

strength of the materials involved. 

Therefore, from a safety perspective, the 

hydrostatic test is not used to verify the 

structural integrity of the component or 

system being tested. It only proves leak 

tightness, which is also accomplished 

by a system leakage test. Hence, the end 

results of the hydrostatic test and the 

system leakage test are the same (leak 

tightness is verified). The additional 

nondestructive examination (NDE) 

being suggested by the NRC is of no 

value in verifying leak tightness, and 

thus is not related to the safety 

significance of not performing a 

hydrostatic test. The construction code 

NDE that is implemented by ASME 

Code, Section XI (IWA–4500, 

[‘‘Examination and Testing’’]), is all that 

is needed to verify any welding 

discontinuities that could affect the 

required joint efficiency for the required 

quality of the weld or brazed joint. 

NRC Response: 

Subarticle IWA–4540(a) of the 1995 

Edition of the ASME BPV Code, Section 

XI, requires that after repair and 

replacement activities, a system 

hydrostatic pressure test be performed. 

The industry asserted that the 

hydrostatic pressure test creates a 

significant hardship. Subsequently, the 

ASME Committee developed Code Case 

N–416–3, ‘‘Alternative Pressure Test 

Requirements for Welded Repairs or 


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Installation of Replacement Items by 

Welding Class 1, 2, and 3, Section XI, 

Division 1,’’ to allow the use of system 

leakage testing and NDE to replace the 

hydrostatic test. Later, the technical 

provisions of Code Case N–416–3 were 

incorporated into the 2001 Edition of 

ASME Section XI, IWA–4540(a) and 

maintained through the 2002 Addenda. 

However, the NDE requirements of 

IWA–4540(a) were eliminated from the 

2003 Addenda of the Code. Therefore, 

the NRC proposed a condition in 

§ 50.55a(b)(2)(xx) requiring Section III 

NDE be performed following repair and 

replacement activities if a system 

leakage test was to be used in lieu of a 

hydrostatic test under the 2003 

Addenda through the latest edition and 

addenda incorporated by reference in 10 

CFR 50.55a(b)(2). 

The piping systems in some vintage 

nuclear power plants were fabricated in 

accordance with American National 

Standards Institute (ANSI)/ASME B31.1, 

‘‘Power Piping,’’ Code. ANSI/ASME 

B31.1 does not require a volumetric 

examination for those systems that 

would now be classified as ASME Class 

2 and Class 3 piping systems during 

original construction. The current 

ASME BPV Code, Section XI (IWA–

4500), allows licensees to use the NDE 

requirement of the original construction 

code as part of repair/replacement 

activities. Licensees of these vintage 

plants would not need to perform 

volumetric examinations after repair/ 

replacement activities for piping 

classified as ASME Class 2 or Class 3 

piping for which ANSI B31.1 does not 

require NDE. A system pressure test or 

hydrostatic pressure test does not verify 

the structural integrity of the repaired 

piping components. However, it is 

generally recognized in the industry that 

the volumetric examinations do provide 

significant information relative to the 

structural integrity of the repaired 

piping components. For those Class 2 

and 3 piping systems that may not 

receive a volumetric examination for the 

life of the systems, the NRC is 

concerned that performance of a system 

leakage test without associated 

volumetric examinations would not 

adequately ensure high quality welds 

for the repaired or replaced component. 

Therefore, performance of a Section III 

volumetric examination in connection 

with a system leakage test in repair/ 

replacement activities is necessary. 

Public Comment: 

In letter dated June 13, 2007, ASME 

stated that § 50.55a(b)(2)(xx) does not 

explicitly state that the NDE shall be 

performed after the system leakage test. 

As written, a licensee could comply 

with this requirement by performing the 

required NDE before the system leakage 

test. It is common practice to perform 

this NDE prior to the system leakage 

test. 

NRC Response: 

The NRC agrees with the commenter 

that an ASME BPV Code, Section III, 

1992 Edition, volumetric examination 

performed as part of the repair/ 

replacement activities prior to the 

system leakage test can be accepted to 

fulfill the NDE requirement of 

§ 50.55a(b)(2)(xx)(B). The NRC’s 

position has been, and continues to be, 

that the NDE performed as part of the 

repair/replacement activities satisfies 

the NDE provision of subarticle IWA–

4540(a) of the 2002 Addenda of the 

ASME Code, Section XI. 

Public Comment: 

In letter dated June 19, 2007, Duke 

Energy stated that § 50.55a(b)(2)(xx) 

does not restrict a licensee from using 

the provisions of IWA–5213(a) in the 

2003 Addenda of Section XI. Therefore, 

licensees may currently use the 

provisions of IWA–4540(a) in the 2003 

Addenda without having to perform 

NDE in accordance with the 

requirements of IWA–4540(a)(2) of the 

2002 Addenda after a system leakage 

test. Because the proposed change 

imposes additional requirements on 

licensees, the change should be 

evaluated to determine whether the 

change is a backfit. 

NRC Response: 

The NRC agrees with the commenter 

that the proposed requirement would 

result in a backfit for some licensees 

because this final rule would now 

require them to perform the required 

NDE in conjunction with the system 

leakage test in lieu of the hydrostatic 

test. In the October 1, 2004 (69 FR 

58804), rulemaking of the 2003 

Addenda of the ASME Code, the NRC 

neglected to incorporate the above NDE 

requirement in 10 CFR 50.55a(b)(2). 

However, the oversight needs to be 

corrected to ensure that during repair or 

replacement activities, the volumetric 

examination, in conjunction with a 

system leakage test, is performed to 

ensure structural integrity of the 

repaired or replaced piping system. The 

NRC discusses its backfit analysis for 

those licensees who may be affected by 

this rule in Section XI, Backfit Analysis, 

of this document. 

5. 10 CFR 50.55a(b)(2)(xxi)(A)—Table 

IWB–2500–1 Examination Requirements 

Public Comment: 

In letter dated June 13, 2007, ASME; 

in letter dated June 19, 2007, Nuclear 

Energy Institute; and in letter dated June 

19, 2007, Duke Energy disagree with 

modifying the limitation to require 

visual examination of Class 1 

pressurizer and steam generator nozzle 

inner radius areas (ASME Code Case N–

619) based on the previous reactor 

vessel nozzle inner radius limitation 

(ASME Code Case N–648–1). The 

commenters believe that the original 

limitation (to continue examination of 

the inner nozzle radius region) is 

unnecessary because of the following: 

a. Inner nozzle radius regions in Class 

1 systems have been examined for over 

25 years without detecting cracking. 

b. Structural integrity evaluations 

demonstrated a large tolerance for flaws. 

c. Risk informed evaluations 

demonstrated that these nozzles have a 

large tolerance for flaws. 

d. Risk informed evaluations 

demonstrated a low probability of 

failure under plant operating 

conditions. 

e. There is a negligible change in risk 

if inspections are eliminated. 

f. The term enhanced VT–1 is not 

defined in Code, and studies show that 

VT–1 character heights provide the 

same or better resolution than the 1 mil 

wire. 

NRC Response: 

The NRC disagrees with the 

commentors. The limitation on the 

visual examination in 10 CFR 

50.55a(b)(2)(xxi)(A) did not differentiate 

between vessel components. The 

limitation is an alternative for 

volumetric examinations. The proposed 

change in the rule is to provide a visual 

examination criterion for determining 

fatigue crack flaw depth. 

With respect to Item 5.a above, the 

commentor’s information on 25 years of 

inservice ultrasonic examinations with 

no evidence of inner radius cracking on 

nozzles covered by the ASME Code 

cases is from an ASME document issued 

in 2001. At that time, ultrasonic 

examinations of pressurized-water 

reactors were normally performed from 

the inside surface, and were normally 

performed from the outside surface for 

boiling-water reactors. The NRC took 

issue with the effectiveness of ultrasonic 

examinations of the inner nozzle radius 

performed prior to performance-based 

qualification requirements. 

Performance-based examinations of all 

reactor pressure vessel (RPV) inner 

nozzle radii became mandatory on 

November 22, 2002. On July 26, 2006, 

the Electric Power Research Institute—

Boiling Water Reactor Vessel & Internal 

Project (BWRVIP) provided a summary 

of results from inner nozzle radius 

performance-based examinations to 

support reducing RPV inner nozzle radii 

examination frequency by 75 percent. 

By letter dated December 19, 2007, 

the NRC issued a safety evaluation 


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accepting BWRVIP–108 which reduced 

the inspection frequency of reactor 

nozzle-to-vessel shell welds and nozzle 

inner radius for BWRs (NRC ADAMS 

Accession Number ML073600374). 

Operating conditions, such as 

fluctuating temperature, and fabricating 

conditions, such as work hardening can 

cause cracking of the inner nozzle 

radius. The ASME Code Cases (N–619 

and N–648–1) are silent on conditions 

that are associated with cracking. These 

conditions may appear, or be affected, at 

various times during the operating cycle 

and may not be specific to vessel design. 

To detect degradation that appears 

during operations, NDE of inner nozzle 

radii are warranted. 

Items 5b, 5c, and 5d pertained to riskinformed computations. Of the riskinformed piping programs reviewed to 

date, none of the programs contained 

risk data for Class 1 inner nozzle radius 

regions. The NRC did not find 

documentation of a review on the ASME 

2001 article. Recently, the BWRVIP 

submitted to the NRC information on 

structural integrity and probability of 

failure and risk calculations concerning 

the inspection of inner nozzle radius 

regions to the NRC for review, which is 

ongoing. 

With respect to Item 5f, the 

commentors referenced proprietary 

documents that were not made available 

to the NRC. Therefore, the NRC was 

unable to verify the data used to 

validate the adequacy of VT–1 and of 

character recognition for examinations 

of the inner radii regions. While 

characters are useful for distinguishing 

shapes, NUREG/CR 6860, ‘‘An 

Assessment of Visual Testing,’’

identified the crack open width 

dimension as a key variable for visually 

detecting cracks. In 10 CFR 

50.55a(b)(2)(xxi)(A), the 1-mil width 

wire or crack is a measurable criterion 

for a postulated crack open width 

dimension. Therefore, the 1-mil width 

wire or crack requirement provides a 

minimum criteria for performance-based 

demonstrations of examination 

effectiveness. 

The commentors stated that the term 

‘‘enhanced VT–1’’ was not recognized 

by the ASME BPV Code. The term 

‘‘enhanced VT–1’’ is being used by 

knowledgeable personnel for 

conversational expediency. The term 

‘‘enhanced VT–1’’ is not used in the 

regulation. However, the use of the term 

‘‘enhanced magnification’’ is used in the 

rule and may have been misleading. 

Therefore, the term ‘‘enhanced’’ will be 

removed from the regulation. 

6. 10 CFR 50.55a(b)(2)(xxviii)—

Evaluation Procedure and Acceptance 

Criteria for PWR Reactor Vessel Head 

Penetration Nozzles 

Public Comment: 

In a letter dated June 13, 2007, the 

ASME stated that this modification is 

being proposed because of a 

typographical error that the NRC says 

exists in ASME Section XI, Nonmandatory Appendix O, paragraph O–

3220(b), equation SR, = [l—0.82R]¥22, 

where the exponent ¥22 should be 

¥2.2. ASME has identified this error 

and is publishing an ERRATA in July 

2007 to correct this error retroactively to 

include the 2004 Edition of Section XI. 

As such, the proposed amendment to 10 

CFR 50.55a(b)(2)(xxviii) is unnecessary. 

NRC Response: 

The NRC finds that ASME has 

published an ERRATA in July 2007 to 

correct the error in the SR equation of 

paragraph O–3220(b) retroactively to 

include the 2004 Edition of ASME BPV 

Code, Section XI. The condition 

imposed in § 50.55a(b)(2)(xxviii) will 

not be necessary. Therefore, the NRC is 

not including § 50.55a(b)(2)(xxviii) in 

this final rule. 

7. 10 CFR 50.55a(b)(3)(v)—Subsection 

ISTD 

Public Comments: 

By electronic mail dated June 11, 

2007, George L. Fechter of Southern 

Nuclear Operating Company stated that 

Article IWF–5000, ‘‘Inservice Inspection 

Requirements for Snubbers,’’ was 

deleted from the 2006 Addenda of the 

ASME BPV Code, Section XI. With 

adequate verification of training 

provided to personnel performing visual 

exams, removal, testing, and 

reinstallation of snubbers per applicable 

Subsection ISTD, ‘‘Inservice Testing of 

Dynamic Restraints (Snubbers) in LightWater Reactor Power Plants,’’ of the 

ASME OM Code and site licensing and 

maintenance criteria, it should be 

justifiable to allow performance of this 

type of visual examination versus a VT–

3 visual examination. The knowledge 

obtained from such snubber-specific 

training and experience commonly 

exceeds the VT–3 visual examination 

criteria for snubbers. While IWA–2317 

of the 2003 Addenda through 2004 

Edition of the ASME BPV Code, Section 

XI, provides alternative VT–3 

examination qualification requirements, 

the administrative burden incurred for 

the VT–3 certification may not be 

commensurate with any convenience 

provided by qualifying additional VT–3 

personnel in this manner and, for 

reasons stated previously, does not 

provide higher quality examinations. 

The commenter requested that the 

permissive for allowing personnel 

trained specifically on snubber 

requirements per the applicable ISTD 

and site licensing and maintenance 

criteria be allowed to perform visual 

examinations for snubbers as an 

alternative to performing a VT–3 

examination per the method described 

in IWA–2213 of the ASME BPV Code, 

Section XI. 

NRC Response: 

The commenter requested that the 

visual examination method required by 

§ 50.55a(b)(3)(v) when performing 

examination and testing of snubbers be 

revised. The NRC declines to adopt the 

commenter’s suggestion because the 

proposed rule did not suggest an 

amendment to the visual examination 

method in § 50.55a(b)(3)(v), and the 

NRC currently does not have a basis for 

supporting such a revision. There were 

no other public comments received on 

§ 50.55a(b)(3)(v). Therefore, the NRC 

declines to adopt the commenter’s 

suggestion. No change was made to 

§ 50.55a(b)(3)(v) in the final rule as a 

result of the comment. 

8. 10 CFR 50.55a(g)(6)(ii)(B)—

Containment ISI Programs 

Public Comments: 

In a letter dated June 19, 2007, Duke 

Energy stated that when compliance 

with the requirements of the ASME BPV 

Code, Section XI, Subsections IWE and 

IWL was initially imposed by 10 CFR 

50.55a, the requirements of 

§ 50.55a(g)(6)(ii)(B) did not require 

licensees to submit ISI programs that 

were developed to comply with the 

Code during the expedited examination 

period (September 9, 1996, through 

September 9, 2001). However, when the 

initial expedited examination 

requirements were removed from 

§ 50.55a after September 9, 2001, 

§ 50.55a(g)(6)(ii)(B) was not deleted, 

leaving some licensees to believe that 

the NRC wanted to retain this provision. 

As a result, many licensees continue to 

believe that the NRC does not want 

updated containment ISI plans to be 

submitted. The NRC should take action 

to clarify whether it is the intent of 10 

CFR 50.55a(g)(6)(ii)(B) that licensees be 

required to submit ISI plans for Class 

MC and Class CC components for all ISI 

plans developed after the expedited 

examination period. 

NRC Response: 

The NRC notes that the comment was 

not related to the proposed rule but to 

seek clarification on § 50.55a(g)(6)(ii)(B) 

in the current regulation. It is the NRC’s 

position to retain the current 

§ 50.55a(g)(6)(ii)(B) provision in the 

final rule. § 50.55a(g)(6)(ii)(B) states that 


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licensees do not have to submit to the 

NRC for approval of their containment 

in-service inspection (CISI) programs for 

Class MC and Class CC pressure 

retaining components that were 

developed to meet the requirements of 

the ASME BPV Code, Section XI, 

Subsections IWE and IWL, with 

specified modifications and limitations, 

under § 50.55a(g)(5)(i) and/or 

§ 50.55a(g)(4). The provision requires 

that program elements and the required 

documentation of the developed plan 

must be maintained on site for audit. 

The provision applies to the CISI 

programs developed for each operating 

license for the initial 120-month 

inspection interval, including the CISI 

program revisions made by licensees of 

operating reactors during the September 

1996 to September 2001 timeframe (i.e., 

expedited examination period) when 

the rule for ASME BPV Code, Section 

XI, compliance was initially imposed. 

Further, the provision applies to 

subsequent revisions to the CISI 

programs for successive 120-month 

inspection intervals under 

§ 50.55a(g)(4)(ii). Therefore, as stated in 

§ 50.55a(g)(6)(ii)(B), licensees do not 

have to submit to the NRC for approval 

of their CISI program that meets the 

ASME Code, Subsections IWE and IWL 

with specified modifications and 

limitations after the expedited 

examination period. 

However, the NRC would like to 

clarify a situation which does not affect 

50.55a(g)(6)(ii)(B) directly but which 

involves the use of Subsections IWE and 

IWL. If a licensee wishes to use 

Subsections IWE and IWL of later 

editions and addenda (i.e., later than the 

code of record for the ISI interval in 

question) of the ASME Code that are 

incorporated by reference in 10 CFR 

50.55a(b) to be applied to the specific 

10-year inservice inspection interval at 

its nuclear plant, the licensee needs to 

submit a request for the NRC’s approval 

to use the later editions and addenda of 

the ASME Code. As stated in 

§ 50.55a(g)(4)(iv), licensees are required 

to obtain NRC approval before using 

subsequent editions and addenda (or 

portions thereof) of the ASME BPV 

Code, Section XI, issued after their Code 

of Record for any 120-month inspection 

interval, if they choose to implement 

their ISI programs under 

§ 50.55a(g)(4)(iv). The regulatory issue 

of using later editions and addenda of 

the Code has been previously clarified 

in NRC Regulatory Issue Summary 

2004–12, ‘‘Clarification on Use of Later 

Editions and Addenda to the ASME OM 

Code and Section XI.’’ The intent of the 

commenter is to seek a clarification 

rather than a suggestion. Therefore, no 

change was made to § 50.55a(g)(6)(ii)(B) 

in the final rule as a result of this 

comment. 

9. 10 CFR 50.55a(g)(6)(ii)(D)—Reactor 

Vessel Head Inspections 

9a. Condition 10 CFR 

50.55a(g)(6)(ii)(D)(1), Regarding the 

Implementation of Code Case N–729–1, 

as Amended, in Lieu of the First 

Revised NRC Order EA–03–009 

Some commenters requested 

additional information on the 

implementation of these requirements, 

and asked the NRC about the process of 

changing the current NRC requirements 

for RPV closure head inspection 

requirements from the First Revised 

NRC Order EA–03–009, issued on 

February 20, 2004, (Order) to the 

requirements provided in the proposed 

rule language for 10 CFR 

50.55a(g)(6)(ii)(D). (Comment Numbers 

14, 19 and 20) 

NRC Response: 

To allow an orderly implementation 

of 10 CFR 50.55a(g)(6)(ii)(D), the NRC 

finds an implementation date of no later 

than December 31, 2008, for the 

requirements provided in this section is 

warranted. The requirements of NRC 

Order EA–03–009 will remain in effect 

until the provisions of 10 CFR 

50.55a(g)(6)(ii)(D) are implemented. 

Once a licensee implements this 

requirement, the First Revised NRC 

Order EA–03–009 no longer applies to 

that licensee and under 10 CFR 

50.55a(g)(6)(D)(1) shall be deemed to be 

withdrawn. All relaxations from the 

requirements of the Order will then no 

longer apply. If a licensee cannot meet 

the proposed requirements of 10 CFR 

50.55a(g)(6)(ii)(D), then an alternative 

may be requested in accordance with 10 

CFR 50.55a(a)(3)(i) or 10 CFR 

50.55a(a)(3)(ii) or impracticality must be 

shown under 10 CFR 50.55a(g)(6)(i). To 

incorporate this implementation date, 

section 50.55a(g)(6)(ii)(D)(1) is revised 

to incorporate this implementation date. 

9b. Condition 10 CFR 

50.55a(g)(6)(ii)(D)(2), Regarding the 

Frequency of Reactor Vessel Head 

Inspection for ‘‘Resistant’’ Materials 

Public Comment: 

Some commenters disagreed with the 

proposed NRC position regarding the 

frequency of inspection of Item No. 

B4.40 of ASME Code Case N–729–1. 

The commenters made several remarks 

regarding previous and ongoing 

laboratory work with primary water 

stress corrosion cracking (PWSCC) 

‘‘resistant’’ materials. Further, they 

noted operational experience with these 

materials had provided a sufficient basis 

to allow the inspection interval as stated 

in ASME Code Case N–729–1 without 

the NRC-proposed condition, as 

provided in proposed 10 CFR 

50.55a(g)(6)(ii)(D)(2). One commenter, 

number 13, recommended extending the 

interval of inspection from every seven 

(7) years to every eight (8) years. 

(Comment Numbers 7, 9, 11, 13, 15, 16, 

17, 19, 21, 22 and 23) 

NRC Response: 

During the writing of the proposed 

rule, the NRC disagreed with the NDE 

re-inspection frequency for ‘‘resistant’’

materials, in Item B4.40 of Table 1 of 

ASME Code Case N–729–1, of every ten 

(10) calendar years beyond the first 10 

years. Therefore, the NRC proposed the 

condition 10 CFR 50.55a(g)(6)(ii)(D)(2) 

to limit the inspection frequency for 

‘‘resistant’’ materials to every four 

refueling outages not to exceed seven (7) 

calendar years beyond the first 10 years. 

The proposed condition was based on 

two main factors: the availability of 

limited crack initiation and growth data 

on the Alloy 152/52 weld metal, and the 

accelerated susceptibility increases of 

replaced U.S. RPV heads versus the 

current operational experience data 

from international experience which 

demonstrates the resistance of Alloy 

690/152/52 materials against PWSCC. 

The available data on Alloy 152/52 

weld metal resistance to PWSCC is an 

NRC concern. However, considering the 

comments on this issue and ongoing 

PWSCC research programs at Pacific 

Northwest National Laboratories and 

Argonne National Laboratory sponsored 

by the NRC Office of Nuclear Regulatory 

Research, NRC now finds that the 

current data is sufficient to support the 

re-inspection frequency of Item B4.40 of 

Table 1 of ASME Code Case N–729–1. 

NRC research on these materials is 

scheduled to continue through CY 2010. 

Accordingly, there should be enough 

time to address any items of concern 

regarding the resistance of these 

materials to PWSCC, if and when they 

develop, prior to becoming a significant 

safety issue. 

The NRC acknowledges that current 

operating experience shows the 

resistance of Alloy 152/52 weld material 

to PWSCC to be superior to that of Alloy 

82/182. However, RPV head 

temperatures at numerous international 

plants with replaced RPV upper heads 

are significantly less than U.S. upperhead temperatures. As PWSCC 

susceptibility in nickel based alloys like 

Alloy 600 has been shown to have a 

significant temperature dependence, 

NRC analysis of international head 

replacement data has shown that RPV 

heads in the U.S. will, with time, have 


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a greater susceptibility to PWSCC than 

a majority of the international plants in 

terms of accumulated, effective 

degradation years. Therefore, NRC has 

found that long-term operating 

experience is limited for components 

that contain Alloy 690/52/152 materials 

with indications and repairs of the 

scope and nature found in recently 

replaced U.S. RPV heads. Nevertheless, 

the NRC finds the operational 

experience is sufficient to support Code 

Case N–729–1 inspection frequencies 

while research on these materials 

continues. 

The NRC agrees with the commenters 

and finds that there is sufficient Alloy 

690/152/52 laboratory data and 

operational experience to allow the 

inspection frequency of Item B4.40 of 

Table 1 of ASME Code Case N–729–1 

for RPV upper heads containing Alloy 

690/152/52 components. Therefore, the 

proposed condition in 10 CFR 

50.55a(g)(6)(ii)(D)(2) of the proposed 

rule will not be adopted. 

9c. Condition 10 CFR 

50.55a(g)(6)(ii)(D)(3), Regarding RPV 

Head Inspection Requirements and 

Frequencies 

Public Comment: 

Some commenters disagreed with the 

proposed NRC condition regarding the 

implementation of Note 6 of Table 1 of 

ASME Code Case N–729–1, which is 

stated in the 10 CFR 50.55a proposed 

rule language as 10 CFR 

50.55a(g)(6)(ii)(D)(3). Several comments 

were concerned with the surface and 

volumetric examination coverage 

requirements and the surface 

examination requirement of the J-groove 

weld. The commenters requested to 

allow a UT ‘‘leak-path’’ examination in 

lieu of surface examination of the Jgroove weld, and that a note be added 

to document that Appendix I of the 

Code Case may be used when approved 

as required in 10 CFR 

50.55a(g)(6)(ii)(D)(6). In addition 

comments noted that the impact of Note 

9 is not addressed in the elimination of 

the original Code Case N–729–1, Note 6. 

(Comment Numbers 7, 9, 11, 12, 13, 16, 

17, 18, 19, 20, 22 and 23) 

NRC Response: 

In development of the proposed rule, 

the NRC did not find sufficient basis to 

allow an inspection regime of 3.0 reinspection years (RIY) as described in 

Code Case N–729–1. Further, the NRC 

noted that due to the lack of a nonvisual leak path assessment requirement 

in Code Case N–729–1, surface 

examination of all J-groove welds, 

commensurate with the volumetric 

examination of the penetration nozzle, 

should be required. Therefore the NRC 

proposed the condition in 10 CFR 

50.55a(g)(6)(ii)(D)(3). The NRC found 

the inspection coverage as defined by 

Code Case N–729–1 using the ASME 

Code definition of ‘‘essentially 100 

percent’’ inspection acceptable and 

therefore retained that language in the 

condition. No increase in inspection 

coverage is intended in the condition. 

The NRC disagrees that the 

supporting probabilistic basis is 

adequate to support the 3.0 RIY option. 

A probabilistic fracture mechanics 

analysis was used as a basis for the 3.0 

RIY inspection frequency option. NRC 

finds the supporting probabilistic model 

is based on an assumption of essentially 

no cracking in RPV head penetrations or 

welds with less than 4 effective years of 

degradation (EDY). The NRC considers 

this assumption to be non-conservative 

as used in the supporting probabilistic 

model. One U.S. plant at approximately 

2 EDY identified cracking attributable to 

PWSCC. Many of the other near-cold-leg 

temperature RPV heads (cold-head 

plants) with susceptible material will 

not accumulate a total of 4 EDY through 

the next 15 to 30 years of operation. 

Development of flaws in these heads 

would cause adjustment of the 

probabilistic model output for all 

temperature ranges of RPV heads. 

Cracking attributed to PWSCC has been 

identified internationally in head 

penetration nozzles and associated 

welds at operating temperatures similar 

to U.S. cold-head plants. In the U.S., 

flaws in other components have been 

attributed to PWSCC in similar cold-leg 

temperature environments. The NRC 

finds that relatively few more instances 

of flaws attributed to PWSCC in the 

cold-head sub-population could 

significantly change the probabilistic 

model upon which the 3.0 RIY 

inspection frequency is justified. 

Therefore, NRC concludes that the 

supporting probabilistic model does not 

provide an adequate basis for extending 

the non-visual NDE inspection 

frequency to 3.0 RIY. 

The conditional requirement for 

surface examinations of all J-groove 

welds is based on the need for a 

defense-in-depth method to ensure 

reactor coolant pressure boundary 

integrity through the J-groove weld. In 

Code Case N–729–1, the mechanism to 

identify a through-weld flaw in a Jgroove weld is through the bare-metal 

visual exam using visual leak detection 

at the top of the RPV head. This method 

alone is not consistent with previous 

NRC inspection requirements under the 

Order which require a non-visual leak 

path assessment in conjunction with a 

bare-metal visual examination of the 

RPV head. The NRC finds that not 

performing a leak path assessment 

would limit the ability of an inspection 

plan to provide sufficient defense-indepth to identify leakage through the Jgroove weld. In the past, the NRC has 

accepted ultrasonic (UT) leak path 

assessments as an adequate inspection 

to provide this assurance. However, the 

UT leak path assessment was not 

included in Code Case N–729–1 because 

it had not been qualified through the 

ASME Code process. Surface 

examination of the J-groove weld was 

included in Code Case N–729–1, but 

only as an option to increase inspection 

frequency. Under the proposed 

condition, performance of a surface 

examination of the J-groove weld would 

have been the only option in terms of 

a leak path assessment. 

The commenters stated that there are 

current plans to demonstrate the 

effectiveness of the ultrasonic leak path 

assessment technique for use within 

Code Case N–729–1. As the ultrasonic 

leak path assessment was a previously 

acceptable alternative to surface 

examination of the J-groove weld, due to 

physical constraints and radiological 

dose concerns in performing a surface 

exam in this area, the condition stated 

in 10 CFR 50.55a(g)(6)(ii)(D)(3) has been 

modified in this final rule. 

As noted previously the Condition 

stated in 10 CFR 50.55a(g)(6)(ii)(D)(2) 

was removed. To address stakeholder 

comments about confusion between 

Notes 6 and 9 of Code Case N–729–1, 

condition in 10 CFR 

50.55a(g)(6)(ii)(D)(2) of the proposed 

rule will simply state in the final rule 

that: ‘‘Note 9 of ASME Code Case N–

729–1 shall not be implemented.’’ Note 

9 of ASME Code Case N–729–1 provides 

the path for use of the 3.0 RIY 

inspection frequency interval. As 

previously stated, and as directed in the 

change to Note 6, the 3.0 RIY inspection 

frequency will not be included in the 

final rule. 

9d. Condition 10 CFR 

50.55a(g)(6)(ii)(D)(4), Regarding 

Qualification Requirements for 

Volumetric Inspection of RPV Head 

Penetration Nozzles 

Public Comment: 

Some commenters disagreed with the 

NRC-proposed condition regarding 

qualification requirements for 

volumetric examination as stated in 

Paragraph–2500 of ASME Code Case N–

729–1. This proposed condition is 

stated in 10 CFR 50.55a(g)(6)(ii)(D)(4) of 

the proposed rule. (Comment Numbers 

2, 7, 9, 11, 12, 13, 17, 19 and 22). 

NRC Response: 

The NRC notes that the condition 

stated in 10 CFR 50.55a(g)(6)(ii)(D)(4) 


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requires that reliable and effective 

ultrasonic examinations be performed to 

ensure adequate protection for public 

health and safety. Because of the 

emphasis placed on inspections of the 

penetrations, it is appropriate to 

incorporate requirements for a robust 

blind demonstration of the ability of 

personnel, procedures and equipment to 

reliably detect and characterize 

indications, consistent with the 

approach articulated in Appendix VIII 

of Section XI of the ASME BPV Code. 

As RPV head inspection frequencies 

transition to every 8 or 10 years due to 

replacement heads being installed, 

clearly defined performance 

demonstration requirements are 

necessary to ensure effective NDE. Due 

to the lack of current ASME BPV Code 

ultrasonic performance demonstration 

qualification requirements in Section 

XI, Appendix VIII, for RPV head 

penetrations, the NRC is adopting the 

conditions stated in 10 CFR 

50.55a(g)(6)(ii)(D)(4) in the final rule. 

With respect to the performance 

demonstration requirements of the 

ASME BPV Code, Section XI, Appendix 

VIII, have increased the effectiveness 

and reliability of ultrasonic 

examinations, most notably in the area 

of inspection of dissimilar metal welds. 

The development of a qualification 

program to meet the intermediate rigor 

requirements of ASME BPV Code, 

Section V, Article 14 would require an 

additional process beyond this 

rulemaking activity. As noted in 

paragraph 10 CFR 50.55a(g)(6)(ii)(C), 

implementation of performance 

demonstration requirements of 

Appendix VIII of Section XI of the 

ASME BPV Code is currently required 

by 10 CFR 50.55a for Supplements 1 

through 8, 10 and 11. At this time, there 

is no ASME BPV Code supplement to 

address performance demonstration 

requirements for the qualification of 

ultrasonic inspection of Alloy 600 base 

material. The conditions identified in 

the paragraphs 10 CFR 

50.55a(g)(6)(ii)(D)(4)(i) through 10 CFR 

50.55a(g)(6)(ii)(D)(4)(iv) of the final rule 

are consistent with the performance 

demonstration requirements of 

Appendix VIII. 

10 CFR 50.55a(g)(6)(ii)(D)(4), as stated 

in the proposed rule, is modified in the 

final rule to incorporate an 

implementation date of September 1, 

2009, in order to address the comment 

which noted that additional time would 

be required to fully implement a 

formalized qualification program. The 

implementation date in the final rule 

addresses the time necessary for 

mockup production and qualification of 

sufficient numbers of NDE personnel. 

NRC determined that the 

implementation date of September 1, 

2009, is adequate to address the current 

frequency of inspections and allow for 

enough qualified personnel resources to 

be available. During the interval 

between the effective date of the final 

rule and the implementation date, the 

NRC finds that the qualification 

requirements of Code Case N–729–1 

will provide reasonable assurance of 

public health and safety. 

With respect to the expansion of 

specimen qualification set applicability 

for a range of pipe diameters and 

thicknesses, 10 CFR 

50.55a(g)(6)(ii)(D)(4)(i) was modified. 

The commenters noted that current 

demonstrations are performed on 

typical-sized control rod drive 

mechanism penetration nozzles. These 

demonstrations are used for a variety of 

similar-sized penetration nozzles 

(incore instrumentation, control rod 

drive and control element drive) and for 

smaller-size and thickness vent-line 

nozzles. The proposed draft condition 

specimen set applicability range was 

taken from Section XI, Appendix VIII, 

Supplement 10 requirements for 

dissimilar metal welds. A change to 

increase the range of applicability was 

made to 10 CFR 50.55a(g)(6)(ii)(D)(4)(i) 

to address stakeholder comments 

concerning the number of currently 

available mockup assemblies and the 

continued use of them for a slightly 

larger range of nozzles. The commenter 

noted that a small adjustment would 

allow the current mockups to be 

applicable for similar sized penetration 

nozzles which would fall just outside of 

the range stated in the proposed draft 

rule language. The NRC has reviewed 

the requested increased range of 

applicability and finds that the nozzles 

in question have enough through-wall 

thickness to provide similar response. 

As the weakness of ultrasonic 

examination is near field resolution, an 

expanded range for pipe diameters and 

thicknesses is allowed. The NRC finds 

that the range now stated in 10 CFR 

50.55a(g)(6)(ii)(D)(4)(i) of the final rule 

is adequate to ensure representative 

specimen sets will be used in the 

qualification processes for both 

personnel and procedures over the 

entire range of penetration nozzles in 

the reactor vessel head, and address 

stakeholder concerns. 

With respect to issues that 

recommended an adjustment for 

mockup specimens to include a range of 

blind demonstration mockups 

previously manufactured, 10 CFR 

50.55a(g)(6)(ii)(D)(4)(ii) was modified 

for incorporation into the final rule. 

Specimen set flaw location 

requirements must meet several criteria 

to ensure the wide range of possible 

flaws identified through operational 

experience are captured for qualification 

of procedures, equipment, and 

personnel. The NRC has found that the 

commenters’ flaw location range 

recommendations as stated in public 

comment viii of this section 

satisfactorily meet the intent of 10 CFR 

50.55a(g)(6)(ii)(D)(4)(ii), which were 

established to ensure the entire range of 

flaws identified through operational 

experience are represented in the 

mockups. The NRC accepts the 

comments and, therefore, has modified 

the requirements of the condition stated 

in 10 CFR 50.55a(g)(6)(ii)(D)(4)(ii) for 

incorporation into the final rule. 

With respect to asking for additional 

clarity when an essential variable may 

be changed outside of its demonstration 

range, 10 CFR 50.55a(g)(6)(ii)(D)(4)(iii) 

has been revised for incorporation into 

the final rule. The identification and 

definition of essential variables is 

necessary to ensure proper applicability 

of qualification standards to each 

particular inspection. 10 CFR 

50.55a(g)(6)(ii)(D)(4)(iii) has been 

revised to include specific requirements 

if changes to essential variables occur. 

These requirements are the same as 

those required in Section XI, Appendix 

VIII general requirements of Subarticle 

VIII–2100 which are required for use 

under 10 CFR 50.55a(g)(6)(ii)(C) for 

implementation of performance 

demonstration requirements of 

Appendix VIII of Section XI of the 

ASME BPV Code. 

With respect to the objection to the 

proposed generic qualification 

requirements for depth and length 

sizing qualification, noting that the 

requirements were currently 

unachievable for a generic procedure 

and were not necessary from a safety 

standpoint, 10 CFR 

50.55a(g)(6)(ii)(D)(4)(iv) has been 

revised for incorporation into the final 

rule. Performance demonstration 

requirements provide depth sizing and 

length sizing root mean square (RMS) 

error tolerances to meet the acceptance 

standards of Table VIII–S10–1. The NRC 

reviewed the RMS error tolerances that 

the commenters recommended, and 

found the proposed RMS error 

tolerances of 1⁄8-inch (3 mm) in depth 

and 3⁄8-inch (10 mm) in length were 

adequate to ensure the validity of 

qualification. Therefore, for 

qualification of procedures, equipment, 

and personnel, the acceptance standard 

RMS error tolerance requirements were 

updated in 10 CFR 

50.55a(g)(6)(ii)(D)(4)(iv) as incorporated 

into the final rule. 


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After review and assessment of the 

comments, the NRC is revising the 

proposed condition. 

9e. Condition 10 CFR 

50.55a(g)(6)(ii)(D)(5), Regarding Reinspection Requirements Once a Plant 

has Identified PWSCC Flaws in Their 

RPV Head Penetration Nozzles or 

Associated Welds 

Public Comment: 

Some commenters disagreed with the 

NRC proposed condition 10 CFR 

50.55a(g)(6)(ii)(D)(5). This condition 

requires a volumetric and/or surface reinspection each outage once a plant 

identifies PWSCC in its vessel head 

penetration nozzles or welds. These 

commenters stated that flaw evaluation 

using the crack growth rates for PWSCC 

should provide an acceptable reinspection interval for any flaws that 

were accepted by evaluation, and an 

exemption should be added to exclude 

the condition of ‘‘craze cracking’’ from 

mandating inspections at every outage. 

(Comment Numbers 7, 9, 11, 13, 17, and 

19) 

NRC Response: 

The NRC disagrees with the 

commenters that flaw evaluation using 

the crack growth rates for PWSCC 

would provide an acceptable reinspection interval. The proposed 

condition stated in 10 CFR 

50.55a(g)(6)(ii)(D)(5) is based upon 

operating experience, and that several 

elements of PWSCC susceptibility (e.g., 

cold work, specific material properties, 

etc.) are not fully included in the 

susceptibility and probabilistic models 

of Code Case N–729–1. At least nine 

plants have identified flaws attributable 

to PWSCC in the refueling outage 

immediately following an inspection 

which identified the degradation 

mechanism. One plant identified at least 

four new flaws greater than 50 percent 

through-wall in one operational cycle of 

crack growth. The NRC finds that 

operational experience has shown that 

not all factors affecting the 

susceptibility of Alloy 600 materials are 

included within a standard flaw 

analysis model using the ASME BPV 

Code flaw analysis using the Alloy 600 

crack growth rate identified in 

Subarticle IWB–3660 of Section XI of 

the ASME BPV Code. 

The ASME BPV Code crack growth 

rate curve for Alloy 600 is a mean of the 

upper 50 percent of all acceptable Alloy 

600 laboratory developed crack growth 

rate data points. It is not a bounding 

crack growth curve. Testing on field 

samples of Alloy 600 from the replaced 

RPV head of one plant by Argonne 

National Laboratories identified a crack 

growth rate which is at the upper bound 

(95th percentile) of the data used to 

develop the ASME curve. Additional 

factors may affect the initiation and 

growth of PWSCC in RPV upper head 

penetrations which were not fully 

analyzed in the laboratory tested 

material. These factors include the 

welding process, heats of material, and 

cold work applied in the field or during 

manufacturing conditions. 

If a plant is found to have a flaw 

attributable to PWSCC, the flaw may 

have developed due to any one or a 

combination of the previously 

mentioned susceptibility factors. 

Therefore, the plant may not be fully 

bounded by the Code Case N–729–1 

PWSCC model. The model provides 

appropriate inspection frequencies to 

ascertain when a plant develops PWSCC 

in its RPV upper head penetrations. 

However, to be conservative, the plant 

should perform volumetric and/or 

surface examinations for each outage to 

provide reasonable assurance of the 

integrity of the reactor coolant pressure 

boundary and prevent leakage once 

conditions for PWSCC have been 

verified through inspection results. As 

such, the NRC’s proposed condition is 

that once a plant has identified a flaw 

attributable to PWSCC in a RPV head 

penetration or J-groove weld, that plant 

should perform visual and volumetric 

and/or surface examinations for each 

outage. This is consistent with NRC 

Order EA–03–009. Therefore, the 

proposed provisions in 10 CFR 

50.55a(g)(6)(ii)(D)(5) are adopted 

without change in the final rule. 

Indications of craze cracking have not 

previously been characterized as 

indications of PWSCC, and the NRC 

continues to find that indications of 

craze cracking are not PWSCC. 

Therefore, if a licensee determines that 

the indications in a vessel head 

penetration nozzle are a result of craze 

cracking alone, it would not be within 

the scope of proposed condition stated 

in 10 CFR 50.55a(g)(6)(ii)(D)(5). 

9f. Condition 10 CFR 

50.55a(g)(6)(ii)(D)(6), Regarding the 

Allowance of Licensee Deviation from 

the Requirements of ASME Code Case 

N–729–1 Without NRC Review and 

Approval Public Comments 

Commenters disagreed with the NRCproposed condition for use of Appendix 

I of ASME Code Case N–729–1, which 

is stated in 10 CFR 50.55a(g)(6)(ii)(D)(6). 

The comments concerned the following 

items: 

• It is not the place of the ASME BPV 

Code to require utilities to get NRC 

approval on acceptable alternatives. 

• NRC review of industry 

implementation of Appendix I of Code 

Case N–729–1 relief from the 

requirements of ASME Code Case N–

729–1 is unnecessary. 

• An exemption should be made for 

the need for NRC approval for use of 

Appendix I of Code Case N–729–1 by 

plants with new heads that use 

‘‘resistant’’ material, until PWSCC is 

identified in those heads. 

(Comment Numbers 7, 12, 13, 17 and 

19) 

NRC Response: 

Appendix I of Code Case N–729–1 

gives an analysis procedure that allows 

licensees to demonstrate the adequacy 

of an NDE zone of coverage less than 

that required by Code Case N–729–1. 

Implementation of this analysis 

procedure does not require NRC review 

and approval. In essence, Appendix I 

would allow licensees to self-approve 

relief from the requirements of Code 

Case N–729–1, essentially usurping 

NRC’s authority under 10 CFR 50.55a to 

evaluate alternatives. NRC experience in 

processing relaxation requests to Order 

requirements has shown that there was 

significant variation in technical basis 

approaches between licensees in 

proposing alternatives to the Order. For 

example, probabilistic analyses were 

used in licensee relaxation requests 

from Order requirements that the NRC 

found to have insufficient basis and 

therefore did not approve as a basis for 

relaxation. However, under Appendix I 

of Code Case N–729–1, these relaxation 

requests could be found acceptable 

without NRC review. While the NRC 

agrees that the methods provided in 

Appendix I may be used as a basis to 

request relief from the ASME Code Case 

requirements, NRC review and approval 

shall be required for deviations from 

Code Case N–729–1 examination 

coverage requirements. 

The NRC disagrees with the comment 

that excludes from this proposed 

condition new reactor vessel heads that 

use resistant material, until PWSCC is 

identified in these heads. The NRC 

notes that the flaw evaluation tools and 

susceptibility of new PWSCC resistant 

materials have not been established or 

approved by the NRC. As such, 

implementation of Appendix I of Code 

Case N–729–1 would be open to 

significant variation of interpretation. 

Therefore, the provisions in 10 CFR 

50.55a(g)(6)(ii)(D)(6) are adopted 

without change in the final rule. 

9g. General Public Comments on 10 CFR 

50.55a(g)(6)(ii)(D) 

Two commenters (comment numbers 

8 and 11) stated that Public Law, PL 

104–113, mandates that national 

consensus standards be used by Federal 

agencies where applicable. This 


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includes the use of ASME codes and 

standards. Because the consensus 

process used to develop the Code Case 

specifically considered the NRC 

comments (i.e., additional conditions 

being added with this rule change) and 

found them to be without technical 

merit, one commenter considered it 

inappropriate for NRC to impose 

additional conditions on the use of Code 

Case N–729–1. Therefore, the 

commenter requested that the additional 

conditions be removed from the rule 

language. Alternatively, if the additional 

conditions would not be removed from 

the rule language, the technical 

justifications for the need for these 

additional conditions should be 

included in the supplemental 

information for the final rule. 

NRC Response: 

NRC review of ASME Code Case N–

729–1 concludes that its basis implies 

that leakage is acceptable as long as 

ejection and structural integrity due to 

wastage isn’t likely to occur. All of the 

RPV head penetration and associated 

weld examinations required by the NRC 

to date, have been based on assuring an 

extremely low probability of leakage 

from these components as well as 

assuring their structural integrity. NRC’s 

position for reactor pressure vessel 

upper head inspections is that if an 

active degradation mechanism is 

present, any long term inspection plan 

should be based on assuring an 

extremely low probability of abnormal 

leakage rather than allowing leakage and 

demonstrating the acceptability of its 

consequences. Consistent with this 

position, the NRC sets the conditions 

regarding the use of ASME Code Case 

N–729–1 in order to incorporate its use, 

by reference, into the Code of Federal 

Regulations. The technical justifications 

for the need for these conditions are 

included in the public comment section 

of this rulemaking activity. 

10. 10 CFR 50.55a(g)(6)(ii)(E)—Reactor 

Coolant Pressure Boundary Visual 

Inspections 

Public Comment: 

In a letter dated June 19, 2007, 

Progress Energy stated that the ASME 

has not amended Section XI of the BPV 

Code to include Code Case N–722. 

Therefore, requiring licensees to comply 

with a Code Case that has not been 

incorporated into the ASME Code sets a 

precedence of mandatory 

implementation of a Code Case which 

has not been subject to ASME public 

review and comment during its 

development. 

NRC Response: 

The NRC recognizes that the ASME 

has not amended Section XI of the 

ASME BPV Code to include Code Case 

N–722 and that during development 

code cases may be subjected to different 

ASME public review and comment than 

Section XI. The NRC is incorporating 

Code Case N–722 in the rule to expedite 

the implementation of Code Case N–

722. The NRC is requiring expedited 

implementation of Code Case N–722 

because the NRC concluded from a 

safety perspective that these inspections 

are necessary to ensure the integrity of 

the Alloy 600/82/182 components. The 

NRC has previously incorporated code 

cases in 10 CFR 50.55a prior to the 

ASME taking action to include the code 

cases in the ASME Code. The NRC 

declines to adopt commenter’s 

suggestion. No change was made to the 

final rule as a result of this comment. 

Public Comment: 

In a letter dated June 22, 2007, 

Southern Nuclear Operating Company 

stated that the NRC does not reference 

the industry efforts, especially those 

made through the Electric Power 

Research Institute’s Materials and 

Reliability Program (MRP) to address 

the issue of bare-metal visual 

examination of Alloy 600 welds. Every 

PWR in the United States has agreed to 

the implementation of MRP–139, which 

requires an augmented program to 

perform bare-metal visual examinations 

on the large diameter Alloy-600 welds 

on a frequency that is almost identical 

to the schedule mandated in ASME 

Code Case N–722. Typically, utilities 

are given the option to assess each code 

case and determine if that code case 

should be adopted for use. By 

mandating the use of Code Case N–722, 

the NRC is, in effect, writing their own 

code and deviating from using guidance 

from an international consensus 

standard body (ASME Code 

Committees, of which the NRC is a 

participant and voting member). The 

NRC and the industry have been 

working on this issue, and industry 

programs are in place to cover these 

examinations. Additional time should 

be provided to allow the MRP and 

ASME to develop the necessary 

enhancements. 

NRC Response: 

The MRP–139 report referenced by 

the commenter is an industry guidance 

document which includes guidance on 

bare-metal visual examinations of Alloy 

82/182 butt welds. Because MRP–139 is 

written as inspection guidance, MRP–

139 is not suitable to be incorporated by 

reference in 10 CFR 50.55a. In addition, 

the MRP has not issued inspection 

guidelines for partial-penetration 

welded components with Alloy 600/82/ 

182 materials. The NRC finds Code Case 

N–722 with conditions is suitable to be 

incorporated by reference in the final 

rule. Given the safety significance of 

these inspections, the NRC concluded 

that the reactor coolant pressure 

boundary visual inspections of 10 CFR 

50.55a(g)(6)(ii)(E) are necessary to 

ensure that the appropriate safetysignificant visual inspections are 

performed. 

The NRC recognizes that the ASME is 

an international, consensus standard 

body, and that the ASME Code provides 

necessary requirements for the design 

and inspection of nuclear power plant 

components. Therefore, the NRC has 

incorporated by reference in 10 CFR 

50.55a certain editions and addenda of 

Section III and XI of the ASME BPV 

Code. However, in certain cases, such as 

when an active degradation mechanism 

is affecting the integrity of pressure 

boundary components, the NRC needs 

to take regulatory actions to ensure 

safety and protect the public health and 

safety. As mandated by the Atomic 

Energy Act of 1954, as amended, and 

the Energy Reorganization Act of 1974, 

the NRC has the statutory authority and 

responsibility to enact regulations 

through the rulemaking process as 

necessary to ensure safety. 

The NRC declines to adopt 

commenter’s suggestion. No change was 

made to the final rule as a result of this 

comment. 

Public Comment: 

In a letter dated June 20, 2007, 

Arizona Public Service Company stated 

that 10 CFR 50.55a(g)(6)(ii)(E)(1) 

exempts Alloy 600/82/182 materials 

that have been mitigated by weld 

overlay or stress improvement from the 

inspection requirements of Code Case 

N–722. The commenter recommended 

that nozzles and penetrations that have 

been mitigated by half-nozzle 

replacement or Alloy 690/52/152 weld 

pads should also be exempted from the 

requirements of Code Case N–722. 

NRC Response: 

Code Case N–722, as implemented by 

10 CFR 50.55a(g)(6)(ii)(E), applies to 

examination of pressure retaining 

partial or full penetration welds in Class 

1 components fabricated with Alloy 

600/82/182 material in PWRs. The 

requirements of Code Case N–722, as 

implemented by 10 CFR 

50.55a(g)(6)(ii)(E), applies to nozzles 

and penetrations that have Alloy 600/ 

82/182 materials that form the pressure 

boundary. This requirement is clear 

from the title and wording of Code Case 

N–722. Note the clarification in the 

preceding sentences applies even 

though Alloy 600/82/182 materials may 

not be entirely removed from the 

component, provided that pressure 

retaining penetrations and welds no 


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longer contain Alloy 600, Alloy 82, or 

Alloy 182 materials. In addition, 10 CFR 

50.55a(g)(6)(ii)(E)(1) is revised in the 

final rule. 

Public Comment: 

In a letter dated June 20, 2007, Jack 

Spanner of Electric Power Research 

Institute stated that with respect to 10 

CFR 50.55a(g)(6)(ii)(E)(2), it should be 

sufficient to demonstrate the ability to 

characterize location, orientation and 

length of cracks with calibration blocks 

or mockups containing a notch in the 

axial and circumferential orientation. 

NRC Response: 

The requirements of paragraph 

(g)(6)(ii)(E)(2) state only that additional 

actions must be taken to characterize the 

location, orientation, and length of 

cracks. The comment does not provide 

sufficient information for the NRC to 

respond regarding the adequacy of 

calibration blocks or mockups to meet 

these requirements. Therefore, the NRC 

declines to adopt the commenter’s 

suggestion. No change was made to the 

final rule as a result of this comment. 

Public Comment: 

In a letter dated June 20, 2007, 

Arizona Public Service Company 

recommended that the term ‘‘Non-visual 

NDE’’ used in paragraph (g)(6)(ii)(E)(3) 

be changed to ‘‘surface’’ or ‘‘volumetric’’

examination. 

NRC Response: 

The ASME Code, Section XI, 

paragraph IWA–2200 states that ‘‘three 

types of examinations used during 

inservice inspection are defined as 

visual, surface, and volumetric.’’ It is 

clear from this Code definition that nonvisual examination refers to either 

surface or volumetric examination. The 

NRC declines to adopt the commenter’s 

suggestion. No change was made to the 

final rule as a result of this comment. 

Public Comment: 

In a letter dated June 20, 2007, 

Arizona Public Service Company stated 

that paragraph (g)(6)(ii)(E)(4) imposes 

the rule of Appendix VIII of the ASME 

Code, Section XI, to components where 

qualification may not have been 

performed (possibly due to size and 

thickness). Therefore, the commenter 

recommended that because the 

component causing the implementation 

of this paragraph is leaking, the NDE 

method and techniques utilized to 

characterize the leak in paragraph 

(g)(6)(ii)(E)(2) should be sufficient 

qualification. 

NRC Response: 

The commentor believes that 

paragraph (g)(6)(ii)(E)(4) is unnecessary 

and suggests that the NDE method and 

techniques utilized to characterize the 

leak in (g)(6)(ii)(E)(2) be sufficient [NDE] 

qualification. The NRC disagrees with 

the commentor’s suggestion. Paragraph 

(g)(6)(ii)(E)(2) requires that when 

leakage is detected in a component, 

additional action (e.g., non-visual 

examination) must be performed to 

characterize the location, orientation, 

and length of cracks that cause the 

leakage. Paragraph (g)(6)(ii)(E)(2) does 

not provide specific qualification for 

NDE. The intent of Paragraph 

(g)(6)(ii)(E)(2) is to provide a general 

requirement for non-visual 

examinations to be performed should 

leakage be detected. The NDE method 

and techniques utilized to characterize 

the leak in paragraph (g)(6)(ii)(E)(2) are 

visual examinations which cannot 

characterize flaw sizes. 

Paragraph (g)(6)(ii)(E)(4) requires that 

the ultrasonic examination be 

performed using the appropriate 

supplement of Section XI, Appendix 

VIII of the ASME Code. The intent of 

paragraph (g)(6)(ii)(E)(4) is to provide 

specific NDE qualification requirements 

for ultrasonic examination for Alloy 

600/82/182 butt welds so that the 

requirements of paragraphs 

(g)(6)(ii)(E)(2) or (g)(6)(ii)(E)(3) can be 

satisfied. 

This position is consistent with other 

provisions of 10 CFR 50.55a in that 

ultrasonic examination of butt welds 

must be qualified in accordance with 

the appropriate supplement of Section 

XI, Appendix VIII of the ASME Code. 

Therefore, the NRC declines to adopt 

the commenter’s suggestion. No change 

was made to the final rule as a result of 

this comment. 

Public Comment: 

After the public comment period 

closed, the NRC received an additional 

comment from Florida Power and Light 

Company via a phone call on July 8, 

2008, regarding the schedule for 

implementing the initial inspections 

under Code Case N–722 as required by 

10 CFR 50.55a(g)(6)(ii)(E), Reactor 

coolant pressure boundary visual 

inspections. The commenter pointed out 

that Code Case N–722 specifies 

frequency of examination for each part 

to be examined but does not specify 

when the initial inspections shall be 

performed. The commenter 

recommended that the schedule for the 

initial inspections be specified in the 

rule. 

NRC Response: 

The NRC agrees with the commenter 

that the schedule for the initial 

inspections is not specified in Code 

Case N–722 nor is it specified in a NRCproposed condition applicable to this 

Code Case. Code Case N–722 contains 

three different inspection intervals: 

inspections to be conducted every other 

refueling outage, each refueling outage, 

and once per interval. The NRC has 

specified the following initial 

inspection requirements in a new 

footnote to the new paragraph. 

For inspections to be conducted every 

refueling outage and inspections 

conducted every other refueling outage, 

the initial inspection shall be performed 

at the next refueling outage after January 

1, 2009. For inspections to be conducted 

once per interval, the inspections shall 

begin in the interval in effect on January 

1, 2009, and shall be prorated over the 

remaining periods and refueling outages 

in this interval. For inspections to be 

conducted once per interval, if the 

current interval ends prior to January 1, 

2009, the initial inspection shall be 

performed at the first refueling outage 

after January 1, 2009. These initial 

inspection schedules are believed to be 

reasonable since, in general, the 

inspections are straightforward to 

perform and licensees have been aware 

for over two years of the NRC intent to 

incorporate Code Case N–722 in the 

regulations during which to plan the 

inspections. 

III. Section-by-Section Analysis 

ASME BPV Code, Section III 

10 CFR 50.55a(b)(1) 

The final rule revises § 50.55a(b)(1) in 

the current regulation to incorporate by 

reference the 2004 Edition of Section III, 

Division 1, of the ASME BPV Code into 

10 CFR 50.55a. This paragraph requires 

new applicants for a nuclear power 

plant who submit an application for a 

construction permit under 10 CFR part 

50 after the effective date of this rule use 

the 2004 Edition of Section III, Division 

1 of the ASME BPV Code for the design 

and construction of the reactor coolant 

pressure boundary and Quality Group B 

and C components. This paragraph also 

requires that existing modifications and 

limitations for weld leg dimensions, 

independence of inspection and 

subsection NH in §§ 50.55a(b)(1)(ii), 

50.55a(b)(1)(v), and 50.55a(b)(1)(vi), 

respectively, apply to the 2004 Edition 

of Section III, Division 1 of the ASME 

BPV Code. The NRC is not adopting any 

additional limitations with respect to 

the 2004 Edition of Section III. 

10 CFR 50.55a(b)(1)(iii)—Seismic 

Design of Piping 

As discussed in Section II of this 

document, applicants or licensees may 

use Articles NB–3200, NB–3600, NC–

3600, and ND–3600 for seismic design 

of piping up to and including the 1993 

Addenda, subject to the limitation 

specified in paragraph (b)(1)(ii) of this 

section. Applicants or licensees may not 

use these Articles for seismic design of 


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piping in the 1994 Addenda through the 

latest edition and addenda incorporated 

by reference in paragraph (b)(1) of this 

section. The final rule revises 

50.55a(b)(1)(iii) in the current 10 CFR 

50.55a to clarify the current limitation 

regarding seismic design. Current 

§ 50.55a(b)(1)(iii) states that applicants 

or licensees may use Articles NB–3200, 

NB–3600, NC–3600, and ND–3600 for 

seismic design. However, the rules in 

Article NB–3200 of Section III of the 

ASME BPV Code contain criteria 

applicable to the seismic design of 

components other than piping systems. 

The NRC revises § 50.55a(b)(1)(iii) to 

clarify that the limitation only applies to 

the seismic design of piping. 

ASME BPV Code, Section XI 

The final rule revises § 50.55a(b)(2) to 

incorporate by reference the 2004 

Edition of the ASME BPV Code, Section 

XI, Division 1, subject to the 

modifications and limitations discussed 

in the following paragraphs: 

10 CFR 50.55a(b)(2)(xi)—Class 1 Piping 

Paragraph 50.55a(b)(2)(xi) states that 

‘‘licensees may not apply IWB–1220, 

‘‘Components Exempt from 

Examination,’’ of Section XI, 1989 

Addenda through the latest edition and 

addenda incorporated by reference in 

paragraph (b)(2) of this section, and 

shall apply IWB–1220, 1989 Edition.’’

Subarticle IWB–1220 of the 1989 

Edition of the ASME BPV Code, Section 

XI, exempts certain components (such 

as small bore piping) from the 

volumetric and surface examinations. 

However, welds or portions of welds 

that are inaccessible due to being 

encased in concrete, buried 

underground, located inside a 

penetration, or encapsulated by guard 

pipe were included in components for 

exemption from examination and 

incorporated in the edition and addenda 

of the ASME BPV Code, Section XI, after 

the 1989 Edition. The NRC previously 

did not agree with the incorporation of 

these types of welds for exemption from 

examination because the NRC believed 

that these welds should be examined to 

monitor their structural integrity. 

Therefore, the NRC prohibited the use of 

1989 addenda through the latest 

editions and addenda of the ASME BPV 

Code, Section XI, regarding the 

application of IWB–1220 in 10 CFR 

50.55a(b)(2)(xi) (64 FR 51394; 

September 22, 1999). 

The revision to the regulation 

removes 10 CFR 50.55a(b)(2)(xi), 

thereby permitting the use of ASME 

BPV Code, Section XI, IWB–1220 of any 

edition or addenda of ASME BPV Code, 

Section XI, incorporated by reference in 

10 CFR 50.55a. The condition placed 

upon Section XI, IWB–1220 in 10 CFR 

50.55a(b)(2)(xi) is no longer necessary 

because of the following: 

1. Licensees can select an alternate 

weld for inspection that does not have 

limitations. 

2. Licensees have committed to 

perform augmented inspections of break 

exclusion zone (BEZ) welds which are 

located in inaccessible areas such as 

containment penetrations or 

encapsulated by guard pipe to the extent 

practical under the BEZ criteria. 

3. Boiling water reactor (BWR) 

licensees have followed the provisions 

of Generic Letter 88–01, ‘‘NRC Position 

on IGSCC [intergranular stress corrosion 

cracking] in BWR Austenitic Stainless 

Steel Piping,’’ and the associated NRC 

report, NUREG–0313, ‘‘Technical Report 

on Material Selection and Process 

Guidelines for BWR Coolant Pressure 

Boundary Piping,’’ and the provisions of 

the BEZ criteria (Reference: Branch 

Technical Position MEB 3–1 attached to 

Standard Review Plan 3.6.2) apply to 

the examination of the welds such as 

those that are located inside 

containment penetrations or 

encapsulated by guard pipe. 

4. Licensees of plants whose 

construction permits were issued after 

January 1, 1971, are required to have 

ASME Class 1 and Class 2 components 

designed and provided with access to 

enable the performance of ISIs, and the 

removal of the limitation on the use of 

IWB–1220(d) would not permit welds to 

be located in reactor coolant pressure 

boundary components (including Class 

1 components permitted to be designed 

to Class 2 rules) that are encased in 

concrete, buried underground, located 

inside a penetration, or encapsulated by 

guard pipe. 

10 CFR 50.55a(b)(2)(xiii)—Mechanical 

Clamping Devices 

Paragraph 50.55a(b)(2)(xiii) is 

removed from the regulation. This 

paragraph permitted licensees to use the 

provisions of Code Case N–523–1, 

‘‘Mechanical Clamping Devices for Class 

2 and 3 Piping.’’ Instead, Code Case N–

523–2 provides updated requirements to 

those of Code Case N–523–1, has been 

accepted in Regulatory Guide (RG) 

1.147, Revision 15, ‘‘Inservice 

Inspection Code Case Acceptability, 

ASME BPV Code, Section XI, Division 

1,’’ and Revision 15 is incorporated by 

reference into 10 CFR 50.55a(g)(4)(i) and 

10 CFR 50.55a(g)(4)(ii). Therefore, 10 

CFR 50.55a(b)(2)(xiii) no longer serves 

any useful purpose and is removed. 

10 CFR 50.55a(b)(2)(xv)—Appendix VIII 

Specimen Set and Qualification 

Requirements 

Paragraph 50.55a(b)(2)(xv) in the 

current 10 CFR 50.55a regulation 

specifies provisions that may be used to 

modify implementation of Appendix 

VIII of Section XI, 1995 Edition through 

the 2001 Edition of the ASME BPV Code 

with regard to ultrasonic examinations 

of piping systems. The change specifies 

that licensees who have been approved 

by the NRC to use later editions and 

addenda than the 2001 Edition of the 

ASME BPV Code shall use the 2001 

Edition of Appendix VIII. Licensees 

cannot use Appendix VIII to the 

editions and addenda of the ASME Code 

Section XI that are later than the 

Appendix VIII to 2001 Edition. 

10 CFR 50.55a(b)(2)(xx)—System 

Leakage Tests 

10 CFR 50.55a(b)(2)(xx) in the current 

50.55a regulation requires certain hold 

time when performing system leakage 

tests in accordance with IWA–5213(a) of 

the 1997 through 2002 addenda of the 

ASME Code Section XI. Since the 

publication of the current 10 CFR 

50.55a, the NRC has noticed an NDE 

issue that involves the system leakage 

tests when performed in accordance 

with IWA–4540(a). 10 CFR 

50.55a(b)(2)(xx) is revised to address the 

NDE issue. The requirements in current 

10 CFR 50.55a(b)(2)(xx) are not changed. 

The revised 10 CFR 50.55a(b)(2)(xx) 

provides new requirements. The 

revision requires, as part of repair and 

replacement activities (by welding or 

brazing under the 2003 Addenda 

through the latest edition and addenda 

incorporated by reference in 10 CFR 

50.55a(b)(2)), that NDE be performed in 

accordance with subarticle IWA–

4540(a)(2) of the 2002 Addenda of the 

ASME BPV Code, Section XI, after a 

system leakage test is performed per 

subarticle IWA–4540(a)(2) of the 2003 

Addenda through later editions and 

addenda of the ASME BPV Code, 

Section XI. This provision requires that 

after repair or replacement activities (1) 

the NDE method and acceptance criteria 

of the 1992 Edition, or later, of Section 

III be performed and met prior to 

returning the system to service, and that 

(2) a system leakage test be performed 

in accordance with IWA–5000 prior to, 

or as part of, returning the system to 

service. 

10 CFR 50.55a(b)(2)(xxi)(A)—Table 

IWB–2500–1 Examination Requirements 

Paragraph 10 CFR 50.55a(b)(2)(xxi)(A) 

in the current 50.55a regulation allows 

the use of the visual examination with 


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enhanced magnification in lieu of an 

ultrasonic examination. Because of the 

latest development in visual 

examination requirements in the ASME 

Code, Paragraph 10 CFR 

50.55a(b)(2)(xxi)(A) is revised to be 

consistent with the condition for Code 

Case N–648–1, ‘‘Alternative 

Requirements for Inner Radius 

Examination of Class I Reactor Vessel 

Nozzles, Section XI, Division 1.’’ in RG 

1.147, Revision 15, which requires the 

assumption of a limiting flaw aspect 

ratio when using the allowable flaw 

length criteria in Table IWB–3512–1 

during an enhanced visual examination. 

The revision states ‘‘The provisions of 

Table IWB–2500–1, Examination 

Category B–D, Full Penetration Welded 

Nozzles in Vessels, Items B3.40 and 

B3.60 (Inspection Program A) and Items 

B3.120 and B3.140 (Inspection Program 

B) in the 1998 Edition must be applied 

when using the 1999 Addenda through 

the latest edition and addenda 

incorporated by reference in paragraph 

(b)(2) of this section. A visual 

examination with magnification that has 

a resolution sensitivity to detect a 1-mil 

width wire or crack, utilizing the 

allowable flaw length criteria in Table 

IWB–3512–1, 1997 Addenda through 

the latest edition and addenda 

incorporated by reference in paragraph 

(b)(2) of this section, with a limiting 

assumption on the flaw aspect ratio (i.e., 

a/l=0.5), may be performed instead of an 

ultrasonic examination.’’ The limitation 

on the flaw aspect ratio is needed 

because visual examination cannot 

determine the depth of cracks. A visual 

examination requirement may be 

applied only when a limiting flaw 

aspect ratio of 0.5 is assumed. A flaw 

aspect ratio of less than 0.5 would not 

be conservative. As shown in Table 

IWB–3512–1, there are no flaw aspect 

ratios higher than 0.5. Therefore, 

assuming a flaw aspect ratio of 0.5 is 

appropriate. 

10 CFR 50.55a(g)(6)(ii)(A)—Augmented 

Examination of Reactor Vessel 

Paragraph 50.55a(g)(6)(ii)(A) is 

removed from the regulation. This 

paragraph required a one-time, 

augmented ISI program for those 

systems and components the 

Commission determined that added 

assurance of structural reliability was 

necessary. Paragraph 50.55a(g)(6)(ii)(A) 

was incorporated in the regulations in 

1992 to require all current licensees to 

conduct a one-time, expedited 

examination of reactor vessel shell 

welds. Examination requirements were 

specified in item B1.10, ‘‘Shell Welds,’’

of Examination Category B–A, ‘‘Pressure 

Retaining Welds in Reactor Vessel,’’ in 

Table IWB–2500–1, ‘‘Examination 

Categories’’ of the 1989 Edition of the 

ASME BPV Code, Section XI, Division 

1. Because all the licensees have 

completed the subject augmented 

examination of the reactor vessel shell 

welds, the requirements in 10 CFR 

50.55a(g)(6)(ii)(A) and associated 

subparagraphs are no longer needed. 

Future licensees need not conduct this 

augmented examination, because new 

Code provisions should adequately 

address the degradation to which the 

augmented examination was directed. 

10 CFR 50.55a(g)(6)(ii)(D)—Reactor 

Vessel Head Inspections 

On September 30, 2002, the DavisBesse Lessons Learned Task Force 

(LLTF) issued a report containing 51 

recommendations for actions that the 

NRC should take to address areas that 

the LLTF considered contributors to the 

Davis-Besse event. On November 26, 

2002, the senior NRC management 

review team endorsed all but two of the 

task force’s recommendations. One 

endorsed high-priority recommendation 

was the following: 

The NRC should encourage American 

Society of Mechanical Engineers Boiler and 

Pressure Vessel Code (ASME Code) 

requirement changes for bare metal 

inspections of nickel based alloy nozzles for 

which the code does not require the removal 

of insulation for inspections. The NRC 

should also encourage ASME Code 

requirement changes for the conduct of nonvisual non-destructive examination (NDE) 

inspections of VHP [vessel head penetration] 

nozzles. Alternatively, the NRC should revise 

Title 10 Code of Federal Regulations (10 CFR) 

Part 50.55a to address these areas. 

Section XI of the ASME Code, which 

is incorporated by reference into NRC 

regulations by 10 CFR 50.55a, ‘‘Codes 

and standards,’’ currently specifies that 

inspections of the reactor pressure 

vessel (RPV) head need only include a 

visual check for leakage on the insulated 

surface or surrounding area. Experience 

has shown that these inspections may 

not detect small amounts of leakage 

from an RPV head penetration with 

cracks extending through the nozzle or 

the J-groove weld. Such leakage can 

create an environment that leads to 

circumferential cracks in RPV head 

penetration nozzles and/or corrosion of 

the RPV head. 

The NRC issued Order EA–03–009, 

‘‘Interim Inspection Requirements for 

Reactor Pressure Vessel Heads at 

Pressurized Water Reactors,’’ dated 

February 11, 2003, which modified 

licensees’ licenses to require specific 

inspections of the reactor pressure 

vessel head and associated penetration 

nozzles at pressurized water reactors. In 

September 2003, industry 

representatives through the Materials 

Reliability Program provided industry 

input to support industry alternative 

inspection programs through various 

public meetings and MRP–95, 

‘‘Materials Reliability Program: Generic 

Evaluation of Examination Coverage 

Requirements for the Reactor Pressure 

VHP Nozzles, (ML032740424).’’ In 

response to internal review and 

stakeholder input, the NRC issued First 

Revised Order EA–03–009, February 20, 

2004 (Order), which refined the 

inspection requirements of NRC Order 

EA–03–009 by taking into account 

lessons learned from inspections 

performed from February 2003 to 

January 2004. 

On July 7, 2004, after an assessment 

which concluded that ASME Code 

requirement revisions would not be 

complete in 2004, the NRC issued a 

Commission Paper (SECY–04–0115) 

requesting Commission approval of a 

rulemaking plan to incorporate into 10 

CFR 50.55a the RPV head and 

associated head penetration inspection 

requirements contained in the Order. 

The Commission, in a Staff 

Requirements Memorandum, dated 

August 6, 2004, approved an alternative 

option to evaluate the RPV inspection 

requirements of an upcoming ASME 

Code Case or revision of the ASME Code 

for incorporation into 10 CFR 50.55a. 

In March 2006, the ASME approved 

Code Case N–729–1, Alternative 

Examination Requirements for PWR 

Reactor Vessel Upper Heads With 

Nozzles Having Pressure-Retaining 

Partial-Penetration Welds, which 

provides an alternative long-term 

inspection program for RPV upper 

heads. The NRC participated in ASME 

Code development and approval of N–

729–1. The NRC has reviewed the final 

version of Code Case N–729–1, and with 

conditions, finds it provides reasonable 

assurance of public health and safety 

from failure of the reactor pressure 

vessel upper head and penetration 

nozzles. Therefore, the NRC is pursuing 

this rulemaking activity to incorporate 

by reference the inspection 

requirements of Code Case N–729–1, as 

conditioned, into 10 CFR 50.55a. 

The experience of the Davis-Besse 

RPV head degradation and the discovery 

of leaks and nozzle cracking at other 

plants over the past seven years 

reinforce the need for effective 

regulatory required inspections of the 

RPV head and penetration nozzles. The 

absence of an effective inspection 

regime could, over time, result in 

unacceptable circumferential cracks in 

RPV head penetration nozzles or in the 

degradation of the RPV head by 


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corrosion from leaks in the reactor 

coolant pressure boundary. These 

degradation mechanisms increase the 

probability of a loss of reactor coolant 

pressure boundary event through 

ejection of a nozzle or other rupture of 

the RPV head. The result of this 

rulemaking would be the establishment 

of inspection requirements that result in 

an extremely low probability of 

abnormal leakage, of rapidly 

propagating failure and of gross rupture 

of the reactor pressure vessel head and 

penetration nozzles. 

The Code Case N–729–1 inspection 

plan for RPV upper heads with Alloy 

600/182/82 penetration nozzles requires 

periodic bare metal visual (BMV) 

examinations and periodic nonvisual 

examinations using ultrasonic testing 

(UT), eddy current testing (ET), or dye 

penetrant testing of the penetration 

nozzle base metal. BMV examinations 

are performed in order to identify 

primary coolant leakage based on the 

presence of boric acid deposit 

accumulations. Nonvisual examinations 

are performed in order to identify flaws 

which could lead to leakage or failure of 

the penetration nozzle. 

These same inspections are required 

to be performed for RPV upper heads 

with Alloy 690/152/52 penetration 

nozzles, but the frequency of inspection 

is greatly reduced. This reduction is due 

to the enhanced resistance these 

materials have demonstrated against 

PWSCC. 

Paragraph 50.55a(g)(6)(ii)(D) is added 

to the regulation to require licensees to 

comply with the reactor vessel head 

inspection requirements of ASME Code 

Case N–729–1, subject to conditions, by 

December 31, 2008. Compliance to Code 

Case N–729–1; with conditions 

regarding inspection frequency, 

examination coverage, qualification of 

ultrasonic examination, and reinspection intervals; would be 

equivalent to complying with NRC 

Order EA–03–009, dated February 11, 

2003, and First Revised Order EA–03–

009, dated February 20, 2004. Thus, 

once a licensee implements Code Case 

N–729–1, with conditions, the First 

Revised NRC Order EA–03–009 no 

longer applies to that licensee and is 

deemed to be withdrawn. This allows 

licensees to transfer from the Order 

requirements to the requirements of 10 

CFR 50.55a(g)(6)(ii)(D). 

Footnote 10 to 10 CFR 50.55a(b)(2) is 

removed because Code Case N–729–1, 

as conditioned, replaces the 

requirements of the NRC Order EA–03–

009 cited in that footnote. 

10 CFR 50.55a(g)(6)(ii)(E)—Reactor 

Coolant Pressure Boundary Visual 

Inspections 

A new paragraph 10 CFR 

50.55a(g)(6)(ii)(E) is added to require all 

current and future licensees to apply 

ASME Section XI, Code Case N–722, 

with conditions. Code Case N–722 

provides requirements for bare metal 

visual examination of full and partial 

penetration welds in Class 1 

components that are fabricated with 

Alloy 600/82/182 material. Surfaces 

required to be examined by the bare 

metal visual method have to be 

unobstructed by debris, paint, 

insulation or other sources of 

interference. 10 CFR 50.55a(g)(6)(ii)(E) 

requires the use of N–722 plus four 

additional conditions. Condition (1) 

requires that PWR licensees implement 

N–722 except for those welds that have 

been mitigated by weld overlay or stress 

improvements. Condition (2) requires 

that if leakage occurs from a component, 

licensees take additional actions to 

characterize the orientation of the crack 

that caused the leakage. Condition (3) 

requires that if the crack that leads to 

leakage is circumferentially oriented 

and potentially the result of primary 

water stress-corrosion cracking, 

licensees perform non-visual sample 

inspections of the population of the 

components. Condition (4) requires that 

the ultrasonic examinations of the butt 

welds as required by Condition (2) and 

(3) follow the appropriate supplement of 

Appendix VIII of the ASME Code, 

Section XI. 

The visual examinations specified in 

Code Case N–722 are additional 

requirements beyond the current NDE 

requirements of Table IWB–2500–1 in 

the ASME Code, Section XI. The 

application of ASME Code Case N–722 

is necessary because current inspections 

are inadequate and the safety 

consequences can be significant should 

the components fail due to cracking. 

NRC’s determination that existing 

inspections of the reactor coolant 

pressure boundary (RCPB) are 

inadequate is based upon the 

degradation of RPV head penetration 

nozzles at Davis-Besse and the 

discovery of leaks and cracking at other 

plants, such as Oconee and Arkansas 

Nuclear One Unit 1. The absence of an 

effective inspection regime could, over 

time, result in unacceptable 

circumferential cracking or the 

degradation of reactor coolant system 

(RCS) components by corrosion from 

leaks in the RCPB. These degradation 

mechanisms increase the probability of 

a loss-of-coolant accident. The 

inspections required by the 2004 

Edition of the ASME BPV Code, Section 

XI, are inadequate because Examination 

Category B–P, ‘‘All Pressure Retaining 

Components,’’ of Table IWB–2500–1, 

only requires a visual examination of 

the reactor vessel with the insulation in 

place during a system leakage test each 

refueling outage. Visual inspections may 

not detect gradual leakage as confirmed 

by industry experience. 

Both the NRC and the industry took 

short-term actions to address PWSCC in 

the RCPB because of limitations of the 

ASME BPV Code inspection programs to 

address PWSCC in the RCPB. In 

addition to issuing bulletins, the NRC 

issued Order EA–03–009 and First 

Revised Order EA–03–009 to quickly 

establish interim inspection 

requirements for RPV upper heads at 

PWRs. However, these measures 

addressed the issue only temporarily, 

and for specific locations. The industry 

also responded with compensatory 

measures (e.g., by specifying that a onetime, bare-metal visual inspection of all 

RCS nickel-based alloy components and 

weld locations be performed within two 

refueling outages). However, these were 

only short-term measures. 

The ASME also took actions to 

address PWSCC. An ASME task group 

concluded that more rigorous 

inspections than those currently 

provided by the ASME BPV Code were 

needed in the areas most susceptible to 

PWSCC. The task group developed 

ASME Code Case N–722 to enhance the 

current ASME BPV Code requirements 

for detection of leakage and corrosion in 

the components considered to be 

susceptible to PWSCC. The Code Case 

specifies bare-metal visual examinations 

for all RCS pressure retaining 

components fabricated from Alloy 600/ 

82/182 materials. This Code Case was 

approved by ASME in July 2005 and 

was published in Supplement 6 to the 

2004 Code Cases. However, the Code 

Case is not mandatory for industry to 

follow. The Code Case improves upon 

existing ASME BPV Code inspection 

requirements, because it specifies bare 

metal visual examinations. 

Beyond the bare metal visual 

inspection requirements and 

frequencies of inspections, ASME Code 

Case N–722 is relatively limited in 

scope. The NRC is requiring non-visual 

inspection for items where leakage is 

identified in Class 1 components. The 

additional non-visual NDE is required to 

determine whether circumferential 

cracking is present in the flawed 

material and if multiple circumferential 

flaws have initiated. Leakage detected 

by visual examination only identifies 

that a flaw exists, and is not able to 

characterize flaw orientations and 


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locations. The NRC is requiring NDE 

scope expansion once a circumferential 

flaw is identified in these components 

because once flaws are found, favorable 

conditions must be assumed to exist for 

additional flaws to develop in other 

similar components in similar 

environments. Circumferential cracking 

has occurred, and is a particularly 

serious safety concern because it could, 

if undetected by NDE, lead to a 

complete severing of the piping and a 

loss-of-coolant accident. 

Therefore, the NRC is requiring the 

application of Code Case N–722 with 

additional conditions. The conditions 

require additional NDE when leakage is 

detected and expansion of the sample 

size if a circumferential PWSCC flaw is 

found. Operating experience has shown 

that bare metal visual inspections alone 

are not sufficient and that NDE is 

necessary in order to detect cracking. 

The requirements for the schedule for 

conducting the initial inspections are 

specified in a new footnote to the new 

paragraph. 

ASME OM Code 

The revision to § 50.55a(b)(3) 

incorporates by reference the 2004 

Edition of the ASME OM Code subject 

to no new modifications or limitations. 

Paragraph (b)(3)(iv)(D) is revised to be 

less specific with regard to paragraph 

references in subsection ISTC [Inservice 

testing, the Code for Operation and 

Maintenance of Nuclear Power Plants] 

to eliminate inconsistencies in 

paragraph numbering. This is 

considered to be an editorial change that 

does not affect the intent or 

implementation of the current 

modification regarding the 

discontinuance of Appendix II 

condition monitoring programs of check 

valves. 

IV. Generic Aging Lessons Learned 

Report 

In September 2005, the NRC issued, 

‘‘Generic Aging Lessons Learned (GALL) 

Report,’’ NUREG–1801, Volumes 1 and 

2, Revision 1, for applicants to use in 

preparing their license renewal 

applications. The GALL report evaluates 

existing programs and documents the 

bases for determining when existing 

programs are adequate without change 

or augmentation for license renewal. 

Section XI, Division 1, of the ASME 

BPV Code is one of the existing 

programs in the GALL report that is 

evaluated as an aging management 

program (AMP) for license renewal. 

Subsections IWB, IWC, IWD, IWE, IWF, 

and IWL of the 2001 Edition up to and 

including the 2003 Addenda of Section 

XI of the ASME BPV Code for ISI were 

evaluated in the GALL report and the 

conclusions in the GALL report are 

valid for this edition and addenda. 

In the GALL report, Sections XI.M1, 

‘‘ASME Section XI Inservice Inspection, 

Subsections IWB, IWC, and IWD,’’

XI.S1, ‘‘ASME Section XI, Subsection 

IWE,’’ XI.S2, ‘‘ASME Section XI, 

Subsection IWL,’’ and XI.S3, ‘‘ASME 

Section XI, Subsection IWF,’’ describe 

the evaluation and technical bases for 

determining the adequacy of 

Subsections IWB, IWC, IWD, IWE, IWF, 

and IWL, respectively. In addition, 

many other AMPs in the GALL report 

rely in part, but to a lesser degree, on 

the requirements in the ASME BPV 

Code, Section XI. 

The NRC has evaluated Subsections 

IWB, IWC, IWD, IWE, IWF, and IWL of 

Section XI of the ASME BPV Code, 2004 

Edition as part of the § 50.55a 

amendment process to incorporate by 

reference the 2004 Edition of the ASME 

BPV Code to determine if the 

conclusions of the GALL report also 

apply to AMPs that rely upon the ASME 

BPV Code edition that is incorporated 

by reference into § 50.55a by this final 

rule. The NRC finds that the 2004 

Edition of Sections III and XI of the 

ASME BPV Code, as modified and 

limited in this final rule, are acceptable 

and the conclusions of the GALL report 

remain valid. Accordingly, an applicant 

may use Subsections IWB, IWC, IWD, 

IWE, IWF, and IWL of Section XI of the 

2004 Edition of the ASME BPV Code, as 

modified and limited in this final rule, 

as acceptable alternatives to the 

requirements of the 2001 Edition up to 

and including the 2003 Addenda of the 

ASME BPV Code, Section XI, referenced 

in the GALL AMPs in its plant-specific 

license renewal application. Similarly, a 

licensee approved for license renewal 

that relied on the GALL AMPs may use 

Subsections IWB, IWC, IWD, IWE, IWF, 

and IWL of Section XI of the 2004 

Edition of the ASME BPV Code as 

acceptable alternatives to the AMPs 

described in the GALL report. 

However, a licensee must assess and 

follow applicable NRC requirements 

with regard to changes to its licensing 

basis. 

The GALL report includes AMPs that 

are based on the requirements in the 

2001 Edition through the 2003 Addenda 

of Section XI of the ASME BPV Code 

but in which the AMPs may recommend 

additional augmentation of the Code 

requirements in order to achieve aging 

management for license renewal. The 

technical or regulatory aspects of the 

AMPs, for which augmentation is 

recommended, also apply when 

implementing the 2004 Edition of 

Section XI of the ASME BPV Code. A 

license renewal applicant may either 

augment its AMPs in these areas, as 

described in the GALL report, or 

propose alternatives (exceptions) for the 

NRC to review as part of a plant-specific 

program element aspect of its AMP. 

The NRC currently provides license 

renewal guidance for augmented 

inspections of PWR upper reactor vessel 

heads and their penetration nozzles in 

GALL AMP XI.M11A, ‘‘Nickel-Alloy 

Penetration Nozzles Welded to the 

Upper Reactor Vessel Closure Heads of 

Pressurized Water Reactors (PWR 

Only).’’ The current program elements 

and aging management 

recommendations in GALL AMP 

XI.M11A are based on the augmented 

inspection requirements in the First 

Revised Order EA–03–009, ‘‘Issuance of 

First Revised Order (EA–03–009) 

Establishing Interim Inspection 

Requirements for Reactor Pressure 

Vessel Heads at Pressurized Water 

Reactors.’’ For licensees that have been 

granted a renewed operating license and 

have committed to an AMP that is based 

on both conformance with GALL AMP 

XI.M11A and compliance with First 

Revised Order EA–03–009, the licensees 

may update the program elements of 

their AMP to reflect compliance with 

the new requirements in 10 CFR 

50.55a(g)(6)(ii)(D) and (E) without 

having to identify an exception to GALL 

AMP XI.M11A. For new or current 

license renewal applicants, they may 

reference conformance with GALL AMP 

XI.M11A and compliance with the new 

augmented inspection requirements in 

paragraphs 10 CFR 50.55a(g)(6)(ii)(D) 

and (E) without the need for taking an 

exception to the program elements in 

GALL AMP XI.M11A. 

V. Availability of Documents 

Document Public document room 

Electronic 

reading room ADAMS No. 

ASME BPV Code* .................................................................................................................... ........................ ........................ N/A 

ASME OM Code* ...................................................................................................................... ........................ ........................ N/A 

ASME Code Case N–722 ......................................................................................................... X ........................ ML070170676 

ASME Code Case N–729–1 ..................................................................................................... X ........................ ML070170679 


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52744 Federal Register / Vol. 73, No. 176 / Wednesday, September 10, 2008 / Rules and Regulations 

Document Public document room 

Electronic 

reading room ADAMS No. 

Regulatory Analysis .................................................................................................................. X ........................ ML081550317 

EA–03–009 ............................................................................................................................... X X ML030380470 

First Revised NRC Order EA–03–009 ..................................................................................... X X ML040220181 

GALL Report, NUREG–1801 .................................................................................................... ........................ X ML012060392 

........................ ........................ ML012060514 

........................ ........................ ML012060521 

........................ ........................ ML012060539 

Staff Requirements Memorandum dated September 10, 1999 ............................................... ........................ ........................ ML003751061 

RG 1.147, Revision 15 ............................................................................................................. X X ML072070419 

*Available on the ASME Web site. 

VI. Voluntary Consensus Standards 

The National Technology Transfer 

and Advancement Act of 1995, Public 

Law 104–113, requires agencies to use 

technical standards that are developed 

or adopted by voluntary consensus 

standards bodies unless the use of such 

a standard is inconsistent with 

applicable law or is otherwise 

impractical. Public Law 104–113 

requires Federal agencies to use 

industry consensus standards to the 

extent practical; it does not require 

Federal agencies to incorporate by 

reference a standard into the regulations 

in its entirety. The law does not prohibit 

an agency from generally adopting a 

voluntary consensus standard while 

taking exception to specific portions of 

the standard if those provisions are 

deemed to be ‘‘inconsistent with 

applicable law or otherwise 

impractical.’’ Furthermore, taking 

specific exceptions furthers the 

Congressional intent of Federal reliance 

on voluntary consensus standards 

because it allows the adoption of 

substantial portions of consensus 

standards without the need to reject the 

standards in their entirety because of 

limited provisions which are not 

acceptable to the agency. 

The NRC is amending its regulations 

to incorporate by reference a more 

recent edition of Sections III and XI of 

the ASME BPV Code and ASME OM 

Code, for construction, ISI, and 

inservice testing of nuclear power plant 

components. ASME BPV and OM Codes 

are national consensus standards 

developed by participants with broad 

and varied interests, in which all 

interested parties (including the NRC 

and licensees of nuclear power plants) 

participate. In an SRM dated September 

10, 1999, the Commission indicated its 

intent that a rulemaking identify all 

parts of an adopted voluntary consensus 

standard that are not adopted, and to 

justify not adopting such parts. The 

parts of the ASME BPV Code and OM 

Code that the NRC is not adopting; or 

is adopting with conditions, 

modifications, or limitations under 

which the Codes may be applied; are 

identified in Section III of this 

document and in the regulatory 

analysis. If the NRC did not 

conditionally accept ASME Code 

Editions and Addenda, it would 

disapprove these items entirely. The 

effect would be that licensees would 

need to submit a larger number of relief 

requests which would be an 

unnecessary additional burden for both 

the licensee and the NRC. This situation 

fits the definition of ‘‘impractical’’

under Public Law 104–113. For these 

reasons, the treatment of ASME Code 

Editions and Addenda, and conditions, 

modifications, or limitations placed on 

them in this final rule do not conflict 

with any policy on agency use of 

consensus standards specified in Office 

of Management and Budget Circular A–

119. 

VII. Finding of No Significant 

Environmental Impact: Environmental 

Assessment 

This action is in accordance with 

NRC’s policy to incorporate by reference 

in 10 CFR 50.55a new editions and 

addenda of the ASME BPV and OM 

Codes to provide updated rules for 

constructing and inspecting components 

and testing pumps, valves, and dynamic 

restraints (snubbers) in light-water 

nuclear power plants. ASME Codes are 

national voluntary consensus standards 

and are required by the National 

Technology Transfer and Advancement 

Act of 1995, Public Law 104–113, to be 

used by government agencies unless the 

use of such a standard is inconsistent 

with applicable law or otherwise 

impractical. 

NEPA requires Federal government 

agencies to study the impacts of their 

‘‘major Federal actions significantly 

affecting the quality of the human 

environment’’ and prepare detailed 

statements on the environmental 

impacts of the proposed action and 

alternatives to the proposed action (42 

U.S.C. 4332(C); NEPA § 102(C)). 

The Commission has determined 

under NEPA, as amended, and the 

Commission’s regulations in subpart A 

of 10 CFR part 51, that this rule, is not 

a major Federal action significantly 

affecting the quality of the human 

environment and, therefore, an 

environmental impact statement is not 

required. 

The rulemaking will not significantly 

increase the probability or consequences 

of accidents; no changes are being made 

in the types of effluents that may be 

released off-site; there is no increase in 

occupational exposure; and there is no 

significant increase in public radiation 

exposure. Some of the changes 

concerning ensuring the integrity of the 

RCPB would reduce the probability of 

accidents and radiological impacts on 

the public. The rulemaking does not 

involve non-radiological plant effluents 

and has no other environmental impact. 

Therefore, no significant nonradiological impacts are associated with 

the action. 

The determination of this 

environmental assessment is that there 

will be no significant off-site impact to 

the public from this action. 

VIII. Paperwork Reduction Act 

Statement 

This rule increases the burden on 

licensees to report requirements and 

maintain records for examination 

requirements in ASME BPV Code 

Section XI IWB–2500(b). The public 

burden for this information collection is 

estimated to average 3 hours every ten 

years per request. Because the burden 

for this information collection is 

insignificant, OMB clearance is not 

required. Existing requirements were 

approved by the OMB, approval number 

3150–0011. 

Public Protection Notification 

The NRC may not conduct or sponsor, 

and a person is not required to respond 

to, a request for information or an 

information collection requirement 

unless the requesting document 

displays a currently valid OMB control 

number. 


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IX. Regulatory Analysis 

The NRC has prepared a regulatory 

analysis on this final rule. The analysis 

is available for review in the NRC’s 

PDR, located in One White Flint North, 

11555 Rockville Pike, Rockville, 

Maryland. In addition, copies of the 

regulatory analysis may be obtained as 

indicated in Section V of this document. 

X. Regulatory Flexibility Certification 

In accordance with the Regulatory 

Flexibility Act of 1980, 5 U.S.C. 605(b), 

the Commission certifies that this 

amendment will not, if promulgated, 

have a significant economic impact on 

a substantial number of small entities. 

This amendment affects the licensing 

and operation of nuclear power plants. 

The companies that own these plants do 

not fall within the scope of the 

definition of small entities set forth in 

the Regulatory Flexibility Act or the 

Small Business Size Standards set forth 

in regulations issued by the Small 

Business Administration at 13 CFR part 

121. 

XI. Backfit Analysis 

The NRC’s Backfit Rule in 10 CFR 

50.109 states that the Commission shall 

require the backfitting of a facility only 

when it finds the action to be justified 

under specific standards stated in the 

rule. Section 50.109(a)(1) defines 

backfitting as the modification of or 

addition to systems, structures, 

components, or design of a facility; or 

the design approval or manufacturing 

license for a facility; or the procedures 

or organization required to design, 

construct or operate a facility; any of 

which may result from a new or 

amended provision in the Commission 

rules or the imposition of a regulatory 

staff position interpreting the 

Commission rules that is either new or 

different from a previously applicable 

NRC position after issuance of the 

construction permit or the operating 

license or the design approval. 

Section 50.55a requires nuclear power 

plant licensees to construct ASME BPV 

Code Class 1, 2, and 3 components in 

accordance with the rules provided in 

Section III, Division 1, of the ASME BPV 

Code; inspect Class 1, 2, 3, Class MC, 

and Class CC components in accordance 

with the rules provided in Section XI, 

Division 1, of the ASME BPV Code; and 

test Class 1, 2, and 3 pumps, valves, and 

dynamic restraints (snubbers) in 

accordance with the rules provided in 

the ASME OM Code. This rule 

incorporates by reference the 2004 

Edition of Section III, Division 1, of the 

ASME BPV Code; Section XI, Division 

1, of the ASME BPV Code; and the 

ASME OM Code. 

Incorporation by reference of more 

recent editions and addenda of Section 

III, Division 1, of the ASME BPV Code 

does not affect a plant that has received 

a construction permit or an operating 

license or a design that has been 

approved, because the edition and 

addenda to be used in constructing a 

plant are, by rule, determined on the 

basis of the date of the construction 

permit, and are not changed thereafter, 

except voluntarily by the licensee. Thus, 

incorporation by reference of a more 

recent edition and addenda of Section 

III, Division 1, does not constitute a 

‘‘backfitting’’ as defined in 

§ 50.109(a)(1). 

Incorporation by reference of more 

recent editions and addenda of Section 

XI, Division 1, of the ASME BPV Code 

and the ASME OM Code affect the ISI 

and IST programs of operating reactors. 

However, the Backfit Rule does not 

apply to incorporation by reference of 

later editions and addenda of the ASME 

BPV Code (Section XI) and OM Code. 

The NRC’s policy has been to 

incorporate later versions of the ASME 

Codes into its regulations. This practice 

is codified in § 50.55a which requires 

licensees to revise their ISI and IST 

programs every 120 months to the latest 

edition and addenda of Section XI of the 

ASME BPV Code and the ASME OM 

Code incorporated by reference in 

§ 50.55a that is in effect 12 months prior 

to the start of a new 120-month ISI and 

IST interval. 

Other circumstances where the NRC 

does not apply the Backfit Rule to the 

incorporation by reference of a later 

Code into the regulations are as follows: 

(1) When the NRC takes exception to 

a later ASME BPV Code or OM Code 

provision but merely retains the current 

existing requirement, prohibits the use 

of the later Code provision, limits the 

use of the later Code provision, or 

supplements the provisions in a later 

Code, the Backfit Rule does not apply 

because the NRC is not imposing new 

requirements. However, the NRC 

explains any such exceptions to the 

Code in the Statement of Considerations 

and regulatory analysis for the rule; 

(2) When an NRC exception relaxes an 

existing ASME BPV Code or OM code 

provision but does not prohibit a 

licensee from using the existing Code 

provision, the Backfit Rule does not 

apply because the NRC is not imposing 

new requirements and; 

(3) Modifications and limitations 

imposed during previous routine 

updates of § 50.55a have established a 

precedent for determining which 

modifications or limitations are backfits 

or require a backfit analysis (e.g., final 

rule dated October 1, 2004 (69 FR 

58804). The application of the backfit 

requirements to modifications and 

limitations in the current rule are 

consistent with the application of 

backfit requirements to modifications 

and limitations in previous rules. 

There are some circumstances in 

which the incorporation by reference of 

a later ASME BPV Code or OM Code 

into 10 CFR 50.55a introduces a backfit. 

In these cases, the NRC performs a 

backfit analysis or documented 

evaluation in accordance with § 50.109. 

These include the following: 

(1) When the NRC incorporates by 

reference a later provision of the ASME 

BPV Code or OM Code that takes a 

substantially different direction from 

the existing requirements, the action is 

treated as a backfit, e.g., 61 FR 41303 

(August 8, 1996). 

(2) When the NRC requires 

implementation of later ASME BPV 

Code or OM Code provision on an 

expedited basis, the action is treated as 

a backfit. This applies when 

implementation is required sooner than 

it would be required if the NRC simply 

incorporated the Code by reference 

without any expedited language, e.g., 64 

FR 51370 (September 22, 1999). 

(3) When the NRC takes an exception 

to an ASME BPV Code or OM Code 

provision and imposes a requirement 

that is substantially different from the 

existing requirement as well as 

substantially different than the later 

Code, e.g., 67 FR 60529 (September 26, 

2002). 

The backfitting discussion for the 

revisions to 10 CFR 50.55a is set forth 

as follows: 

1. Remove 10 CFR 50.55a(b)(2)(xi) 

Concerning Components Exempt From 

Examination 

This change removes an existing 

limitation on the use of 1989 Addenda 

and later editions and addenda of the 

ASME BPV Code, Section XI, regarding 

the use of subarticle IWB–1220 in the 

examinations of welds in the 

inaccessible locations. Licensees have 

either committed to perform augmented 

inspection or have followed the 

provisions of Generic Letter 88–01 and 

NUREG–0313 in examining the 

inaccessible welds. Therefore, this 

change is not considered as a backfit 

under 10 CFR 50.109. 

2. Remove 10 CFR 50.55a(b)(2)(xiii) 

Concerning the Provisions of Code Case 

N–523–1, ‘‘Mechanical Clamping 

Devices for Class 2 and 3 Piping’’

10 CFR 50.55a(b)(2)(xiii) states that 

‘‘Licensees may use the provisions of 


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Code Case N–523–1, ‘‘Mechanical 

Clamping Devices for Class 2 and 3 

Piping.’’ 10 CFR 50.55a(b)(2)(xiii) does 

not require, but provides an option for, 

licensees to use Code Case N–523–1. In 

2000, ASME updated Code Case N–523–

1 to N–523–2 without changes to 

technical requirements. Code Case N–

523–2, ‘‘Mechanical Clamping Devices 

for Class 2 and 3 Piping,’’ has been 

accepted in RG 1.147, Revision 15, 

which is incorporated by reference into 

10 CFR 50.55a(g)(4)(i) and 10 CFR 

50.55a(g)(4)(ii). Code Case N–523–2 may 

be used by licensees without requesting 

authorization. According to RG 1.147, 

Revision 15, Code Case N–523–1 has 

been superseded by Code Case N–523–

2. It is stated in RG 1.147, Revision 15, 

that ‘‘After the ASME annuls a Code 

Case and the NRC amends 10 CFR 

50.55a and this guide [RG 1.147], 

licensees may not implement that Code 

Case for the first time. However, a 

licensee who implemented the Code 

Case prior to annulment may continue 

to use that Code Case through the end 

of the present ISI interval. An annulled 

Code Case cannot be used in the 

subsequent ISI interval unless 

implemented as an approved alternative 

under 10 CFR 50.55a(a)(3) * * *’’ The 

NRC has not annulled or prohibited the 

use of Code Case N–523–1 in RG 1.147, 

Revision 15. Licensees who have used 

Code Case N–523–1 may continue to use 

it. The NRC is not imposing new 

requirements by removing 10 CFR 

50.55a(b)(2)(xiii). Therefore, the removal 

of 10 CFR 50.55a(b)(2)(xiii) is not a 

backfit. 

3. Modify 10 CFR 50.55a(b)(2)(xv) To 

Implement Appendix VIII of Section XI, 

the 1995 Edition Through the 2004 

Edition of the ASME BPV Code 

This change updates the edition of the 

ASME BPV Code in 10 CFR 

50.55a(b)(2)(xv). Therefore, is not 

considered as a backfit under 10 CFR 

50.109. 

4. Add 10 CFR 50.55a(b)(2)(xx) to 

Require NDE Provision in IWA–

4540(a)(2) of the 2002 Addenda of 

Section XI When Performing System 

Leakage Tests 

Subarticle IWA–4540(a)(2) of the 2002 

Addenda of the ASME BPV Code, 

Section XI, requires an NDE be 

performed in combination with a system 

leakage test during repair/replacement 

activities. Subarticle IWA–4540(a)(2) of 

the 2003 Addenda through later editions 

and addenda of the ASME BPV Code, 

Section XI, does not specify an NDE 

after a system leakage test. The addition 

requires, as part of repair and 

replacement activities, that a NDE be 

performed per IWA–4540(a)(2) of the 

2002 Addenda of the ASME BPV Code, 

Section XI, after a system leakage test is 

performed per subarticle IWA–

4540(a)(2) of the 2003 Addenda through 

later editions and addenda of the ASME 

BPV Code, Section XI. 

As stated previously, when the NRC 

takes exception to a later ASME BPV 

Code provision but merely retains the 

existing requirement, prohibits the use 

of the later Code provision, limits the 

use of the later Code provision, or 

supplements the provisions in a later 

Code, the Backfit Rule does not apply 

because the NRC is not imposing new 

requirements. The addition retains the 

system leakage test requirement in 

IWA–4540(a)(2) of the 2003 Addenda 

through the later editions and addenda 

of the ASME BPV Code, Section XI, but 

supplements it with the NDE of IWA–

4540(a)(2) of the 2002 Addenda of the 

Code. However, the NRC has approved 

a few licensees to use IWA–4540(a) of 

the 2003 addenda of the ASME Code, 

Section XI without imposing the NDE 

requirement in conjunction with the 

system leakage tests. Therefore, some 

licensees may currently use the 

provisions of IWA–4540(a) in the 2003 

Addenda without having to perform 

NDE. Because 10 CFR 50.55a(b)(2)(xx) 

imposes NDE requirements after these 

licensees are allowed not to perform the 

required NDE, the additional NDE 

requirements in 10 CFR 50.55a(b)(2)(xx) 

may be considered backftting under 10 

CFR 50.109(a)(1) for these licensees. 

However, the NRC believes that the NDE 

requirements are necessary for 

compliance with Commission 

requirements and/or license provisions. 

Therefore, a backfit analysis need not be 

prepared under the ‘‘compliance’’

exception in 10 CFR 50.109(a)(4)(i). The 

following discussion constitutes the 

documented evaluation to support the 

invocation of the compliance exception. 

A system leakage test does not verify 

fully the structural integrity of the 

repaired or replaced piping 

components. NDE examinations will 

most likely detect whether cracks exist 

and thereby ensure the structural 

integrity of the repaired or replaced 

components. The general design criteria 

(GDC) for nuclear power plants 

(Appendix A to 10 CFR part 50) provide 

the regulatory requirements for the 

NRC’s assessment of the potential for, 

and consequences of, degradation of the 

reactor coolant pressure boundary 

(RCPB). The applicable GDCs include 

GDC 14 and GDC 31. GDC 14 specifies 

that the RCPB be designed, fabricated, 

erected, and tested so as to have an 

extremely low probability of abnormal 

leakage, of rapidly propagating failure, 

and of gross rupture. GDC 31 specifies 

that the probability of rapidly 

propagating fracture of the RCPB be 

minimized. 

The nuclear plants that were licensed 

before GDC were incorporated in 10 

CFR Part 50 also would not be in 

compliance with their licensing basis 

which requires maintenance of the 

structural and leakage integrity of the 

RCPB. 

Cracking of primary system piping as 

a result of the repair or replacement is 

a non-compliance with GDC 14 because 

the RCPB must be fabricated and tested 

as to have an extremely low probability 

of abnormal leakage, of rapidly 

propagating failure and of gross rupture. 

Without an NDE, there would be no 

confirmation as to whether cracks exist 

in the component. The volumetric 

examination (NDE) will verify the 

structural integrity of the component as 

part of the repair or replacement 

activity. If a crack, especially a 

circumferential crack in a pipe, is not 

detected, it would increase the 

probability of rapidly propagating 

fracture of RCPB (i.e., a non-compliance 

with GDC 31). Therefore, cracking, if 

undetected, would be detrimental to the 

structural and leakage integrity of the 

RCPB. The NDE requirements in 

conjunction with system leakage testing 

of 50.55a(b)(2)(xx) will ensure the 

structural and leakage integrity of the 

RCPB, assuring an extremely low 

probability of abnormal leakage, and 

minimizing the probability of a rapidly 

propagating fracture of the RCPB. 

The NRC concludes that those 

licensees who use subsection IWA–

4540(a) of the 2003 addenda of the 

ASME Code, Section XI will not be in 

compliance with GDC and their 

licensing basis for the structural 

integrity of piping components 

throughout the term of their license 

(including any renewal periods) absent 

the imposition of NDE examination in 

conjunction with the system leakage 

testing. The NRC concludes, therefore, 

that 10 CFR 50.55a(b)(2)(xx) is a 

compliance backfit under 10 CFR 

50.109(a)(4)(i). 

5. Revise 10 CFR 50.55a(b)(2)(xxi) To Be 

Consistent With the NRC’s Imposed 

Condition for Code Case N–648–1 in RG 

1.147, Revision 15 

This change aligns the conditions 

imposed on visual examinations in 10 

CFR 50.55a(b)(2)(xxi) with the 

conditions imposed on Code Case N–

648–1 in RG 1.147, Revision 15. The 

imposed conditions do not represent a 

new NRC position. Therefore, this 

change is not considered as a backfit 

under 10 CFR 50.109. 


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6. Remove 10 CFR 50.55a(g)(6)(ii)(A) 

and Associated Subparagraphs on the 

Augmented Examination of the Reactor 

Vessel 

This change removes a one-time 

examination requirement which has 

been completed by all current licensees, 

and, therefore, is not considered as a 

backfit under 10 CFR 50.109. Future 

licensees will be subject to other Code 

provisions that preclude the need for 

this one-time examination. 

7. Add Paragraph (D) to 10 CFR 

50.55a(g)(6)(ii)—Reactor Vessel Head 

Inspections 

The current regulatory requirements 

for RPV head inspection are set forth in 

the First Revised NRC Order EA–03–

009, dated February 20, 2004. Order 

EA–03–009 was issued to ensure that 

boric acid corrosion of RPV heads and 

PWSCC of RPV head penetration 

nozzles and welds, which could result 

in failure of the RPV head or head 

penetrations, are promptly identified 

and corrected. The NRC determined that 

Order EA–03–009 constitutes backfitting 

as defined in 10 CFR 50.109(a)(1), but 

that the actions mandated by the Order 

were necessary for reasonable assurance 

of adequate protection to public health 

and safety. Therefore, a backfit analysis 

was not prepared for the Order in 

accordance with § 50.109(a)(4)(ii). 

Section III of the Order also stated, in 

part, ‘‘It is appropriate and necessary to 

the protection of public health and 

safety to establish a clear regulatory 

framework, pending the incorporation 

of revised inspection requirements into 

10 CFR 50.55a.’’

This rule revokes Order EA–03–009 as 

the current regulatory requirement for 

RPV head inspection, and replace it 

with ASME Code Case N–729–1, as 

modified in 10 CFR 50.55a per 10 CFR 

50.55a(g)(6)(ii)(D)(1). All current 

licensees will be required to implement 

ASME Code Case N–729–1, with the 

limitations and conditions denoted by 

this rule. The Code Case provisions on 

RPV head and head penetration 

inspections are somewhat different from 

those established in Order EA–03–009, 

and will require a licensee to modify its 

procedures for inspection of its RPV 

head and head penetrations to meet the 

requirements on the Code Case. 

Accordingly, NRC imposition of the 

Code Case may be deemed to be a 

modification of the procedures to 

operate a facility resulting from the 

imposition of new regulation, and as 

such, this rulemaking provision may be 

considered backfitting under 10 CFR 

50.109(a)(1). The NRC continues to find 

that RPV head inspections are necessary 

for adequate protection of public health 

and safety, and that the requirements of 

Code Case N–729–1, with the 

limitations and conditions denoted by 

this rule, represents an acceptable 

approach, developed by a voluntary 

consensus standards organization, for 

performing future RPV head and head 

penetration inspections. The NRC 

believes, in keeping with the intent of 

the National Technology Transfer and 

Advancement Act, that it is preferable to 

endorse a voluntary consensus standard 

such as Code Case N–729–1, with the 

limitations and conditions denoted by 

this rule, rather than continuing to rely 

upon the requirements embodied in 

Order EA–03–009. Therefore, the NRC 

concludes that NRC approval of Code 

Case N–729–1, with the limitations and 

conditions denoted by this rule, by 

incorporation by reference of that Code 

Case into § 50.55a, constitutes a 

redefinition of the requirements 

necessary to provide reasonable 

assurance of adequate protection of 

public health and safety. Therefore, a 

backfit analysis was not prepared for 

this portion of the final rule, in 

accordance with § 50.109(a)(4)(iii). 

8. Add Paragraph (E) to 10 CFR 

50.55a(g)(6)(ii)—Reactor Coolant 

Pressure Boundary Visual Inspections 

The NRC is adding 10 CFR 

50.55a(g)(6)(ii)(E) to require augmented 

inspections of Class 1 components 

fabricated with Alloy 600/82/182 

materials. The augmented inspection 

will consist of the requirements in Code 

Case N–722 which specifies ISI for PWR 

ASME Code Class 1 components 

containing materials susceptible to 

PWSCC and NRC imposed conditions to 

the Code Case to require additional NDE 

when leakage is detected and expansion 

of the inspection sample size if a 

circumferential PWSCC flaw is detected. 

The intent of conditioning the Code 

Case is to identify leakage of and 

prevent unacceptable cracks and 

corrosion in Class 1 components, which 

are part of RCPB. The requirements may 

be considered backfitting under 10 CFR 

50.109(a)(1). However, the NRC believes 

that the requirements are necessary for 

compliance with Commission 

requirements and/or license provisions. 

Therefore a backfit analysis need not be 

prepared under the ‘‘compliance’’

exception in 10 CFR 50.109(a)(4)(i). The 

following discussion constitutes the 

documented evaluation to support the 

invocation of the compliance exception. 

Failure of the RCPB could result in 

unacceptable challenges to reactor 

safety systems that, combined with 

other failures, could lead to the release 

of radioactivity to the environment. 

Based on PWSCC experience in PWRs, 

the NRC concludes that there is a 

reasonable likelihood that PWR 

licensees would not be in compliance 

with appropriate regulatory 

requirements and current licensing basis 

with respect to structural integrity and 

leak-tightness throughout the term of 

the operating license, should PWSCC 

occur in their plants. The general design 

criteria (GDC) for nuclear power plants 

(Appendix A to 10 CFR part 50) provide 

the regulatory requirements for the 

NRC’s assessment of the potential for, 

and consequences of, degradation of the 

RCPB. The applicable GDCs include 

GDC 14 and GDC 31. GDC 14 specifies 

that the RCPB be designed, fabricated, 

erected, and tested so as to have an 

extremely low probability of abnormal 

leakage, of rapidly propagating failure, 

and of gross rupture. GDC 31 specifies 

that the probability of rapidly 

propagating fracture of the RCPB be 

minimized. 

The nuclear plants that were licensed 

before GDC were incorporated in 10 

CFR Part 50 also would not be in 

compliance with their licensing basis 

which requires maintenance of the 

structural and leakage integrity of the 

RCPB. 

Leakage of primary system coolant as 

a result of PWSCC in Alloy 600/82/182 

material is a non-compliance with GDC 

14 and 31 and licensing bases because 

there have been many cases of leakage 

as a result of PWSCC of Alloy 600/82/ 

182 material in PWRs. Therefore, 

leakage as a result of PWSCC has not 

been shown to be of extremely low 

probability (i.e., a non-compliance with 

GDC 14). In addition, the operating 

experience has shown that the crack 

growth rate of PWSCC in Alloy 600/82/ 

182 material can be rapid. If PWSCC is 

not detected and removed, a crack, 

especially a circumferential crack in a 

pipe, would increase the probability of 

rapidly propagating fracture of RCPB 

(i.e., a non-compliance with GDC 31). 

Therefore, PWSCC in Alloy 600/82/182 

material, if undetected, would be 

detrimental to the structural and leakage 

integrity of the RCPB. Code Case N–722 

with conditions provides inspection 

requirements to detect PWSCC so that 

licensees can repair or replace the 

affected components, thereby 

maintaining the structural and leakage 

integrity of the RCPB, assuring an 

extremely low probability of abnormal 

leakage, and minimizing the probability 

of a rapidly propagating fracture of the 

RCPB. 

The NRC concludes that licensees 

will not be in compliance with GDC and 

their licensing basis for structural and 

leakage integrity of Class 1 components 


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that were made of Alloy 600/82/182 

material throughout the term of their 

license (including any renewal periods) 

absent the imposition of Code Case N–

722 with conditions. The NRC 

concludes, therefore, that 10 CFR 

50.55a(g)(6)(ii)(E) is a compliance 

backfit under 10 CFR 50.109(a)(4)(i). 

XII. Congressional Review Act 

In accordance with the Congressional 

Review Act of 1996, the NRC has 

determined that this action is not a 

major rule and has verified this 

determination with the Office of 

Information and Regulatory Affairs of 

OMB. 

List of Subjects in 10 CFR Part 50 

Antitrust, Classified information, 

Criminal penalties, Fire protection, 

Incorporation by reference, 

Intergovernmental relations, Nuclear 

power plants and reactors, Radiation 

protection, Reactor siting criteria, 

Reporting and recordkeeping 

requirements. 

■ For the reasons set forth in the 

preamble and under the authority of the 

Atomic Energy Act of 1954, as amended, 

the Energy Reorganization Act of 1974, 

as amended, and 5 U.S.C. 552 and 553, 

the NRC is adopting the following 

amendments to 10 CFR part 50. 

PART 50—DOMESTIC LICENSING OF 

PRODUCTION AND UTILIZATION 

FACILITIES 

■ 1. The authority citation for part 50 

continues to read as follows: 

Authority: Secs 102, 103, 104, 105, 161, 

182, 183, 186, 189, 68 Stat. 936, 937, 938, 

948, 953, 954, 955, 956, as amended, sec. 

234, 83 Stat. 444, as amended (42 U.S.C. 

2132, 2133, 2134, 2135, 2201, 2232, 2233, 

2236, 2239, 2282); secs. 201, as amended, 

202, 206, 88 Stat. 1242, as amended, 1244, 

1246 (42 U.S.C. 5841, 5842, 5846); sec. 1704, 

112 Stat. 2750 (44 U.S.C. 3504 note); sec. 

651(e), Pub. L. 109–58, 119 Stat. 806–810 (42 

U.S.C. 2014, 2021, 2021b, 2111). 

Section 50.7 also issued under Pub. L. 95–

601, sec. 10, 92 Stat. 2951 as amended by 

Pub. L. 102–486, Sec. 2902, 106 Stat. 3123 

(42 U.S.C. 5841). Section 50.10 also issued 

under secs. 101, 185, 68 Stat. 955, as 

amended (42 U.S.C. 2131, 2235), sec. 102, 

Pub. L. 91–190, 83 Stat. 853 (42 U.S.C. 4332). 

Sections 50.13, 50.54(d), and 50.103 also 

issued under sec. 108, 68 Stat. 939, as 

amended (42 U.S.C. 2138). Sections 50.23, 

50.35, 50.55, and 50.56 also issued under sec. 

185, 68 Stat. 955 (42 U.S.C. 2235). Sections 

50.33a, 50.55a and Appendix Q also issued 

under sec. 102, Pub. L. 91–190, 83 Stat. 853 

(42 U.S.C. 4332). Sections 50.34 and 50.54 

also issued under sec. 204, 88 Stat. 1245 (42 

U.S.C. 5844). Sections 50.58, 50.91, and 

50.92 also issued under Pub. L. 97–415, 96 

Stat. 2073 (42 U.S.C. 2239). Section 50.78 

also issued under sec. 122, 68 Stat. 939 (42 

U.S.C. 2152). Sections 50.80–50.81 also 

issued under sec. 184, 68 Stat. 954, as 

amended (42 U.S.C. 2234). Appendix F also 

issued under sec. 187, 68 Stat. 955 (42 U.S.C. 

2237). 

■ 2. Section 50.55a is amended by: 

■ A. Revising paragraph (b) introductory 

text, (b)(1) introductory text, (b)(1)(iii), 

(b)(2) introductory text , (b)(2)(xv) 

introductory text, (b)(2)(xx) and 

(b)(2)(xxi)(A), (b)(3) introductory text, 

and (b)(3)(iv)(D); 

■ B. Removing and reserving paragraphs 

(b)(2)(xi) and (b)(2)(xiii), and 

(g)(6)(ii)(A); and 

■ C. Adding paragraphs (g)(6)(ii)(D) and 

(g)(6)(ii)(E), to read as follows: 

§ 50.55a Codes and standards. 

* * * * * 

(b) The following standards have been 

approved for incorporation by reference 

by the Director of the Federal Register 

pursuant to 5 U.S.C. 552(a) and 1 CFR 

part 51: Sections III and XI of the ASME 

Boiler and Pressure Vessel Code and the 

ASME Code for Operation and 

Maintenance of Nuclear Power Plants, 

which are referenced in paragraphs 

(b)(1), (b)(2), and (b)(3) of this section; 

NRC Regulatory Guide 1.84, Revision 

34, ‘‘Design, Fabrication, and Materials 

Code Case Acceptability, ASME Section 

III’’ (October 2007); NRC Regulatory 

Guide 1.147, Revision 15, ‘‘Inservice 

Inspection Code Case Acceptability, 

ASME Section XI, Division 1’’ (October 

2007); and Regulatory Guide 1.192, 

‘‘Operation and Maintenance Code Case 

Acceptability, ASME OM Code’’ (June 

2003), which list ASME Code cases that 

the NRC has approved in accordance 

with the requirements in paragraphs 

(b)(4), (b)(5), and (b)(6) of this section; 

ASME Code Case N–729–1, ‘‘Alternative 

Examination Requirements for PWR 

Reactor Vessel Upper Heads With 

Nozzles Having Pressure-Retaining 

Partial-Penetration Welds, Section XI, 

Division 1’’ (Approval Date: March 28, 

2006), which has been approved by the 

NRC with conditions in accordance 

with the requirements in paragraph 

(g)(6)(ii)(D) of this section; and ASME 

Code Case N–722, ‘‘Additional 

Examinations for PWR Pressure 

Retaining Welds in Class 1 Components 

Fabricated with Alloy 600/82/182 

Materials, Section XI, Division 1’’

(Approval Date: July 5, 2005), which has 

been approved by the NRC with 

conditions in accordance with the 

requirements in paragraphs (g)(6)(ii)(E) 

of this section. Copies of the ASME 

Boiler and Pressure Vessel Code, the 

ASME Code for Operation and 

Maintenance of Nuclear Power Plants, 

ASME Code Case N–729–1, and ASME 

Code Case N–722 may be purchased 

from the American Society of 

Mechanical Engineers, Three Park 

Avenue, New York, NY 10016 or 

through the Web http://www.asme.org/ 

Codes/. Single copies of NRC Regulatory 

Guides 1.84, Revision 34; 1.147, 

Revision 15; and 1.192 may be obtained 

free of charge by writing the 

Reproduction and Distribution Services 

Section, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555–

0001; or by fax to 301–415–2289; or by 

e-mail to DISTRIBUTION@nrc.gov. 

Copies of the ASME Codes and NRC 

Regulatory Guides incorporated by 

reference in this section may be 

inspected at the NRC Technical Library, 

Two White Flint North, 11545 Rockville 

Pike, Rockville, MD 20852–2738 or call 

301–415–5610, or at the National 

Archives and Records Administration 

(NARA). For information on the 

availability of this material at NARA, 

call 202–741–6030, or go to: http:// 

www.archives.gov/federal_register/ 

code_of_federal_regulations/ 

ibr_locations.html. 

(1) As used in this section, references 

to Section III of the ASME Boiler and 

Pressure Vessel Code refer to Section III, 

and include the 1963 Edition through 

1973 Winter Addenda, and the 1974 

Edition (Division 1) through the 2004 

Edition (Division 1), subject to the 

following limitations and modifications: 

* * * * * 

(iii) Seismic design of piping. 

Applicants and licensees may use 

Articles NB–3200, NB–3600, NC–3600, 

and ND–3600 for seismic design of 

piping, up to and including the 1993 

Addenda, subject to the limitation 

specified in paragraph (b)(1)(ii) of this 

section. Applicants and licensees may 

not use these Articles for seismic design 

of piping in the 1994 Addenda through 

the latest edition and addenda 

incorporated by reference in paragraph 

(b)(1) of this section. 

* * * * * 

(2) As used in this section, references 

to Section XI of the ASME Boiler and 

Pressure Vessel Code refer to Section XI, 

and include the 1970 Edition through 

the 1976 Winter Addenda, and the 1977 

Edition (Division 1) through the 2004 

Edition (Division 1), subject to the 

following limitations and modifications: 

* * * * * 

(xi) [Reserved] 

* * * * * 

(xiii) [Reserved] 

* * * * * 

(xv) Appendix VIII specimen set and 

qualification requirements. The 

following provisions may be used to 

modify implementation of Appendix 

VIII of Section XI, 1995 Edition through 


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the 2001 Edition. Licensees choosing to 

apply these provisions shall apply all of 

the following provisions under this 

paragraph except for those in 

§ 50.55a(b)(2)(xv)(F) which are optional. 

Licensees who use later editions and 

addenda than the 2001 Edition of the 

ASME Code shall use the 2001 Edition 

of Appendix VIII. 

* * * * * 

(xx) System leakage tests. 

(A) When performing system leakage 

tests in accordance with IWA–5213(a), 

1997 through 2002 Addenda, the 

licensee shall maintain a 10-minute 

hold time after test pressure has been 

reached for Class 2 and Class 3 

components that are not in use during 

normal operating conditions. No hold 

time is required for the remaining Class 

2 and Class 3 components provided that 

the system has been in operation for at 

least 4 hours for insulated components 

or 10 minutes for uninsulated 

components. 

(B) The NDE provision in IWA–

4540(a)(2) of the 2002 Addenda of 

Section XI must be applied when 

performing system leakage tests after 

repair and replacement activities 

performed by welding or brazing on a 

pressure retaining boundary using the 

2003 Addenda through the latest edition 

and addenda incorporated by reference 

in paragraph (b)(2) of this section. 

(xxi) * * * 

(A) The provisions of Table IWB–

2500–1, Examination Category B–D, Full 

Penetration Welded Nozzles in Vessels, 

Items B3.40 and B3.60 (Inspection 

Program A) and Items B3.120 and 

B3.140 (Inspection Program B) of the 

1998 Edition must be applied when 

using the 1999 Addenda through the 

latest edition and addenda incorporated 

by reference in paragraph (b)(2) of this 

section. A visual examination with 

magnification that has a resolution 

sensitivity to detect a 1-mil width wire 

or crack, utilizing the allowable flaw 

length criteria in Table IWB–3512–1, 

1997 Addenda through the latest edition 

and addenda incorporated by reference 

in paragraph (b)(2) of this section, with 

a limiting assumption on the flaw aspect 

ratio (i.e., a/l=0.5), may be performed 

instead of an ultrasonic examination. 

* * * * * 

(3) As used in this section, references 

to the OM Code refer to the ASME Code 

for Operation and Maintenance of 

Nuclear Power Plants, and include the 

1995 Edition through the 2004 Edition 

subject to the following limitations and 

modifications: 

* * * * * 

(iv) * * * 

(D) The applicable provisions of 

subsection ISTC must be implemented if 

the Appendix II condition monitoring 

program is discontinued. 

* * * * * 

(g) * * * 

(6) * * * 

(ii) * * * 

(A) [Reserved] 

* * * * * 

(D) Reactor vessel head inspections. 

(1) All licensees of pressurized water 

reactors shall augment their inservice 

inspection program with ASME Code 

Case N–729–1 subject to the conditions 

specified in paragraphs (g)(6)(ii)(D)(2) 

through (6) of this section. Licensees of 

existing operating reactors as of [insert 

final date of rule] shall implement their 

augmented inservice inspection 

program by December 31, 2008. Once a 

licensee implements this requirement, 

the First Revised NRC Order EA–03–009 

no longer applies to that licensee and 

shall be deemed to be withdrawn. 

(2) Note 9 of ASME Code Case N–

729–1 shall not be implemented. 

(3) Instead of the specified 

‘examination method’ requirements for 

volumetric and surface examinations in 

Note 6 of Table 1 of Code Case N–729–

1, the licensee shall perform volumetric 

and/or surface examination of 

essentially 100 percent of the required 

volume or equivalent surfaces of the 

nozzle tube, as identified by Figure 2 of 

ASME Code Case N–729–1. A 

demonstrated volumetric or surface leak 

path assessment through all J-groove 

welds shall be performed. If a surface 

examination is being substituted for a 

volumetric examination on a portion of 

a penetration nozzle that is below the 

toe of the J-groove weld [Point E on 

Figure 2 of ASME Code Case N–729–1], 

the surface examination shall be of the 

inside and outside wetted surface of the 

penetration nozzle not examined 

volumetrically. 

(4) By September 1, 2009, ultrasonic 

examinations shall be performed using 

personnel, procedures and equipment 

that have been qualified by blind 

demonstration on representative 

mockups using a methodology that 

meets the conditions specified in 

(50.55a(g)(6)(ii)(D)(3)(i) through 

(50.55a(g)(6)(ii)(D)(3)(iv), instead of the 

qualification requirements of Paragraph 

–2500 of ASME Code Case N–729–1. 

References herein to Section XI, 

Appendix VIII shall be to the 2004 

Edition with no Addenda of the ASME 

BPV Code. 

(i) The specimen set shall have an 

applicable thickness qualification range 

of +25 percent to ¥40 percent for 

nominal depth through-wall thickness. 

The specimen set shall include 

geometric and material conditions that 

normally require discrimination from 

primary water stress corrosion cracking 

(PWSCC) flaws. 

(ii) The specimen set shall have a 

minimum of ten (10) flaws which 

provide an acoustic response similar to 

PWSCC indications. All flaws shall be 

greater than 10 percent of the nominal 

pipe wall thickness. A minimum of 20 

percent of the total flaws shall initiate 

from the inside surface and 20 percent 

from the outside surface. At least 20 

percent of the flaws shall be in the 

depth ranges of 10–30 percent through 

wall thickness and at least 20 percent 

within a depth range of 31–50 percent 

through wall thickness. At least 20 

percent and no more than 40 percent of 

the flaws shall be oriented axially. 

(iii) Procedures shall identify the 

equipment and essential variables and 

settings used for the qualification, and 

are consistent with Subarticle VIII–2100 

of Section XI, Appendix VIII. The 

procedure shall be requalified when an 

essential variable is changed outside the 

demonstration range as defined by 

Subarticle VIII–3130 of Section XI, 

Appendix VIII and as allowed by 

Articles VIII–4100, VIII–4200 and VIII–

4300 of Section XI, Appendix VIII. 

Procedure qualification shall include 

the equivalent of at least three personnel 

performance demonstration test sets. 

Procedure qualification requires at least 

one successful personnel performance 

demonstration. 

(iv) Personnel performance 

demonstration test acceptance criteria 

shall meet the personnel performance 

demonstration detection test acceptance 

criteria of Table VIII—S10–1 of Section 

XI, Appendix VIII, Supplement 10. 

Examination procedures, equipment, 

and personnel are qualified for depth 

sizing and length sizing when the RMS 

error, as defined by Subarticle VIII–3120 

of Section XI, Appendix VIII, of the flaw 

depth measurements, as compared to 

the true flaw depths, do not exceed 1⁄8

inch (3 mm), and the root mean square 

(RMS) error of the flaw length 

measurements, as compared to the true 

flaw lengths, do not exceed 3⁄8 inch (10 

mm), respectively. 

(5) If flaws attributed to PWSCC have 

been identified, whether acceptable or 

not for continued service under 

Paragraphs –3130 or –3140 of ASME 

Code Case N–729–1, the re-inspection 

interval must be each refueling outage 

instead of the re-inspection intervals 

required by Table 1, Note (8) of ASME 

Code Case N–729–1. 

(6) Appendix I of ASME Code Case 

N–729–1 shall not be implemented 

without prior NRC approval. 


pwalker on PROD1PC71 with RULES2
52750 Federal Register / Vol. 73, No. 176 / Wednesday, September 10, 2008 / Rules and Regulations 

1For inspections to be conducted every refueling 

outage and inspections conducted every other 

refueling outage, the initial inspection shall be 

performed at the next refueling outage after January 

1, 2009. For inspections to be conducted once per 

interval, the inspections shall begin in the interval 

in effect on January 1, 2009, and shall be prorated 

over the remaining periods and refueling outages in 

this interval. 

(E) Reactor coolant pressure boundary 

visual inspections.1

(1) All licensees of pressurized water 

reactors shall augment their inservice 

inspection program by implementing 

ASME Code Case N–722 subject to the 

conditions specified in paragraphs 

(g)(6)(ii)(E)(2) through (4) of this section. 

The inspection requirements of ASME 

Code Case N–722 do not apply to 

components with pressure retaining 

welds fabricated with Alloy 600/82/182 

materials that have been mitigated by 

weld overlay or stress improvement. 

(2) If a visual examination determines 

that leakage is occurring from a specific 

item listed in Table 1 of ASME Code 

Case N–722 that is not exempted by the 

ASME Code, Section XI, IWB–

1220(b)(1), additional actions must be 

performed to characterize the location, 

orientation, and length of crack(s) in 

Alloy 600 nozzle wrought material and 

location, orientation, and length of 

crack(s) in Alloy 82/182 butt welds. 

Alternatively, licensees may replace the 

Alloy 600/82/182 materials in all the 

components under the item number of 

the leaking component. 

(3) If the actions in paragraph 

(g)(6)(ii)(E)(2) of this section determine 

that a flaw is circumferentially oriented 

and potentially a result of primary water 

stress corrosion cracking, licensees shall 

perform non-visual NDE inspections of 

components that fall under that ASME 

Code Case N–722 item number. The 

number of components inspected must 

equal or exceed the number of 

components found to be leaking under 

that item number. If circumferential 

cracking is identified in the sample, 

non-visual NDE must be performed in 

the remaining components under that 

item number. 

(4) If ultrasonic examinations of butt 

welds are used to meet the NDE 

requirements in paragraphs 

(g)(6)(ii)(E)(2) or (g)(6)(ii)(E)(3) of this 

section, they must be performed using 

the appropriate supplement of Section 

XI, Appendix VIII of the ASME Boiler 

and Pressure Vessel Code. 

* * * * * 

For the U.S. Nuclear Regulatory 

Commission. 

Dated at Rockville, Maryland, this 18th day 

of August 2008. 

R.W. Borchardt, 

Executive Director for Operations. 

[FR Doc. E8–20624 Filed 9–9–08; 8:45 am] 

BILLING CODE 7590–01–P 


pwalker on PROD1PC71 with RULES2