Proposed Rule Draft Regulatory Guide-1344 (RG 1.193, Rev. 6)

U.S. NUCLEAR REGULATORY COMMISSION
DRAFT REGULATORY GUIDE DG-1344
Proposed Revision 6 to Regulatory Guide RG 1.193
Issue Date: August 2018
Technical Lead: Giovanni Facco

ASME CODE CASES NOT APPROVED FOR USE
A. INTRODUCTION
Purpose
This regulatory guide (RG) lists the American Society of Mechanical Engineers (ASME) Code
Cases that the U.S. Nuclear Regulatory Commission (NRC) has determined not to be acceptable for use
on a generic basis. This regulatory guide does not approve the use of the Code Cases listed herein.
Applicable Rules and Regulations
•

•

Title 10, Part 50, of the Code of Federal Regulations (10 CFR Part 50), “Domestic Licensing of
Production and Utilization Facilities” (Ref. 1).
o

Section 50.55a(c) (10 CFR 50.55a(c)), “Reactor Coolant Pressure Boundary,” requires, in
part, that components of the reactor coolant pressure boundary be designed, fabricated,
erected, and tested in accordance with the requirements for Class 1 components of
Section III, “Rules for Construction of Nuclear Power Plant Components,” of the ASME
Boiler and Pressure Vessel (BPV) Code or equivalent quality standards.

o

10 CFR 50.55a(f), “Inservice Testing Requirements,” requires, in part, that Class 1, 2,
and 3 components and their supports meet the requirements of the ASME “Operation
and Maintenance of Nuclear Power Plants” (OM Code) or equivalent quality standards.

o

10 CFR 50.55a(g), “Inservice Inspection Requirements,” requires, in part, that Class 1, 2,
3, MC (metal containment), and CC (concrete containment) components and their
supports meet the requirements of Section XI, “Rules for Inservice Inspection of Nuclear
Power Plant Components,” of the ASME BPV Code or equivalent quality standards.

10 CFR 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants” (Ref. 2), Section
10 CFR 52.79(a)(11) requires that, “[The final safety analysis report shall include the following
information:] A description of the program(s), and their implementation, necessary to ensure that

This RG is being issued in draft form to involve the public in the development of regulatory guidance in this area. It has not received final staff
review or approval and does not represent an NRC final staff position. Public comments are being solicited on this DG and its associated
regulatory analysis. Comments should be accompanied by appropriate supporting data. Comments may be submitted through the Federal
rulemaking Web site, http://www.regulations.gov, by searching for draft regulatory guide DG-1344. Alternatively, comments may be submitted
to the Rules, Announcements, and Directives Branch, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 205550001. Comments must be submitted by the date indicated in the Federal Register notice.
Electronic copies of this DG, previous versions of this guide, and other recently issued guides are available through the NRC’s public Web site
under the Regulatory Guides document collection of the NRC Library at http://www.nrc.gov/reading-rm/doc-collections/reg-guides/. The DG is
also available through the NRC’s Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/readingrm/adams.html, under Accession No. ML18114A227. The regulatory analysis may be found in ADAMS under Accession No. ML18099A041.

the systems and components meet the requirements of the ASME Boiler and Pressure Vessel
Code and the ASME Code for Operation and Maintenance of Nuclear Power Plants in accordance
with 50.55a of this chapter.”
Related Guidance
•

RG 1.84, “Design, Fabrication, and Materials Code Case Acceptability, ASME Section III”
(Ref. 3), lists the ASME BPV Code, Section III, Code Cases, that the NRC has approved for use
as voluntary alternatives to the mandatory ASME BPV Code provisions that are incorporated into
10 CFR 50.55a.

•

RG 1.147, “Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1”
(Ref. 4), lists the ASME BPV Code, Section XI, Code Cases, that the NRC has approved for use
as voluntary alternatives to the mandatory ASME BPV Code provisions that are incorporated into
10 CFR 50.55a.

•

RG 1.192, “Operation and Maintenance Code Case Acceptability, ASME OM Code” (Ref. 5),
lists the ASME Operation and Maintenance Code (OM Code) (Ref. 6) Code Cases that the NRC
has approved for use as voluntary alternatives to the mandatory ASME OM Code provisions that
are incorporated into 10 CFR 50.55a.

Purpose of This Regulatory Guide
This RG is issued to provide information to applicants and licensees regarding those Code Cases
that the NRC has determined not to be acceptable for use on a generic basis. A brief description of the
basis for the determination is provided with each Code Case. Applicants or licensees may submit a
request to implement one or more of the Code Cases listed below through 10 CFR 50.55a(z), which
permits the use of alternatives to the Code requirements referenced in 10 CFR 50.55a, provided that the
proposed alternatives result in an acceptable level of quality and safety. Applicants or licensees must
submit a plant-specific request that addresses the NRC’s concerns about the Code Case at issue. The NRC
will revise this regulatory guide as needed to address subsequent new or revised Code Cases.
Paperwork Reduction Act
This RG provides guidance for implementing the mandatory information collections in 10 CFR
Parts 50 and 52 that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et. seq.). These
information collections were approved by the Office of Management and Budget (OMB), under control
numbers 3150-0011 and 3150-0151. Send comments regarding this information collection to the
Information Services Branch, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by
e-mail to Infocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory
Affairs, NEOB-10202 (3150-0011, 3150-0151), Office of Management and Budget, Washington, DC
20503.
Office of the Chief Information Officer (OCIO) will review this paragraph to ensure that the
correct control number is being used. The list of OCIO control numbers are located here:
http://fusion.nrc.gov/ois/team/CSD/FPIB/ICT/Shared Documents/Clearance List.xlsx
Public Protection Notification
The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of

DG-1344, Page 2

information unless the document requesting or requiring the collection displays a currently valid OMB
control number.

DG-1344, Page 3

B. DISCUSSION
Reason of Revision
Revision 6 of RG 1.193 includes new information reviewed by the NRC of Section III and
Section XI BPV Code Cases listed in Supplement 11 to the 2010 Edition and Supplements 0 through 7 to
the 2013 Edition, and the OM Code Cases listed in the 2015 and 2017 Editions. This is an update to RG
1.193, Revision 5, that included information from Supplement 11 of the 2010 Edition and Supplements 1
through 10 to the 2010 Edition (Sections III and XI), and the 2009 Edition through the 2012 Edition of
the OM Code.
Background
The ASME publishes a new edition of the BPV and OM Codes every 2 years. The latest editions
and addenda of the ASME BPV Code, Section III and Section XI, and the ASME OM Code that the NRC
has approved for use by applicants and licensees are referenced in 10 CFR 50.55a(a). The ASME also
publishes Code Cases for Section III and Section XI quarterly and Code Cases for the OM Code
biennially. Code Cases provide alternatives developed and approved by the ASME.
The NRC staff reviewed Section III and Section XI Code Cases listed in Supplement 11 to the 2010
Edition and Supplements 0 through 7 to the 2013 Edition. Revision 38 of Regulatory Guide 1.84 and
Revision 19 of Regulatory Guide 1.147 have been published concurrently with this guide to identify the
Code Cases that the NRC has determined to be acceptable alternatives to applicable parts of Section III
and Section XI. The NRC staff also reviewed the OM Code Cases listed in the 2015 Edition through the
2017 Edition. Revision 3 of Regulatory Guide 1.192 has also been published concurrently with this guide
to identify the Code Cases that the NRC has determined to be acceptable alternatives to applicable parts
of the OM Code.

DG-1344, Page 4

C. STAFF REGULATORY GUIDANCE
Licensees should not implement Code Cases from the Section III and Section XI Codes listed in
Supplement 11 to the 2010 Edition, Supplements 0 through 7 to the 2013 Edition, and the OM Code
Cases listed in the 2015 Edition through the 2017 Edition of the OM Code, that are listed in this guide,
without prior NRC approval. The Code Cases addressed by this regulatory guide are listed in three tables:
(1) Table 1, “Unacceptable Section III Code Cases,” contains Section III Code Cases that are
unacceptable for use by licensees in their Section III design and construction programs.
(2) Table 2, “Unacceptable Section XI Code Cases,” contains Section XI Code Cases that are
unacceptable for use by licensees in their Section XI inservice inspection programs.
(3) Table 3, “Unacceptable OM Code Cases,” contains OM Code Cases that were determined to be
unacceptable for use by licensees in their inservice testing programs.
1.

Unacceptable Section III Code Cases

The NRC determined that the following Section III Code Cases are unacceptable for use
by licensees in their Section III design and construction programs. To assist users, new Code Cases
are shaded to distinguish them from those listed in previous versions of this guide.
Table 1. Unacceptable Section III Code Cases
CODE CASE
NUMBER

TABLE 1
UNACCEPTABLE SECTION III CODE CASES
SUMMARY

DATE OR
SUPPLEMENT/
EDITION

N-201-6

Class CS Components in Elevated Temperature Service, Section
III, Division 1
Code Case is applicable for high temperature applications beyond
that of light water reactors.

10/18/10

N-284-1

Metal Containment Shell Buckling Design Methods, Section III,
Division 1, Class MC
(1) The following errata, misprints, recommendations, and errors
have been identified:
 Fig. 1511.1, The curve for αθL should not exceed 0.8
for any value of (R/t).
 -1512, The statement “See Fig. 1512-1 then see -1713.1.2
for method of calculating M” should be rephrased as:
“See -1713.1.2 for method of calculating M, then see
Fig. -1512-1.”
 -1513, Recommend “Use the value of αil given for
spherical shells in accordance with -1512.”
 -1521, (i) In (a) Axial Compression, “αθG = αθL” should be
changed to “αφG = αφL.” (ii) The source of the equations
shown under “(a) Axial Compression” provided separate
instability equations for stringer-stiffened and
ring-stiffened cylindrical shells. The Code Case adopted

5/9/03
5/9/03

DG-1344, Page 5

CODE CASE
NUMBER

TABLE 1
UNACCEPTABLE SECTION III CODE CASES
SUMMARY

DATE OR
SUPPLEMENT/
EDITION

the instability equations pertaining to ring-stiffened shells,
which are less conservative than those for stringer
instability, for both ring and/or stringer stiffened
cylindrical shells. The Code Case should use the most
limiting case (that gives a lower allowable stress for
instability based on a smaller value of capacity reduction
factor), or provide separate equations for the stringer
stiffened case and ring stiffened case.
 -1712.1.1, The equation “Cθh = 0.92/(Mθ - 0.636)” should
be changed to “Cθh = 0.92/(Mφ - 0.636).”
 -1712.1.1-1, The leftmost curve should be labeled Cθh.
 -1712.2.2, (a) Axial Compression, (i) In the formula for
σφej, the denominator should be (mπ/Lj)2 Χ tφ. (ii) The
expressions for Cφ and Cθ should be separated.
 -1712.2.3, (i) The factor 1.944 in an older edition has
been changed to 2.00. No basis is apparent. (ii) The
misprint “t1¼.” should be corrected to “t1¼.”
 -1713.1.1, (i) The equation “στa=αφθΧσφθel/FS” should be
changed to “στa=αφθLΧσφθel/FS.” (ii) The title of (c) should
be changed to “Axial Compression Plus In-Plane Shear.”
 -1713.1-1, In (b), the lower value “Ks=σra” on the vertical
axis should be changed to “Ks=σha.”
 -1713.2.1, (i) The headings for (b) and (c) should include
the words “In-Plane.” (ii) In (b) “Axial Compression
Plus Shear,” “σθ” should be changed to “σφ.”
(2) Applicants intending to use Code Case N-284-1 shall submit
a request to the NRC staff for its review and approval on
a plant-specific basis.
(3) The rules applicable to evaluate the buckling and instability
of containment shells for Section III, Division 3, are under
development. Currently, use of Code Case N-284-1 by
licensees for storage canisters and transportation casks
is permissible provided it has been reviewed and approved by
the NRC.
N-483-2
N-483-3

Alternative Rules to the Provisions of NCA-3800, Requirements for
Purchase of Material, Section III, Divisions 1 and 3
The Code Case lacks sufficient detail to ensure that the supplied
material is as represented by the Certified Material Test Report.

5/7/99
2/25/02

N-510
N-510-1

Borated Stainless Steel for Class CS Core Support Structures and
Class 1 Component Supports, Section III, Division 1
No technical basis was provided for expanding the Code Case
to include borated stainless steel Types 304B, 304B1, 304B2,
and 304B3. A considerable amount of information was required
to support the types presently contained in the Code Case.

12/9/93
8/14/01

DG-1344, Page 6

CODE CASE
NUMBER

TABLE 1
UNACCEPTABLE SECTION III CODE CASES
SUMMARY

DATE OR
SUPPLEMENT/
EDITION

The revised Code Case would permit borated stainless steel to be
used for component supports within the reactor vessel.
The technical basis to support the Code Case only addresses
the use of these materials as component supports in spent fuel
racks and transportation casks.
N-519

Use of 6061-T6 and 6061-T651 Aluminum for Class 1 Nuclear
Components
Code Case is applicable to only one DOE aluminum vessel.

Annulled 2/3/03

N-530

Provisions for Establishing Allowable Axial Compressive
Membrane Stresses in the Cylindrical Walls of 0-15 Psi Storage
Tanks, Classes 2 and 3 Section III, Division 1
There are numerous errors in the equations. The errors must be
corrected before the Code Case can be approved for use.

2/3/03

N-565

Alternative Methods of Nozzle Attachment for Class 1 Vessels
Section III, Division 1
The Code Case essentially requires a design using a seal to protect
the threads from the contained fluid, and seals are not a Code item.
The seal, which plays a very important part in the integrity of the
joint, imposes too great a vulnerability in the design. The
supporting information for the Code Case does not demonstrate the
resulting threaded nozzle configuration is equivalent in integrity to
that of a welded connection.

12/3/99

N-595
N-595-1
N-595-2
N-595-3
N-595-4

Requirements for Spent Fuel Storage Canisters, Section III,
Division 1
Regulatory approval for the use of multi-purpose casks is presently
addressed by the NRC Spent Fuel Project Office Interim Staff
Guidance No. 4 (ISG-4), Rev. 1, “Cask Closure Weld Inspections”
(Ref. 7), and NRC Spent Fuel Storage and Transportation Division
Interim Staff Guidance No. 18 (ISG-18) Rev. 1, “The Design and
Testing of Lid Welds on Austenitic Stainless Steel Canisters as
Containment Boundary for Spent Fuel Storage” (Ref. 8). The
interim staff guidance provides a framework to ensure that the cask
system, as designed, and when fabricated and used in accordance
with the conditions specified in its Certificate of Compliance,
meets the requirements of 10 CFR Part 72, “Licensing
Requirements for the Independent Storage of Spent Nuclear Fuel,
High-Level Radioactive Waste, and Reactor-Related Greater than
Class C Waste” (Ref. 9). It should be noted that Code Case N-717
replaces Code Case N-595-X.

2/26/99
9/24/99
12/8/00
04/08/02
Annulled
10/14/11

N-645
N-645-1

Use of Rupture Disk Devices on Nuclear Fuel Storage Canisters,
Class 1, Section III, Division 1

DG-1344, Page 7

6/14/00
2/3/03

CODE CASE
NUMBER

TABLE 1
UNACCEPTABLE SECTION III CODE CASES
SUMMARY

DATE OR
SUPPLEMENT/
EDITION

The NRC does not permit the use of rupture disk devices in spent
nuclear fuel storage canister designs.
N-659
N-659-1

Use of Ultrasonic Examination in Lieu of Radiography for Weld
Examination, Section III, Division 1
The NRC conditionally approved Code Case N-659 in Revision 34
of Regulatory Guide 1.84. The NRC’s issues and proposed
conditions were discussed in the statement of considerations for the
proposed rule. The public comments discussed a number of
concerns with the proposed conditions. Given the number of issues
raised by NRC staff and the concerns expressed in the public
comments, the NRC determined that a more effective approach for
developing a suitable performance demonstration program was to
work with ASME International to resolve the issues. Accordingly,
the NRC is not going to endorse Code Case N-659 or Code Case
N-659-1 at this time. NRC staff continue to interact with the
cognizant ASME committees, and the industry is working to
provide additional data and information.

9/17/02
11/18/03

N-659-2

Use of Ultrasonic Examination in Lieu of Radiology for Weld
Examination, Section III, Divisions 1 and 3
The NRC is not going to endorse Code Case N-659-2 at this time.
Research is currently being conducted on a number of issues with
respect to using ultrasonic testing (UT) to replace radiographic
testing (RT). While preliminary results suggest that replacement of
RT with UT may be feasible, the interchangeability of these
techniques has not yet been fully demonstrated, UT acceptance
criteria for fabrication/construction weld inspection have not yet
been adequately defined, and the applicability of UT in the
presence of high levels of acoustic noise such as that found in
austenitic materials is not fully understood. The impact and
implications of the expanded examination volume (full-thickness)
required for UT for fabrication/construction must also be
addressed.

6/09/08

In addition, the Code Case would allow the examinations to be
performed in accordance with Section V, Article 5 up to and
including the 2001 Edition or Article 4 for later edition and
addenda. The reliability UT performed to the provisions of Section
V has been shown to be inferior to UT techniques developed
through a program where the performance characteristics have
been shown to be sufficient and reliable.
Furthermore, the qualification specimens do not specify an
adequate number of flaws required to be in the sample set, the
required flaw distribution within the specimen, nor the required

DG-1344, Page 8

CODE CASE
NUMBER

TABLE 1
UNACCEPTABLE SECTION III CODE CASES
SUMMARY

DATE OR
SUPPLEMENT/
EDITION

size distribution within the specimen. Therefore, performance
demonstration requirements including acceptance criteria for UT
equipment, procedures, and personnel used for
construction/fabrication activities must be addressed.
Until studies are complete that demonstrate the ability of UT to
replace RT for fabrication/construction, the NRC will not endorse
UT in lieu of RT Code Cases or generically allow the substitution
of UT in lieu of RT for fabrication/construction examinations.
N-670

Use of Ductile Cast Iron Conforming to ASTM A 874/ A 874M-98
or JIS G5504-1992 for Transport Containments, Section III,
Division 3
The NRC has not yet endorsed Section III, Division 3,
“Containments for Transportation and Storage of Spent Nuclear
Fuel and High Level Radioactive Material and Waste.” Thus, it
would not be appropriate to approve a Code Case that is an
alternative to the Section III, Division 3, provisions.

7/01/05

N-673

Boron Containing Power Metallurgy Aluminum Alloy for Storage
and Transportation of Spent Nuclear Fuel, Section III, Division 1
The Code Case does not address the following:
(1) Corrosion properties of this material in spent fuel pool
chemistry and/or clean water.
(2) Impact properties for use as a structural material.
(3) Uniform distribution of boron carbide in the aluminum matrix.
(4) Mechanical properties for the use of the material in hightemperature conditions.

8/7/03

N-693

Alternative Method to the Requirements of NB-3228.6 for
Analyzing Piping Subjected to Reversing Dynamic Load, Section
III, Division 1
The Code Case would permit the use of the design, service,
and test limits in Paragraph NB-3656(b) for Level D Service
Limits. The limits in Paragraph NB-3656(b) are prohibited
per 10 CFR 50.55a(b)(1)(iii).

5/21/03

N-707

Use of SA-537, Class 1 Plate Material for Spent-Fuel Containment
Internals in Non-pressure Retaining Applications Above 700°F
(370°C), Section III, Division 3
The NRC has not yet endorsed Section III, Division 3,
“Containments for Transportation and Storage of Spent Nuclear
Fuel and High Level Radioactive Material and Waste.” Thus, it
would not be appropriate to approve a Code Case that is an
alternative to the Section III, Division 3, provisions.

11/02/04

DG-1344, Page 9

CODE CASE
NUMBER

TABLE 1
UNACCEPTABLE SECTION III CODE CASES
SUMMARY

DATE OR
SUPPLEMENT/
EDITION

N-717

Requirements for Construction of Storage Containments for Spent
Nuclear Fuel and High Level Radioactive Waste and Material,
Section III, Division 3
The NRC has not yet endorsed Section III, Division 3,
“Containments for Transportation and Storage of Spent Nuclear
Fuel and High Level Radioactive Material and Waste.” Thus, it
would not be appropriate to approve a Code Case that is an
alternative to the Section III, Division 3, provisions.
The provisions of the Code Case are copied from the July 1, 2005,
addenda to Section III, Division 3. The changes to the ASME Code
contained in the addenda are scheduled to be reviewed by the NRC
staff in FY 2009. The Code Case is listed in this guide pending the
results of the NRC staff review.

5/04/04

N-721

Alternative Rules for Linear Piping Supports, Section III, Division
1
Code Case N-721 allows the use of ANSI/AISC N690L-03, “Load
and Resistance Factor Design (LRFD) Specification for SafetyRelated Steel Structures for Nuclear Facilities.” ANSI/AISC
N690L-03 provides an alternative method of design to that given in
ANSI/AISC N690-1994, “Specification for the Design,
Fabrication, and Erection of Safety-Related Steel Structures for
Nuclear Facilities,” including Supplement No. 2, which is based on
Allowable Stress Design (ASD) specification.

9/09/08

The LRFD method is a probabilistic method developed to provide
uniform practice in the design of steel structures for nuclear
facilities. The LRFD method uses many factors including one
factor per resistance, and one factor for each of the different load
types whereas the ASD method uses one factor of safety. The ASD
method is a deterministic and normally conservative method and
has been approved by the NRC for use in the design of new
reactors.
The LRFD method continues to undergo development. Code Case
N-721 was developed based on N690L-03 which has subsequently
been superseded by N690L-06. Thus, the Code Case is not up-todate. In addition, questions regarding uncertainty remain with
regard to the probabilistic treatment of loads and resistances. Thus,
the LRFD method has not yet been approved by the NRC for use in
the design of new reactor facilities.
N-728

Use of ASTM B 932-04 Plate Material for Nonpressure Retaining
Spent Fuel Containment Internals to 650°F (343°C), Section III,
Division 3

DG-1344, Page 10

10/11/05

CODE CASE
NUMBER

TABLE 1
UNACCEPTABLE SECTION III CODE CASES
SUMMARY

DATE OR
SUPPLEMENT/
EDITION

The NRC has not yet endorsed Section III, Division 3,
“Containments for Transportation and Storage of Spent Nuclear
Fuel and High Level Radioactive Material and Waste.” Thus, it
would not be appropriate to approve a Code Case that is an
alternative to the Section III, Division 3, provisions.
N-755
N-755-1
N-755-2

Use of Polyethylene (PE) Plastic Pipe, Section III, Division 1 and
XI
Issues have been raised regarding materials, fusion qualification
requirements, nondestructive examination (NDE), crack growth
and lack of data to support operating experience.

3/22/07
7/15/11
1/13E

N-761

Fatigue Design Curves for Light Water Reactor (LWR)
Environments, Section III, Division 1
Research has shown that the effect of environment on reactor
components exposed to reactor water is not bounded by the current
air fatigue curves. Bounding curves and a series of other curves for
known strain rates have been developed to account for the
reduction of fatigue life.
• The proposed curves in Code Case N-761 for carbon and low
alloy steels (as shown in Fig. 2 & Table 1 of the Code Case, and
the curves for austenitic stainless steels (as shown in Fig. 3 &
Table 2 of the Code Case) are not acceptable as sufficient
technical basis has not been provided.
• These curves are developed based on a factor of 10 on cycles
and a factor of 2 on stress, which are not in agreement with the
factor of 12 on cycles and a factor of 2 on stress as established in
NUREG/CR-6909, “Effect of LWR Coolant Environments on
the Fatigue Life of Reactor Materials” (Ref. 10). The factor of
10 on cycles is technically inconsistent with the factor of 12 in
NUREG/CR-6909. The proposed curves are non-conservative
relative to the estimates based on NUREG/CR-6909 procedure.
The use of a different set of factors for the consideration of the
LWR coolant environmental effects (i.e., a factor of 10 on
cycles and a factor of 2 on stress) for the environmental fatigue
correction factor (Fen) approach versus the environmental fatigue
curves approach is inconsistent from technical and regulatory
perspective.
• The technical basis document for the proposed Code Case does
not describe the process step by step from beginning to end as to
how final design curves for LWR environment are obtained. The
basis document does not provide the expression for the best-fit
S-N curve of the experimental data, and the details of the mean
stress correction for each curve, and how the proposed design
curves were obtained.

9/20/10
9/20/10

DG-1344, Page 11

CODE CASE
NUMBER

TABLE 1
UNACCEPTABLE SECTION III CODE CASES
SUMMARY

DATE OR
SUPPLEMENT/
EDITION

• The proposed Code Case contains five environmental fatigue
curves for carbon and low-alloy steels and five for stainless
steels. These are the air curve, and the worst case environmental
curve, and three other curves for different strain rates. These
environmental curves are not consistent with the experimental
data. The strain rate dependence for the first three curves is
much lower than that observed in experimental data on smooth
cylindrical or tube specimens or even the recent Electric Power
Research Institute sponsored component tests in Germany.
• There is no information provided in the basis document about
the operating conditions that were used to represent the worst
case environmental curve. Also, no information is provided in
the basis document regarding the equation for the best-fit curve
of the experimental data.
• The technical basis document for the code case should address
the effect of strain threshold and tensile hold time in fatigue
evaluations.
N-791

Shear Screw and Sleeve Splice, Section III, Division 2
There is no slip criterion for this code case. The staff believe that
ASTM A 1034/A1034M-05b, “Standard Test Methods for Testing
Mechanical Splices for Steel Reinforcing Bars” (Ref. 11), could be
used as a good model to develop definition and test methods for
slip.
Concrete containments in nuclear power plants are important
structures and therefore their criteria for mechanical splices should
not be less stringent than that of other seismic Category I
structures, as defined in ACI 349-06, “Code Requirements for
Nuclear Safety-Related Concrete Structures & Commentary” (Ref.
12). The design criterion for concrete containment structures is
based on allowable strains for the steel reinforcing bars. The
purpose of this strain criterion is partially to prevent the tearing of
steel liner plates, which are attached to the inside face of the
containment and serve as a leak tight pressure boundary, by
limiting strains in both concrete and steel reinforcing bars in
containment. The mechanical splices should not be allowed to have
a significant slip that would cause the strain from the steel
reinforcing bars to be transferred to the steel liner plates.
Therefore, the code case needs to develop a slip criterion for
mechanical splices.
Concrete and Commentary,” Section 21.1.6.1, classifies
mechanical splices in two types: Type 1 and 2. The criterion for
Type 1 mechanical splices is that a mechanical splice shall develop
no less than 125% of the specified minimum yield strength of the

DG-1344, Page 12

9/20/10

CODE CASE
NUMBER

TABLE 1
UNACCEPTABLE SECTION III CODE CASES
SUMMARY

DATE OR
SUPPLEMENT/
EDITION

spliced bar, as stated in Section 12.14.3.2 of the Code. Type 1
mechanical splices are not allowed to be used in regions that may
experience steel yielding. The criterion for Type 2 mechanical
splices is that a mechanical splice shall develop the specified
tensile strength of the spliced bar, as stated in Section 21.1.6.1 of
the Code. The specified, or actual tensile strength of the steel
reinforcing bars are used to calculate the ultimate capacity of
concrete containment structures against the internal pressure, as a
measure of the safety margin above the design basis accident
pressure. Consequently, Type 2 mechanical splices must be used in
concrete containment structures. Therefore, the criterion in Section
2.3 of N-791 code case is the equivalent criterion for Type 1
mechanical splices of ACI 318, “Building Code Requirements for
Structural Concrete and Commentary” (Ref. 13), which is not an
adequate criterion for qualifying mechanical splices for use in
concrete containment structures. Therefore, the code case should
develop a more stringent strength criterion, and the same criterion
should also be used for continuing splices performance tests in the
field, as stated in Section 5 of the code case.
N-792
N-792-1

Fatigue Evaluations Including Environmental Effects, Section III,
Division 1
This code case does not implement the latest methodology
developed from NRC/RES research activities. That methodology
was presented to ASME in May 2012, as reflected in the material
posted in ADAMS at ML13008A005. There are also further
adjustments to that information based on the finalization of our
research efforts.
To be more specific, the six most significant differences
betweenthe code case and the latest NRC research are shown
below:
a. Carbon and Low Alloy Steel Fatigue Curve: Figure -21001 (and -1M) and Table -2100-1 of the code case define the
design fatigue air curve for carbon and low alloy steels.
Both material types are combined into one fatigue curve,
whereas the NRC approach defines a separate fatigue
curve for each material type. The code case fatigue curve
matches the design fatigue air curve currently in Appendix
I of Section III (2011 Addenda). The code case fatigue
curve does not matchthe carbon or low alloy steel design
fatigue air curves from the initial revision of NUREG/CR6909 (which are the same curves NRC intends to use in
Rev. 1 of NUREG/CR-6909) because the code case fatigue
curve utilizes a margin of 20 on cycles whereas the NRC
curves use a margin of 12. The code case design fatigue air
curve is conservative with respect to the NRC fatigue

DG-1344, Page 13

9/20/10
11/10E

CODE CASE
NUMBER

TABLE 1
UNACCEPTABLE SECTION III CODE CASES
SUMMARY
design air curves; however, Item (b) below must also be
considered when evaluating the adequacy of Fen usage
factors calculated using the design curve.
b. Carbon and Low Alloy Steel Fen Expression: Equation (1)
of the code case uses the carbon steel Fen expression from
the initial revision of NUREG/CR-6909 adjusted to
account for the difference in the margin term used to
develop the ASME and NRC design fatigue curves. This
equation is different than the Fen expression recently
developed by the NRC, and the equations for the
transformed environmental parameters are different, so the
Fen equation may yield non-conservative values of Fen
because:
a. The use of average temperature with the code case
Fen expression may be non-conservative (see Item
(f)).
b. The code case Fen expression was adjusted to
account for the difference in margin used for to
develop the design curve, i.e., the factor of 20 vs.
12 discussed under Item (a) above. As a result, the
constant in the Fen expression is 0.121 compared
to 0.632 for carbon steel material in the initial
revision to NUREG/CR-6909. Such adjustment is
not appropriate and may be non-conservative for
Fen application to the portion of the fatigue design
air curve that is controlled by the factor of 2 on
stress rather than the factor of 20 .on cycles.
c. The code case Fen expression is for carbon steel
material, and it is used for application to both
carbon and low alloy steel materials. Use of this
expression for low alloy steel may be nonconservative since the constant is higher for low
alloy steel compared to carbon steel (0.702 vs.
0.632).
d. The code case Fen expression is nonconservative
for some environmental conditions compared to
the new NRC expressions, i.e., for T < 200°C,
strain rate = 0.001 %/sec, and dissolved oxygen
values higher than 0.04 ppm.
c. Stainless Steel Fatigue Curve: Figure -2100-2 (and -2M)
and Table -2100-2 of the code case define the design
fatigue air curve for stainless steels. The code case fatigue
curve matches the design fatigue air curve currently in
Appendix I of Section III (2011 Addenda). The code case
fatigue urve matches the stainless steel design fatigue air
curve from the initial revision of NUREG/CR-6909 (which

DG-1344, Page 14

DATE OR
SUPPLEMENT/
EDITION

CODE CASE
NUMBER

TABLE 1
UNACCEPTABLE SECTION III CODE CASES
SUMMARY
is the same curve NRC intends to use in Rev. 1 of
NUREG/CR-6909). However, Item (d) below must also be
considered when evaluating the adequacy of Fen usage
factors calculated using the design curve.
d. Stainless Steel Fen Expression: Equation (2) of the code
case uses the stainless steel Fen expression from the initial
revision of NUREG/CR-6909. This equation is different
than the Fen expression recently developed by the NRC,
and the equations for the transformed environmental
parameters are different, and the Fen equation may yield
non-conservative values of Fen in cases where the average
temperature is used (see Item (f)).
e. Ni-Cr-Fe Steel: The same observations under Item (c)
applies for Ni-Cr-Fe steels since the stainless steel fatigue
curve is used for Ni-Cr-Fe materials. Equation (3) of the
code case uses the Ni-Cr-Fe steel Fen expression from the
initial revision of NUREG/CR-6909. This equation is the
same as the Fen expression recently developed by the
NRC, but the equations for the transformed environmental
parameters are different, and the Fen equation may yield
non-conservative values of Fen in cases where the average
temperature is used (see Item (f)).
f. -2420 Determination of Transformed Temperature:
a. -2421 of the code case states that the transformed
temperature is based on, “…the average of the
highest and lowest metal temperatures of the
surface in contact with the fluid in the transients
constituting the stress cycle…” NRC disagrees
with this approach as it is not consistent with the
Fen methodology, and it can be non-conservative.
i. To be consistent with the Fen
methodology, an average temperature for
the transient should consider the threshold
temperature to estimate Fen during a load
cycle, which may be significantly higher
than the minimum temperature of the
transient.
ii. Limited NRC calculations indicate that
using either an average transient
temperature or an average of the transient
maximum temperature and the Fen
threshold temperature does not always
yield a conservative Fen estimate when
compared to the results obtained from an
integrated Fen using the modified rate
approach.

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DATE OR
SUPPLEMENT/
EDITION

CODE CASE
NUMBER

TABLE 1
UNACCEPTABLE SECTION III CODE CASES
SUMMARY

DATE OR
SUPPLEMENT/
EDITION

b. -2422 defines the transformed temperature for
carbon and low alloy steels for temperatures up to
350°C (660°F). NRC’s updated research only
includes data up to 325°C (615°F), so the updated
Fen expression for carbon and low alloy steels is
only applicable for temperatures up to 325°C.
c. -2423 defines the transformed temperature for
wrought and cast austenitic stainless steels for
temperatures above 325°C (615°F) as constant (T*
= 1). NRC’s updated research only includes data
up to 325°C (615°F), and the updated Fen
expression for wrought and cast austenitic stainless
steels does not plateau at temperatures above
325°C. Therefore, the code case may provide nonconservative estimates of Fen for temperatures
above 325°C.
g. -2424 defines the transformed temperature for Ni-Cr-Fe
steels for temperatures above 325°C (615°F) as constant
(T* = 1). NRC’s updated research only includes data up to
325°C (615°F), and the updated Fen expression for Ni-CrFe steels does not plateau at temperatures above 325°C.
Therefore, the code case may provide non-conservative
estimates of Fen for temperatures above 325°C.
NRC recommends that Code Case N-792-1 be revised to reflect
NUREG/CR-6909 Rev. 1 after it is published.
The NRC staff abstained from voting on this item at Standards
Committee and commented that the staff does not support the Code
Case based on NRC sponsored research that is on-going.
N-793

Extruded Steel Sleeves With Parallel Threaded Ends, Section III,
Division 2
See comments for N-791.

9/20/10

N-794

Swaged Splice With Threaded Ends, Section III
See comments for N-791.

9/20/10

N-796

Alternative Preheat Temperature for Austenitic Welds in P-No. 1
Material Without PWHT, Section III, Division 1
See comments for N-791.

10/18/10

N-804

Alternative Preheat Temperature for Austenitic Welds in P-No. 1
Material Without PWHT, Section III, Division 1
The NRC believes that the test data provided is insufficient to
support a reduction in the ASME Code required preheat of 200°F.
Data for the welds in the production valve bodies tested indicate
the presence of martensite resulting in unacceptably high hardness

10/14/11

DG-1344, Page 16

CODE CASE
NUMBER

TABLE 1
UNACCEPTABLE SECTION III CODE CASES
SUMMARY

DATE OR
SUPPLEMENT/
EDITION

values. Hydrogen cracking of the welds could result in the absence
of proper preheat.
N-807

Use of Grades 75 and 80 Reinforcement in Concrete
Containments, Section III, Division 2
The NRC considers the higher grades of steel to be unacceptable
for the reinforcement of containment construction. Higher grades
will reduce the ductility of the steel reinforcement, and thus the
overall ductility of the containment, which is undesirable.

4/20/11

N-812
N-812-1

Alternate Creep-Fatigue Damage Envelope for 9Cr-1Mo-V Steel,
Section III, Division 1
Code Case N-812 utilizes Section III, Division, Subsection NH,
“Class 1 Components in Elevated Temperature Service.”
Subsection NH is not approved for use by the NRC.

8/5/11
1/13E

N-818
N-818-1

Alternative Requirements for Preservice Volumetric and Surface
Examination, Section III, Division 1
Code Case N-818-1 contains provisions for applying the results of
nondestructive examinations and fracture mechanics calculations to
accept flaws in full penetration butt welds of ferritic vessels and
austenitic and ferritic piping in lieu of repair in accordance with the
ASME Code, Section III, when the radiography indicates that the
welds cannot satisfy NB-5000 or NC-5000 of Section III during
preservice examinations.
The NRC staff has following concerns regarding the provisions of
UT and other issues in this code case.

12/6/11
7/13E

1. The code case applies to ferritic, austenitic stainless steel, and
dissimilar metal welds. However, UT in lieu of radiograph testing
(RT), at this time, has only been qualified as described ASME
Code Case N-831) for ferritic materials. The NRC staff has
reviewed and approved relief requests for UT in lieu of RT that
utilized the qualification approach described in CC N-831 for
ferritic materials only. To date, the technical basis for the use of
UT in lieu of RT for austenitic welds has not been sufficiently
developed to allow The NRC staff to accept UT in lieu of RT on
austenitic stainless steel or dissimilar metal welds.
2. Single side access in not acceptable for fabrication examinations
because some flaws are only detectable from one direction.
3. Second leg of UT V-path may be acceptable to use on a limited
basis for ferritic material, but will not be acceptable for austenitic
stainless steel or dissimilar metal welds.

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CODE CASE
NUMBER

TABLE 1
UNACCEPTABLE SECTION III CODE CASES
SUMMARY

DATE OR
SUPPLEMENT/
EDITION

4. Surface preparation needs to be addressed. Welds must be
conditioned without any gap more than 1/32-inch between
transducer and weld.
5. Paragraph (g) of the code case seems to be the discussion of a
calibration block, not a qualification block.
6. Paragraph I-3.2(d) states that “…Examination procedures,
equipment, and personnel are qualified for depth-sizing when the
RMS [root-mean-square] error of the flaw depth measurements, as
compared to the true flaw depths, does not exceed 0.125 in. (3
mm)…” The RMS error was meat for depth sizing of serviceinduced surface connected flaws. The NRC staff does not find
using this RMS error is appropriate for measurements of
fabrication defects.
7. The location of the fabrication defect is important in that if the
fabrication defect is located closer to the inside surface vs outside
surface of the pipe.
8. The depth of the maximum flaw permitted by the code case for
the preservice examination is 20 percent through wall. The concept
of such fabrication defect permitted to remain in the component
prior to service is contrary to the fundamental design philosophy of
ASME Code, Section III which is that a component is not designed
to have flaws. In addition, the allowable limits for primary and
secondary stresses and cumulative fatigue usage factors in NB3000 and NC-3000 are based on a component without flaws.
9. Permitting a 20 percent depth flaw to remain in a component
prior to service reflects a tacit approval of a lower quality of the
product and subpar workmanship.
N-820

Twisting of Horizontal Prestressing Tendons, Section III, Division
2
New reactor designs will utilize stranded wire sizes up to 0.6 inch.
The Office of New Reactors will determine the appropriate
regulatory approach for approving Code Case N-820 through the
licensing process.

12/6/11

N-828

Alternative Nonmetallic Material Manufacturer’s and Constituent
Suppliers Quality System Program Requirements, Section III,
NCA-3900, 2010 Edition, and Earlier Editions and Addenda,
Section III, Divisions 1 and 2

4/27/12

DG-1344, Page 18

CODE CASE
NUMBER

TABLE 1
UNACCEPTABLE SECTION III CODE CASES
SUMMARY

DATE OR
SUPPLEMENT/
EDITION

Code Case N-828 was developed to support new nuclear plant
construction. The NRC plans to address this Code Case in
Regulatory Guide 1.136, “Design Limits, Loading Combinations,
Materials, Construction, and Testing of Concrete Containments.”
N-837

Alternative to the Registered Professional Engineer Requirements,
Section III, Divisions 1, 2, 3, and 5
This Code Case is only for non-U.S. nuclear facilities, and
therefore, is not applicable to U.S. nuclear facilities regulated by
the U.S. NRC.

DG-1344, Page 19

3/13E

2.

Unacceptable Section XI Code Cases

The NRC determined that the following Section XI Code Cases are unacceptable for use
by licensees in their Section XI inservice inspection programs. To assist users, new Code Cases
are shaded to distinguish them from those listed in previous versions of this guide. The shading will assist
in focusing attention during the public comment period on the changes to the guide.
Table 2. Unacceptable Section XI Code Cases
CODE CASE
NUMBER

TABLE 2
UNACCEPTABLE SECTION XI CODE CASES
SUMMARY

DATE OR
SUPPLEMENT/
EDITION

N-465
N-465-1

Alternative Rules for Pump Testing, Section XI, Division 1
The draft standard referenced in the Code Case is outdated.
The requirements contained in the OM Code, “Code for Operation
and Maintenance of Nuclear Power Plants,” should be used.

11/30/88
Annulled
2/14/03

N-473
N-473-1

Alternative Rules for Valve Testing, Section XI, Division 1
The draft standard referenced in the Code Case is outdated.
The requirements contained in the OM Code, “Code for Operation and
Maintenance of Nuclear Power Plants,” should be used.

3/8/89
Annulled
2/14/03

N-480

Examination Requirements for Pipe Wall Thinning Due to Single Phase
Erosion and Corrosion, Section XI, Division 1
Code Case has been superseded by Code Case N-597, “Requirements for
Analytical Evaluation of Pipe Wall Thinning,” implemented
in conjunction with NSAC-202L, “Recommendations for an Effective
Flow-Accelerated Corrosion Program” (Ref. 14).

N-498-2
N-498-3

Alternative Requirements for 10-Year System Hydrostatic Testing
for Class 1, 2, and 3 Systems, Section XI, Division 1

6/9/95
5/20/98

N-532-2

Alternative Requirements to Repair and Replacement Documentation
Requirements and Inservice Summary Report Preparation and
Submission as Requested by IWA-4000 and IWA-6000, Section XI,
Division 1
The following concerns were identified during review of the Code Case:

7/23/02

(1) The Code Case references new paragraph IWA-6350, which has not
yet been incorporated into the ASME Code.
(2) NRC staff had difficulty reconciling Footnote 1 and Table 4
regarding the applicable edition and addenda.
(3) Submission of Form OAR-1 is at the end of each inspection
period, rather than 90 days following the outage.

DG-1344, Page 20

Annulled 9/18/01

CODE CASE
NUMBER

TABLE 2
UNACCEPTABLE SECTION XI CODE CASES
SUMMARY

DATE OR
SUPPLEMENT/
EDITION

N-542

Alternative Requirements for Nozzle Inside Radius Section Length Sizing
Performance Demonstration, Section XI, Division 1
Code Case N-542 was subsumed by Code Case N-552,
“Alternative MethodsBQualification for Nozzle Inside Radius Section
from the Outside Surface,” which is being implemented by licensees.
Thus, there is no need to approve Code Case N-542.

Annulled 3/28/01

N-547

Alternative Examination Requirements for Pressure Retaining Bolting of
Control Rod Drive (CRD) Housings, Section XI, Division 1
Code Case N-547 states that the examination of CRD housing bolts,
studs, and nuts is not required. However, 10 CFR 50.55a(b)(2)(xxi)(B)
required the examination of CRD bolting material whenever the CRD
housing is disassembled and the bolting material is to be reused.
Examination of CRD bolting material is required to verify that servicerelated degradation has not occurred, or that damage such as bending and
galling of threads has not occurred when performing maintenance
activities that require the removal and reinstallation of bolting.

Annulled 5/20/01

N-560
N-560-1
N-560-2

Alternative Examination Requirements for Class 1, Category B-J Piping
Welds, Section XI, Division 1
(1) The Code Case does not address inspection strategy for existing
augmented and other inspection programs such as intergranular
stress corrosion cracking (IGSCC), flow-assisted corrosion (FAC),
microbiological corrosion (MIC), and pitting.
(2) The Code Case does not provide system-level guidelines for change
in risk evaluation to ensure that the risk from individual system
failures will be kept small and dominant risk contributors will not be
created.

8/9/96
2/26/99
2/14/03

N-561
N-561-1

Alternative Requirements for Wall Thickness Restoration of Class 2 and
High Energy Class 3 Carbon Steel Piping, Section XI, Division 1
Neither the ASME Code nor the Code Case have criteria for determining
the rate or extent of degradation of the repair or the surrounding base
metal. Reinspection requirements are not provided to verify structural
integrity since the root cause may not be mitigated.

12/31/96
3/28/01

N-562
N-562-1

Alternative Requirements for Wall Thickness Restoration of Class 3
Moderate Energy Carbon Steel Piping, Section XI, Division 1
Neither the ASME Code nor the Code Case have criteria for determining
the rate or extent of degradation of the repair or the surrounding base
metal. Reinspection requirements are not provided to verify structural
integrity since the root cause may not be mitigated.

12/31/96
3/28/01

N-574

NDE Personnel Recertification Frequency, Section XI, Division 1
Based on data obtained by the NRC staff during its review
of Appendix VIII, “Performance Demonstration for Ultrasonic
Examination Systems,” to Section XI, the NRC staff noted that
proficiency decreases over time. The data do not support recertification
examinations at a frequency of every 5 years.

DG-1344, Page 21

Annulled 7/14/06

CODE CASE
NUMBER

TABLE 2
UNACCEPTABLE SECTION XI CODE CASES
SUMMARY

DATE OR
SUPPLEMENT/
EDITION

N-575

Alternative Examination Requirements for Full Penetration
Nozzle-to-Vessel Welds in Reactor Vessels with Set-On Type Nozzles,
Section XI, Division 1
The supporting basis for the Code Case applies to the specific
configuration of one plant and is not applicable on a generic basis.
In addition, there are insufficient controls on stress and operating
conditions to permit a generic reduction in examination volume. Finally,
the boundaries of the volume of the weld, cladding, and heat affected
zone from Figure 2 are ambiguous.

2/14/03

N-577
N-577-1

Risk-Informed Requirements for Class 1, 2, and 3 Piping, Method A,
Section XI, Division 1

9/2/97
2/14/03

(1) The Code Case does not address inspection strategy for existing
augmented and other inspection programs such as IGSCC, FAC,
MIC, and pitting.
(2) The Code Case does not provide system-level guidelines for change
in risk evaluation to ensure that the risk from individual system
failures will be kept small and dominant risk contributors will not be
created.
9/2/97
2/14/03

N-578
N-578-1

Risk-Informed Requirements for Class 1, 2, and 3 Piping, Method B,
Section XI, Division 1
(1) The Code Case does not address inspection strategy for existing
augmented and other inspection programs such as IGSCC, FAC,
MIC, and pitting.
(2) The Code Case does not provide system-level guidelines for change
in risk evaluation to ensure that the risk from individual system
failures will be kept small and dominant risk contributors will not be
created.

N-587

Alternative NDE Requirements for Repair/Replacement Activities,
Section XI, Division 1
The NRC believes this Code Case is in conflict with the review process
for approval of alternatives under 10 CFR 50.55a(z). The Code Case
would permit a licensee and the Authorized Nuclear Inspector to choose
unspecified alternatives to regulatory requirements.

Annulled
2/14/03

N-589
N-589-1

Class 3 Nonmetallic Cured-in-Place Piping, Section XI, Division 1
(1) The installation process provides insufficient controls on wall
thickness measurement.
(2) There are no qualification requirements for installers and installation
procedures such as those for welders and welding procedures.
(3) Fracture toughness properties of the fiberglass are such that the
cured-in-place piping (CIPP) could crack during a seismic event.
(4) Equations 4 and 5 in the Code Case contain an “i” term [a stress
intensification factor] that is derived from fatigue considerations.
Stress intensification factors, however, have not been developed for
fiberglass materials.

4/19/02
7/23/02

DG-1344, Page 22

CODE CASE
NUMBER

TABLE 2
UNACCEPTABLE SECTION XI CODE CASES
SUMMARY

DATE OR
SUPPLEMENT/
EDITION

N-590

Alternative to the Requirements of Subsection IWE, Requirements
for Class MC and Metallic Liners of Class CC Components
of Light-Water Cooled Plants, Section XI, Division 1
The provisions of the Code Case were incorporated into the 1998 Edition,
which has been approved by the NRC. Thus, the Code Case is no longer
needed and was annulled by the ASME.

Annulled 4/8/02

N-591

Alternative to the Requirements of Subsection IWL, Requirements for
Class CC Concrete Components of Light-Water Cooled Plants,
Section XI, Division 1
The provisions of the Code Case were incorporated into the 1998 Edition
which has been approved by the NRC. Thus, the Code Case is no longer
needed and was annulled by the ASME.

Annulled 4/8/02

N-593-1

Examination Requirements for Steam Generator Nozzle-to-Vessel Welds,
Section XI, Division 1
The Code Case eliminates the requirement to examine the steam
generator nozzle inner radius. Specifically, the examination volume for
the nozzle inner radius was removed from Section XI, Figures IWB-25007(a) through IWB-2500-7(d). The action is applicable from the 1974
Edition through the 2004 Edition with the 2005 Addenda. A similar
action was taken regarding Code Case N-619. The NRC did not take
exception to Code Case N-619 because 10 CFR 50.55a(b)(2)(xxi)(A)
required licensees to perform the examination in accordance with the
1998 Edition, which includes figures containing the examination volume.
However, Code Case N-593-1 applies to editions prior to the 1998
Edition which do not have the appropriate figures.

10/08/04

N-613

Ultrasonic Examination of Full Penetration Nozzles in Vessels,
Examination Category B-D, Item No’s. B3.10 and B3.90, Reactor VesselTo-Nozzle Welds, Fig. IWB-2500-7(a), (b), and (c), Section XI, Division 1
The Code Case conflicts with and unacceptably reduced the requirements
of 10 CFR 50.55a(b)(2)(xv)(K)(2)(i). A revision to the Code Case has
been developed to address the concerns.

7/30/98

N-615

Ultrasonic Examination as a Surface Examination Method for Category
B-F and B-J Piping Welds, Section XI, Division 1
The Code Case requires that the ultrasonic technique used be
demonstrated capable of detecting certain size flaws on the outside
diameter of the weld, but it does not specify any demonstration
requirements. To be acceptable, Section XI, Appendix VIII, type rules for
performance demonstration need to be developed and applied.

7/28/01

N-618

Use of a Reactor Pressure Vessel as a Transportation Containment
System, Section XI, Division 1
The Code Case was developed as a potential option for shipping
and disposal of a reactor pressure vessel (RPV). The NRC staff
determined, however, that the Code Case was not applicable to the review
and approval process for transportation packages. The use of RPVs as a
transportation package has been addressed under 10 CFR Part 71,
“Packaging and Transportation of Radioactive Material” (Ref. 15).

6/17/03

DG-1344, Page 23

CODE CASE
NUMBER

TABLE 2
UNACCEPTABLE SECTION XI CODE CASES
SUMMARY

DATE OR
SUPPLEMENT/
EDITION

N-622

Ultrasonic Examination of RPV and Piping, Bolts, and Studs,
Section XI, Division 1
The Code Case was published in May 1999. Industry Performance
Demonstration Initiative efforts since that time have made this Code Case
obsolete. Issues associated with supplements to Appendix VIII are being
addressed individually in separate Code Cases.

Annulled on
1/12/05

N-653

Qualification Requirements for Full Structural Overlaid Wrought
Austenitic Piping Welds, Section XI, Division 1
(1) Section XI, Appendix VIII, Supplement 11, requires a personnel
performance qualification as part of the procedure qualification. The
detection acceptance criteria in the Code Case do not require
personnel performance qualification as part of the procedure
qualification. Personnel qualification is necessary to validate
the effectiveness of the procedure qualification.
(2) The minimum grading unit is 1.0 inch in the circumferential
direction. The acceptance tolerance, however, is 0.75 inch root mean
square error. Thus, the length sizing acceptance criteria do not
adequately prevent the use of testmanship rather than skill to pass
length sizing tests.

9/7/01

N-654

Acceptance Criteria for Flaws in Ferritic Steel Components 4 in. and
Greater in Thickness, Section XI, Division 1
Licensees intending to apply the rules of this Code Case must obtain NRC
approval of the specific application in accordance with 10 CFR 50.55a(z).

4/17/02

N-691

Application of Risk-Informed Insights to Increase the Inspection Interval
for Pressurized Water Reactor Vessels, Section XI, Division 1
A response to the NRC staff=s request for additional information has not
yet been received and therefore, insufficient information has been
provided for the staff to make a determination relative to the acceptability
of this Code Case.

11/18/03

N-711

Alternative Examination Coverage Requirements for Examination
Category B-F, B-J, C-F-1, C-F-2, and R-A Piping Welds, Section XI,
Division 1
The Code Case would permit each licensee to independently determine
when achievement of a coverage requirement is impractical, and when
Code-required coverage is satisfied. As a result, application of the Code
Case for similar configurations at different plants could result in
potentially significant quantitative variations. Furthermore, application of
the Code Case is inconsistent with NRC’s responsibility for determining
whether examinations are impractical, and eliminates the NRC’s ability to
take exception to a licensee’s proposed action and impose additional
measures where warranted in accordance with 10 CFR 50.55a(g)(6)(i).

1/05/06

DG-1344, Page 24

CODE CASE
NUMBER

TABLE 2
UNACCEPTABLE SECTION XI CODE CASES
SUMMARY

DATE OR
SUPPLEMENT/
EDITION

N-713

Ultrasonic Examination in Lieu of Radiography, Section XI, Division 1
The requirements of Code Case N-713 were based largely on the
requirements contained in Code Case N-659. The NRC has not approved
Code Cases N-659, N-659-1, nor N-659-2. Refer to the discussion on
Code Case N-659-2 in Table 1 above, “Unacceptable Section III Code
Cases,” for more information.

11/10/08

N-716

Alternative Piping Classification and Examination Requirements,
Section XI, Division 1
The NRC has approved risk-informed inservice inspection (RI-ISI)
programs based, in part, on methods described in Code Case N-716. The
NRC has approved programs for Grand Gulf Nuclear Station 1
(September 21, 2007; ML072430005), Donald C. Cook Nuclear Plant
(September 28, 2007; ML072620553), and Waterford Steam Electric
Station (April 28, 2008; ML080980120). The approvals were specific to
these units and relied on several changes to the methodology described in
Code Case N-716. The NRC is reviewing EPRI Topical Report 1021467,
“Nondestructive Evaluation: Probabilistic Risk Assessment Technical
Adequacy Guidance for Risk-Informed In-service Inspection Programs.”
The purpose of the topical report, in part, is to provide guidance on
determining the technical adequacy of probabilistic risk assessments used
to develop a “streamlined” RI-ISI program in accordance with Code Case
N-716. The staff will consider the revised Code Case for generic approval
when its review of the topical report has been completed.

4/10/06

N-722-2

Visual Examinations for PWR Pressure Retaining Welds in Class 1
Components Fabricated With Alloy 600/82/182 Materials, Section XI,
Division 1
Code Case N-722 has been superseded by Revisions 1 and 2 to the Code
Case. N-722-1 is conditionally approved directly in 10 CFR 50.55a and
not through Regulatory Guide 1.147. Code Case N-722-2 has been
dispositioned as Unacceptable.

9/8/11

N-729-3
N-729-4

Alternative Examination Requirements for PWR Reactor Vessel Upper
Heads With Nozzles Having Pressure-Retaining Partial-Penetration
Nozzles, Section XI, Division 1
Code Case N-729 has been superseded by Revisions 1, 2, and 3 to the
Code Case. N-729-1 is conditionally approved directly in 10 CFR 50.55a
and not through Regulatory Guide 1.147. Code Case N-729-4 is
addressed directly in 10 CFR 50.55a.

4/4/12
11/10E

DG-1344, Page 25

CODE CASE
NUMBER
N-740
N-740-1
N-740-2

TABLE 2
UNACCEPTABLE SECTION XI CODE CASES
SUMMARY

DATE OR
SUPPLEMENT/
EDITION

Dissimilar Metal Weld Overlay for Repair of Class 1, 2, and 3 Items,
Section XI, Division 1
The NRC staff identified many technical issues regarding the provisions
of Revisions 0 and 1. The issues were communicated to the cognizant
Section XI committees, and the staff continues to work with the
committees to resolve the issues. Due to the total number of issues and
the nature of some (e.g., lack of certain fundamental design details), the
staff determined that it would be inappropriate to attempt to conditionally
approve either version 0 or 1 in Regulatory Guide 1.147.

10/12/06
12/25/09
11/10/08

Code Case N-740-2 has been approved and published by the ASME.
While Revision 2 addresses some of the NRC staff concerns, significant
issues remain. For example, the definition of nominal weld and base
material appear to be inconsistent with the provisions of Section III. Also,
additional detail is required on how to perform the flaw growth or design
analysis. Finally, additional detail is required on how the overlays are
designed.
N-766

Nickel Alloy Reactor Coolant Inlay and Onlay for Mitigation of PWR Full
Penetration Circumferential Nickel Alloy Dissimilar Metal Welds of
Class 1 Items, Section XI, Division 1
(1) Paragraph 1.(c)(1) of Code Case N-766 would potentially allow a
75-percent through wall flaw to remain in service in the original
Alloy 82/182 dissimilar metal weld, in accordance with IWB-3600.
The NRC staff finds it is unacceptable to allow such a large flaw to
remain in service in Class 1 piping.
(2) Paragraphs 2.(c)(1) and 2.(c)(2) of Code Case N-766: The postulated
and as-left flaws need to be evaluated because the postulated flaws
are supposed to represent the capabilities of the non-destructive
examination techniques applied. For example, if a 15-degree
circumferential flaw that is 11% through-wall is detected, this would
be evaluated instead of a 360-degree, 10% through-wall flaw. A
360-degree, 10% through-wall flaw should be analyzed to determine
the fatigue and stress corrosion cracking degradation mechanisms.
(3) Paragraph 2.(f) of Code Case N-766 should be revised to include the
following: “The flaw growth calculation due to stress corrosion
cracking should include the welding residual stresses. The flaw
growth calculation shall be performed in accordance with IWB-3640
and/or Appendix C to the ASME Code, Section XI.”

12/20/10

N-770-3
N-770-4

Alternative Examination Requirements and Acceptance Standards for
Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated With UNS
N06082 or UNS W86182 Weld Filler Material With or Without
Application of Listed Mitigation Activities Section XI, Division 1
The NRC requires the N-770-2 examinations to be performed as an
augmented inspection program under 50.55a(g)(6)(ii)(F). The latest
version of N-770-2 approved by the NRC is incorporated by reference in
50.55a. Once the review of the topical report on the technical basis for
peening is complete, the staff expects to review the latest ASME Code
approved version of Code Case N-770 for incorporation directly in 10
CFR 50.55a, under § 50.55a(g)(6)(ii)(F).

1/13E
5/13E

DG-1344, Page 26

CODE CASE
NUMBER

TABLE 2
UNACCEPTABLE SECTION XI CODE CASES
SUMMARY

DATE OR
SUPPLEMENT/
EDITION

N-780

Alternative Requirements for Upgrade, Substitution, or Reconfiguration
of Examination Equipment When Using Appendix VIII Qualified
Ultrasonic Examination Systems, Section XI, Division 1
At this time, the NRC will review application of Code Case N-780 on a
case-by-case basis. The Code Case is a new alternative to the current
requirements in Section XI, Appendix VIII. The technical justification for
the alternative is based largely on the expertise of nondestructive
examination experts and laboratory testing. While the laboratory testing
was well conducted, it was not bounding. The NRC believes that industry
experience in applying the alternative is needed to ensure generic
applicability and demonstrate reliability before the alternative can be
approved in RG 1.147.

4/9/10

N-784

Experience Credit for Ultrasonic Examiner Certification
Code Case N-784 reduces the requirements for training and experience
regarding examination personnel. Examination personnel would receive
less training and experience with respect to the detection of representative
flaws in materials and configurations found in nuclear power plants. In
addition, the Code Case would allow personnel without nuclear ultrasonic
examination experience to qualify without exposure to the variety of
defects, components, examination conditions, and regulations to be
encountered. The impact of reduced training and experience has not been
evaluated.

4/9/10

N-806

Evaluation of Metal Loss in Class 2 and 3 Metallic Piping Buried in a
Back-Filled Trench
NRC staff advised ASME during consideration of Code Case N-806 that
the NRC had concerns and intended to review and approve the Code Case
on a case-by-case basis. Following are the NRC’s concerns:
(1) The rules applicable to determining corrosion rates which lead to the
definition of the evaluation period and re-examination schedules are
currently under development. Accordingly, the Code Case does not
define the method of determining the wall loss rates, the time period
for length of the evaluation, and the reexamination period/frequency.
(2) The ASME Section XI appendices used to calculate some of the
important values are nonmandatory.

6/22/12

Licensees intending to use Code Case N-806 must submit a plant-specific
request to the NRC staff for review and approval prior to implementation.

DG-1344, Page 27

CODE CASE
NUMBER
N-813

TABLE 2
UNACCEPTABLE SECTION XI CODE CASES
SUMMARY

DATE OR
SUPPLEMENT/
EDITION

Alternative Requirements for Preservice Volumetric and Surface
Examination, Section XI, Division 1
Code Case N-813 is an alternative to the provisions of the 2010 Edition of
the ASME Code, Section XI, paragraph IWB-3112. IWB-3112 does not
allow the acceptance of flaws detected in the preservice examination by
analytical evaluation. Code Case N-813 would allow the acceptance of
these flaws through analytical evaluation. Per paragraph IWB-3112, any
preservice flaw that exceeds the acceptance standards of Table IWB3410-1 must be removed. While it is recognized that operating experience
has shown that large through wall flaws and leakages have developed in
previously repaired welds as a result of weld residual stresses, the NRC
has the following concerns regarding the proposed alternative in Code
Case N-813:

10/24/11

(1) The requirements of paragraph IWB-3112 were developed to ensure
that defective welds were not placed in service. A preservice flaw
detected in a weld that exceeds the acceptance standards of Table
IWB-3410-1 demonstrates poor workmanship and/or inadequate
welding practice and procedures. The unacceptable preservice flaw
needs to be removed and the weld needs to be repaired before it is
placed in service.
(2) Under Code Case N-813, large flaws would be allowed to remain in
service because paragraph IWB-3132.3, via paragraph IWB-3643,
allows a flaw up to 75 percent through wall to remain in service.
Larger flaws could grow to an unacceptable size between
inspections reducing structural margin and potentially challenging
the structural integrity of safety-related Class 1 and Class 2 piping.
Paragraph C-3112(a)(3) of Code Case N-813 provides the same
alternatives for Class 2 piping as that of Paragraph B-3122(a)(3). The
staff has the same concerns for Class 2 piping as for Class 1 piping.

DG-1344, Page 28

CODE CASE
NUMBER

TABLE 2
UNACCEPTABLE SECTION XI CODE CASES
SUMMARY

DATE OR
SUPPLEMENT/
EDITION

N-826

Ultrasonic Examination of Full Penetration Vessel Weld Joints in Fig.
IWB-2500-1 Through Fig. IWB-2500-6
Reduction of the inspection volume from ½ t to ½ inch is in conflict with
10 CFR 50.61a, “Alternate Fracture Toughness Requirements for
Protection against Pressurized Thermal Shock Events.” Licensees
implementing 10 CFR 50.61a must first examine the volume described in
the ASME Code, Section XI, Figures IWB-2500-1 and IWB-2500-2 using
Appendix VIII qualified procedures, equipment, and personnel to obtain
the necessary data on flaws to ensure the flaw density requirements of 10
CFR 50.61a are met. Although under Code Case N-826, a licensee would
have examined the full ½ t volume at least once in accordance with
Appendix VIII, the NRC staff finds it unacceptable to allow reduction of
the examination volume for later inservice examinations due to concerns
about detection and sizing accuracy for smaller flaws using the current
UT technology. Current UT technology cannot reliably detect and
accurately size smaller flaws which affects the validity of the comparison
with the flaw density requirement of 10 CFR 50.61a. In addition, recent
experiences at operating plants regarding missed defects during
examinations using qualified methods and conducted in compliance with
Section XI, Appendix VIII, has raised concerns regarding the reliability
of ultrasonic examinations. Finally, the reduction from ½ t to ½ inch
originated with Code Case N-613. The purpose of the reduction in
examination volume was to reduce the number of relief requests caused
by the inability to examine the required volume for typical geometries of
nozzle-to-vessel welds. The full-penetration vessel welds addressed by
Code Case N-826 do not generally have similar geometric restrictions that
would prevent examination of the full ½ t volume.

7/16/12

N-840

Cladding Repair by Underwater Electrochemical Deposition in Class 1
and 2 Applications Section XI, Division 1
This code case was developed specifically to address erosion/corrosion
concerns in a Korean nuclear facility, in a location where cladding
damage in the RPV has exposed low alloy steels. If this were to occur in a
US nuclear facility the NRC staff would want to review the particular
circumstances on a case by case basis. Any licensee desiring to utilize
Code Case N-840 should submitted it for review and approval in
accordance with 10 CFR 50.55a(z).

4/13E

DG-1344, Page 29

3.

Unacceptable OM Code Cases

The following OM Code Cases were determined to be unacceptable for use by licensees in their
inservice testing programs. The ASME issues OM Code Cases annually with publication of a new edition
or addenda. To assist users, new and revised Code Cases are shaded to distinguish them from those
approved in previous versions of this guide. The shading will assist in focusing attention during the public
comment period on the changes to the guide.
Table 3. Unacceptable OM Code Cases
CODE CASE
NUMBER

TABLE 3
UNACCEPTABLE OM CODE CASES

EDITION/
ADDENDA

SUMMARY OF BASIS FOR EXCLUSION
OMN-10

Requirements for Safety Significance Categorization of Snubbers Using Risk
Insights and Testing Strategies for Inservice Testing of LWR Power Plants
The method used for categorizing snubbers could result in certain snubbers
being inappropriately categorized as having low safety significance. These
snubbers would not be adequately tested or inspected to provide assurance of
their operational readiness. In addition, unexpected extensive degradation in
feedwater piping has occurred which would necessitate a more rigorous
approach to snubber categorization than presently contained in this Code Case.
Note: Pages C-31 through C-34 were not included in the 2006 Addenda.

DG-1344, Page 30

2000 Addenda
Reaffirmed
2001 Edition
Reaffirmed
2003 Addenda
Reaffirmed
2004 Edition
Reaffirmed
2006 Addenda
(see Note)
Reaffirmed
2009 Edition
Reaffirmed
2012 Edition
Reaffirmed
2015 Edition
Reaffirmed
2017 Edition

CODE CASE
NUMBER

TABLE 3
UNACCEPTABLE OM CODE CASES

EDITION/
ADDENDA

SUMMARY OF BASIS FOR EXCLUSION
OMN-15

Requirements for Extending the Snubber Operational Readiness Testing
Interval at LWR Power Plants
Following is a summary of the issues that have been identified:
(1) The basis for the snubber degradation rate that is assumed in the White
Paper for the Code Case is not clear.
(2) The Code Case does not address snubber service life monitoring
requirements when using the 1995 Edition of the OM Code.
(3) The Code Case does not address the assignment of unacceptable snubbers
in the Failure Mode Group.
(4) The Code Case does not address treatment of isolated snubber failures.
(5) The Code Case does not address how unacceptable snubbers are
accounted for during the extended test interval. For example, unacceptable
snubbers could be indentified during maintenance, service life monitoring, and
visual examination activities conducted during the extended test interval.
Note: OMN-15 Revision 2 (2017 Edition) is approved for use in Regulatory.
Guide 1.192 Revision 3.

DG-1344, Page 31

2004 Edition
Revised 2006
Addenda
Reaffirmed
2009 Edition
Reaffirmed
2012 Edition

D. IMPLEMENTATION
The purpose of this section is to provide information to applicants and licensees regarding the NRC
staff’s plans for using this regulatory guide. This regulatory guide does not approve the use of the Code
Cases listed herein. Applicants or licensees may submit a plant-specific request to implement one or more
of the Code Cases listed in this regulatory guide. The request should address the NRC’s concerns about
the Code Case at issue.

DG-1344, Page 32

REFERENCES1
1. Code of Federal Regulations (CFR), Title 10, Energy, Part 50, “Domestic Licensing of Production
and Utilization Facilities” (10 CFR Part 50), U.S. Nuclear Regulatory Commission, Washington, DC.
2. CFR, Title 10, Energy, Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,”
U.S. Nuclear Regulatory Commission, Washington, DC
3. Regulatory Guide (RG) 1.84, “Design, Fabrication, and Materials Code Case Acceptability, ASME
Section III,” U.S. Nuclear Regulatory Commission (NRC), Washington, DC.
4. RG 1.147, “Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1,” NRC,
Washington, DC.
5. RG 1.192, “Operation and Maintenance Code Case Acceptability, ASME OM Code,” NRC,
Washington, DC.
6. ASME Operation and Maintenance of Nuclear Power Plants, American Society of Mechanical
Engineers (ASME), New York, NY.2
7. NRC Spent Fuel Project Office Interim Staff Guidance No. 4 (ISG-4), Revision 1, “Cask Closure
Weld Inspections,” NRC, Washington, DC (ADAMS Accession No. ML051520313).
8. NRC Spent Fuel Storage and Transportation Division Interim Staff Guidance No. 18 (ISG-18)
Revision 1, “The Design and Testing of Lid Welds on Austenitic Stainless Steel Canisters as
Containment Boundary for Spent Fuel Storage,” NRC, Washington, D.C. (ADAMS Accession No.
ML031250620).
9.

CFR, Title 10, Energy, Part 72, “Licensing Requirements for the Independent Storage of Spent
Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater than Class C Waste”
(10 CFR Part 72), NRC, Washington, DC.

10. NUREG/CR-6909, Revision 1, “Effect of LWR Coolant Environments on the Fatigue Life of
Reactor Materials.” NRC, Washington, DC.
11. ASTM A 1034/A1034M-05b, “Standard Test Methods for Testing Mechanical Splices for Steel
Reinforcing Bars,” ASTM International, West Conshohocken, PA.3

1

Publicly available NRC published documents are available electronically through the NRC Library on the NRC’s
public Web site at http://www.nrc.gov/reading-rm/doc-collections/ and through the NRC’s Agencywide Documents
Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html The documents can also be
viewed online or printed for a fee in the NRC’s Public Document Room (PDR) at 11555 Rockville Pike, Rockville,
MD. For problems with ADAMS, contact the PDR staff at 301-415-4737 or (800) 397-4209; fax (301) 415-3548; or email pdr.resource@nrc.gov.

2

Copies of American Society of Mechanical Engineers (ASME) standards may be purchased from ASME, Two Park
Avenue, New York, New York 10016-5990; Telephone (800) 843-2763. Purchase information is available through the
ASME Web site store at http://www.asme.org/Codes/Publications/.

3

The American Society for Testing and Materials (ASTM) is now know as ASTM International. Their standards may
be purchased from ASTM International, 100 Barr Harbor Drive, P.O. Box C700, West Conshohocken, Pennsylvania

DG-1344, Page 33

12. ACI 349-06, “Code Requirements for Nuclear Safety-Related Concrete Structures & Commentary,”
American Concrete Institute, Farmington Hills, MI. 4
13. ACI 318, “Building Code Requirements for Structural Concrete and Commentary,” American
Concrete Institute, Farmington Hills, MI.4
14. “Recommendations for an Effective Flow-Accelerated Corrosion Program” (NSAC-202L-R3),
Electric Power Research Institute, Palo Alto, CA.5
15. CFR, Title 10, Energy, Part 71, “Packaging and Transportation of Radioactive Material” (10 CFR
Part 71), U.S. Nuclear Regulatory Commission, Washington, DC.

19428-2959; telephone (877) 909-2786. Purchase information is available through the ASTM Web site at
http://www.astm.org.
4

Documents from the American Concrete Institute (ACI) are available from their bookstore Web site
(http://www.concrete.org/BookstoreNet/bookstore.htm); or by contacting the corporate office at ACI, P.O. Box 9094,
Farmington Hills, MI 48333; telephone (248) 848-3700, fax (248) 848-3701.

5

Copies of Electric Power Research Institute (EPRI) documents may be obtained by contacting the Electric Power
Research Institute, 3420 Hillview Avenue, Palo Alto, CA 94304, Telephone: 650-855-2000 or on-line at
http://my.epri.com/portal/server.pt.

DG-1344, Page 34