Regulatory Guide 1.193: Difference between revisions

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{{Adams
{{Adams
| number = ML19128A269
| number = ML16321A338
| issue date = 10/31/2019
| issue date = 08/31/2017
| title = Rev. 6, ASME Code Cases Not Approved for Use
| title = ASME Code Cases Not Approved for Use
| author name = Lin B
| author name = Facco G
| author affiliation = NRC/RES/DE
| author affiliation = NRC/RES
| addressee name =  
| addressee name =  
| addressee affiliation =  
| addressee affiliation =  
| docket =  
| docket =  
| license number =  
| license number =  
| contact person = Lin B
| contact person = Giovanni Facco, 301-415-6337
| document report number = RG-1.193, Rev. 3
| case reference number = NRC-2012-0059, RIN 3150-AJ13
| package number = ML19156A147
| document report number = RG-1.193, Rev 5
| document type = Regulatory Guidance
| package number = ML16285A003
| page count = 29
| document type = Regulatory Guide
| page count = 30
}}
}}
{{#Wiki_filter:U.S. NUCLEAR REGULATORY COMMISSION  
{{#Wiki_filter:U.S. NUCLEAR REGULATORY COMMISSION  
REGULATORY GUIDE 1.193, Revision 6
August 2017 OFFICE OF NUCLEAR REGULATORY RESEARCH
Revision 5 REGULATORY GUIDE  


Issue Date: October 2019 Technical Lead: Bruce Lin
Technical Lead Giovanni Facco


Written suggestions regarding this guide or development of new guides may be submitted through the NRCs public Web site in the NRC Library at https://nrcweb.nrc.gov/reading-rm/doc-collections/reg-guides/, under Document Collections, in Regulatory Guides, at https://nrcweb.nrc.gov/reading-rm/doc-collections/reg-guides/contactus.htm
Written suggestions regarding this guide or development of new guides may be submitted through the NRCs public Web site under the Regulatory Guides document collection of the NRC Library at http://www.nrc.gov/reading-rm/doc-collections/reg-guides/contactus.html.


====l.     ====
Electronic copies of this regulatory guide, previous versions of this guide, and other recently issued guides are available through the NRCs public Web site under the Regulatory Guides document collection of the NRC Library at http://www.nrc.gov/reading-rm/doc-collections/. 
The regulatory guide is also available through the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html, under ADAMS Accession No. ML16321A338. The regulatory basis for this guide is the regulatory analysis prepared for the amendment to 10 CFR 50.55a (ADAMS Accession No. ML16285A013. The staff responses to the public comments on DG-1298 may be found under ADAMS Accession No. ML16285A012.


Electronic copies of this RG, previous versions of RGs, and other recently issued guides are also available through the NRCs public Web site in the NRC Library at https://nrcweb.nrc.gov/reading-rm/doc-collections/reg-guides/, under Document Collections, in Regulatory Guides. This RG is also available through the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html, under ADAMS Accession Number (No.) ML19128A269. The associated draft guide DG-1344 may be found in ADAMS under Accession No. ML18114A227.  No public comments were received on DG-1344.
REGULATORY GUIDE RG 1.193 (Draft was issued as DG-1298, dated March 2016)  


ASME CODE CASES NOT APPROVED FOR USE  
ASME CODE CASES NOT APPROVED FOR USE  
Line 31: Line 34:
Purpose  
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 RG does not approve the use of the ASME Code Cases listed herein.
This regulatory guide 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.


Applicability
Applicability  


This RG applies to reactor licensees subject to 10 CFR Part 50, Section 50.55a, Codes and standards.
This RG applies to reactor licensees subject to 10 CFR Part 50, Section 50.55a, Codes and Standards.


Applicable Rules and Regulations  
Applicable Rules and Regulations  


*  
*  
Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Domestic Licensing of Production and Utilization Facilities (Ref. 1): 
Title 10, Part 50, of the Code of Federal Regulations (10 CFR Part 50), Domestic Licensing of Production and Utilization Facilities (Ref. 1), Section 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 (Ref. 2), or equivalent quality standards.


o 10 CFR 50.55a(c) 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 (Ref. 2) or equivalent quality standards.
*
Section 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) (Ref. 3), or equivalent quality standards.


o 10 CFR 50.55a(f) requires, in part, that Class 1, 2, and 3 components and their supports meet the requirements of the ASME Operation and Maintenance Code (OM Code) (Ref. 3) or equivalent quality standards.
*
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.


o 10 CFR 50.55a(g) requires, in part, that Class 1, 2, and 3 metal containment (MC),
RG 1.193, Page 2  
and concrete containment (CC) components and their supports meet the requirements of Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, of the ASME BPV Code (Ref. 4) or equivalent quality standards.
 
RG 1.193, Rev. 6, Page 2  
*
10 CFR 52.79(a)(11) (Ref. 5) requires the final safety analysis report to include a description of the program(s), and their implementation, necessary to ensure that 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  
Related Guidance  


*  
*  
RG 1.84, Design, Fabrication, and Materials Code Case Acceptability, ASME Section III  
Regulatory Guide 1.84, Design, Fabrication, and Materials Code Case Acceptability, ASME  
(Ref. 6), 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  
Section III (Ref. 4), 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.
10 CFR 50.55a.


*  
*  
RG 1.147, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1 (Ref. 7), 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  
Regulatory Guide 1.147, Inservice Inspection Code Case Acceptability, ASME Section XI,  
10 CFR 50.55a.
Division 1 (Ref. 5), 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. 8),  
Regulatory Guide 1.192, Operation and Maintenance Code Case Acceptability, ASME OM  
lists the ASME OM 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.
Code (Ref. 6), lists the OM 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  
Purpose of This Regulatory Guide  


The NRC issued the RG to provide information to applicants and licensees on 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 given 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 Case requirements referenced in 10 CFR 50.55a as long as the proposed alternatives result in an acceptable level of quality and safety. Applicants or licensees must submit a plant-specific request that addresses the NRCs concerns about the Code Case at issue. The NRC will revise this RG as needed to address subsequent new or revised Code Cases.
This regulatory guide 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 NRCs 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
Paperwork Reduction Act  


This RG provides voluntary guidance for implementing the mandatory information collections in
This regulatory guide is being issued for information only and does not contain any new or amended information collection requirements.
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),
Public Protection Notification
approval numbers 3150-0011 and 3150-0151. Send comments regarding this information collection to the Information Services Branch (T6-A10M), U.S. Nuclear Regulatory Commission, Washington, DC 20555-
0001, or by e-mail to Infocollects.Resource@nrc.gov, and to the OMB reviewer at:  OMB Office of Information and Regulatory Affairs (3150-0011 and 3150-0151), Attn: Desk Officer for the Nuclear Regulatory Commission, 725 17th Street, NW Washington, DC20503; e- mail: 
oira_submission@omb.eop.gov.


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.


The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of information unless the document requesting or requiring the collection displays a currently valid OMB
RG 1.193, Page 3  
control number.
 
RG 1.193, Rev. 6, Page 3  


==B. DISCUSSION==
==B. DISCUSSION==
Reason for Revision  
Reason of Revision  


RG 1.193, Revision 6, includes information reviewed by the NRC on the Code Cases for Sections III and XI, listed in Supplement 11 to the 2010 Edition and Supplements 0 through 7 to the  
Revision 5 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 2007 Edition and Supplements 0 through 10
2013 Edition of the ASME BPV Code, and on the OM Code Cases listed in the 2015 and the 2017 Editions. This revision updates and supersedes RG 1.193, Revision 5, which included information from Supplement 11 to the 2007 Edition and Supplements 0 through 10 to the 2010 Edition (Sections III and XI) and the 2009 Edition through the 2012 Edition of the OM Code.
to the 2010 Edition, and the OM Code Cases listed in the 2009 Edition through the 2012 Edition. This is an update to RG 1.193, Revision 4, that included information from Supplements 1 through 10 to the 2007 Edition (Sections III and XI), and the 2002 Addenda through the 2006 Addenda of the OM Code.


Background  
Background  


ASME publishes a new edition of the BPV Code every 2 years and periodically publishes a new edition of the OM Code. In 10 CFR 50.55a(a), the NRC references the latest editions and addenda of the BPV Code, Section III and Section XI, and the OM Code that the agency has approved for use by applicants and licensees. ASME also publishes Code Cases for BPV Code, Section III and Section XI,
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
quarterly and Code Cases for the OM Code biennially. Code Cases provide alternatives developed and approved by ASME.
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 Code Cases for Sections III and XI listed in Supplement 11 to the 2010
The NRC staff reviewed Section III and Section XI Code Cases listed in Supplement 11 to the 2007 Edition through Supplement 10 to the 2010 Edition. Revision 37 of Regulatory Guide 1.84 and Revision  
Edition and Supplements 0 through 7 to the 2013 Edition. The NRC has published RG 1.84, Design, Fabrication, and Materials Code Case Acceptability, ASME Section III, Revision 38, and RG 1.147, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1, Revision 19, concurrently with this guide to identify the Code Cases that the NRC has determined to be acceptable alternatives to applicable parts of BPV Code, Sections III and XI. The NRC staff also reviewed the OM  
18 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 reviewed the OM Code Cases listed in the 2009 Edition through the 2012 Edition.  Revision 2 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
Code Cases listed in the 2015 and the 2017 Editions of the OM Code. The NRC published RG 1.192, Operation and Maintenance Code Case Acceptability, ASME OM Code, Revision 3, 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.


RG 1.193, Rev. 6, Page 4 C. STAFF REGULATORY GUIDANCE
RG 1.193, Page 4  


RG 1.193, Revision 6 supersedes the information in Revision 5.  Licensees should not implement the Code Cases from BPV Code, Sections III and XI, 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 the 2017 Editions of the OM Code, that are listed in this guide without prior NRC approval. The following three tables list the Code Cases that this RG addresses:
C.  STAFF REGULATORY GUIDANCE


a. 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.
Licensees may not implement Code Cases from the Section III and Section XI Codes listed in Supplement 11 to the 2010 Edition through Supplement 10 to the Edition, and the OM Code Cases listed in the 2009 Edition through the 2012 Edition that are listed in this guide without prior NRC approval.


b. 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.
1.


c. Table 3, Unacceptable OM Code Cases, contains OM Code Cases that are unacceptable for use by licensees in their inservice testing programs.
Unacceptable Section III Code Cases  


RG 1.193, Rev. 6, Page 5
The NRC determined that the following Section III Code Cases are unacceptable for use by licensees in their Section III design and construction 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  
CODE CASE  
NUMBER  
NUMBER  
TABLE 1 UNACCEPTABLE SECTION III CODE CASES  
TABLE 1 UNACCEPTABLE SECTION III CODE CASES  
DATE OR
SUPPLEMENT/
EDITION
SUMMARY  
SUMMARY  
N-201-6 Class CS Components in Elevated Temperature Service, Section III,
Division 1
4/10E
Code Case is applicable for high temperature applications beyond that of light water reactors.
N-284-1 Metal Containment Shell Buckling Design Methods, Section III,     
Division 1, Class MC
5/9/03
RG 1.193, Page 5
CODE CASE
NUMBER
TABLE 1 UNACCEPTABLE SECTION III CODE CASES
DATE OR  
DATE OR  
SUPPLEMENT/  
SUPPLEMENT/  
EDITION  
EDITION  
N-201-6 Class CS Components in Elevated Temperature Service, Section III,
 
Division 1 Code Case applies to high-temperature applications beyond that of light-water reactors (LWRs).
SUMMARY
10/18/10
 
N-284-1 Metal Containment Shell Buckling Design Methods, Section III, Division 1, Class MC
(1)  
(1) The following errata, misprints, recommendations, and errors have been identified:  
The following errata, misprints, recommendations, and errors have been identified:  
 
Fig. 1511.1, The curve for L should not exceed 0.8 for any value of (R/t).  
Fig. 1511.1, The curve for L 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:  
 
-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.
See -1713.1.2 for method of calculating M, then see Fig. -1512-1.


-1513, Recommend, Use the value of il given for spherical shells in accordance with 1512.
-1513, Recommend Use the value of il given for spherical shells in accordance with -1512.


-1521, (i) In (a), Axial Compression, G = L should be changed to G = L. (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 the instability equations pertaining to ring-stiffened shells, which are less conservative than those for stringer instability, for either or both ring- 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.
-1521, (i) In (a) Axial Compression, G = L should be changed to G = L. (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 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 Ch = 0.92/(M - 0.636) should be changed to Ch = 0.92/(M - 0.636).  
-1712.1.1, The equation Ch = 0.92/(M - 0.636) should be changed to Ch = 0.92/(M - 0.636).  
-1712.1.1-1, The leftmost curve should be labeled Ch.


-1712.2.2, In (a) Axial Compression, (i) the denominator in the formula for ej should be (m/Lj)2  t. (ii) The expressions for C  
-1712.1.1-1, The leftmost curve should be labeled Ch.
 
-1712.2.2, (a) Axial Compression, (i) In the formula for ej, the denominator should be (m/Lj)2  t. (ii) The expressions for C  
and C should be separated.
and C 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 t 11/4.
-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 t11/4.


should be corrected to t11/4.
should be corrected to t11/4.


5/9/03
RG 1.193, Page 6
5/9/03
 
Table 1. Unacceptable Section III Code Cases


RG 1.193, Rev. 6, Page 6 CODE CASE  
CODE CASE  
NUMBER  
NUMBER  
TABLE 1 UNACCEPTABLE SECTION III CODE CASES  
TABLE 1 UNACCEPTABLE SECTION III CODE CASES  
SUMMARY
DATE OR  
DATE OR  
SUPPLEMENT/  
SUPPLEMENT/  
EDITION  
EDITION  
-1713.1.1, (i) The equation a=el/FS should be changed to a=Lel/FS. (ii) The title of (c) should be changed to Axial Compression Plus In-Plane Shear.
 
SUMMARY
 
N-284-1 (contd)
Metal Containment Shell Buckling Design Methods, Section III,     
Division 1, Class MC
5/9/03
 
-1713.1.1, (i) The equation a=el/FS should be changed to a=Lel/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=ra on the vertical axis should be changed to Ks=ha.
-1713.1-1, In (b), the lower value Ks=ra on the vertical axis should be changed to Ks=ha.


-1713.2.1, (i) The headings for (b) and (c) should include the words In-Plane. (ii) In (b) Axial Compression Plus Shear, should be changed to .
-1713.2.1, (i) The headings for (b) and (c) should include the words In-Plane. (ii) In (b) Axial Compression Plus Shear, should be changed to .  
(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 that apply to the evaluation of 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.
(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.


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.
(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
5/7/99  
5/7/99  
2/25/02 N-510
2/25/02  
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. 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.


The Code Case lacks sufficient detail to ensure that the supplied material is as represented by the Certified Material Test Report.
N-510
N-510-1 Borated Stainless Steel for Class CS Core Support Structures and Class 1 Component Supports, Section III, Division 1
12/9/93  
12/9/93  
8/14/01 N-519 Use of 6061-T6 and 6061-T651 Aluminum for Class 1 Nuclear Components Code Case N-519 only applies to one U.S. Department of Energy aluminum vessel.
8/14/01  
 
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.  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 Annulled 2/3/03


Annulled 2/3/03 N-530
Code Case is applicable to only one DOE aluminum vessel.
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.


N-530
Provisions for Establishing Allowable Axial Compressive Membrane Stresses in the Cylindrical Walls of 0-15 Psi Storage Tanks, Classes 2 and 3
2/3/03  
2/3/03  


RG 1.193, Rev. 6, Page 7 CODE CASE  
There are numerous errors in the equations.  The errors must be corrected before the Code Case can be approved for use.
 
RG 1.193, Page 7  
 
CODE CASE  
NUMBER  
NUMBER  
TABLE 1 UNACCEPTABLE SECTION III CODE CASES  
TABLE 1 UNACCEPTABLE SECTION III CODE CASES  
SUMMARY
DATE OR  
DATE OR  
SUPPLEMENT/  
SUPPLEMENT/  
EDITION  
EDITION  
N-565 Alternative Methods of Nozzle Attachment for Class 1 Vessels, Section III,
Division 1 Code Case N-565 essentially requires a design that uses 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 Code Case N-565 does not demonstrate that 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
SUMMARY
 
N-565 Alternative Methods of Nozzle Attachment for Class 1 Vessels
12/3/99  
 
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.


Requirements for Spent Fuel Storage Canisters, Section III, Division 1 Regulatory approval for the use of multipurpose casks is presently addressed by Spent Fuel Storage and Transportation (SFST) Interim Staff Guidance (ISG) 4, Cask Closure Weld Inspections, Revision 1 (Ref. 9),
N-595 N-595-1 N-595-2 N-595-3 N-595-4
and SFST-ISG-18, The Design and Testing of Lid Welds on Austenitic Stainless Steel Canisters as the Containment Boundary for Spent Fuel Storage, Revision 1 (Ref. 10). The ISGs provide a framework to ensure that the as-designed cask system, 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. 11). Note that Code Case N-717 replaces Code Case N-595-X.


Requirements for Spent Fuel Storage Canisters, Section III, Division 1
2/26/99  
2/26/99  
9/24/99  
9/24/99  
12/8/00  
12/8/00  
4/8/02 Annulled  
04/08/02 Annulled 10/14/11  
10/14/11 N-645 N-645-1 Use of Rupture Disk Devices on Nuclear Fuel Storage Canisters, Class 1, Section III, Division 1 The NRC does not permit the use of rupture disk devices in spent nuclear fuel storage canister designs.
 
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 (Ref. 7), and Interim Staff Guidance No. 18 (ISG-
18), Rev. 1 (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.


N-645 N-645-1 Use of Rupture Disk Devices on Nuclear Fuel Storage Canisters, Class 1, Section III, Division 1
6/14/00  
6/14/00  
2/3/03 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 NRCs 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. Because of the number of issues raised by the NRC staff and because of the concerns expressed in the public comments, the NRC
2/3/03 The NRC does not permit the use of rupture disk devices in spent nuclear fuel storage canister designs.
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. The NRC staff continues to interact with the cognizant ASME committees, and the industry is working to provide additional data and information.


N-659 N-659-1 Use of Ultrasonic Examination in Lieu of Radiography for Weld Examination, Section III, Division 1
9/17/02  
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). Although preliminary results suggest that replacing 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
11/18/03  
6/9/08


RG 1.193, Rev. 6, Page 8 CODE CASE  
The NRC conditionally approved Code Case N-659 in Revision 34 of Regulatory Guide 1.84.  The NRCs 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.
 
N-659-2 Use of Ultrasonic Examination in Lieu of Radiology for Weld Examination, Section III, Divisions 1 and 3
6/09/08
 
RG 1.193, Page 8  
 
CODE CASE  
NUMBER  
NUMBER  
TABLE 1 UNACCEPTABLE SECTION III CODE CASES  
TABLE 1 UNACCEPTABLE SECTION III CODE CASES  
SUMMARY
DATE OR  
DATE OR  
SUPPLEMENT/  
SUPPLEMENT/  
EDITION  
EDITION  
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.


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 a later edition and addenda. The reliability UT performed under the provisions of Section V, has been shown to be inferior to UT
SUMMARY
techniques developed through a program under which the performance characteristics have been shown to be sufficient and reliable.
 
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.
 
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 addend


Furthermore, the qualification specimens do not specify an adequate number of flaws required for the sample set, the required flaw distribution within the specimen, or the required size distribution within the specimen.
====a. 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.


Therefore, performance demonstration requirements, including acceptance criteria for UT equipment, procedures, and personnel used for construction/fabrication activities, must be addressed.
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 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  
Until studies are complete that demonstrate the ability of UT to replace RT  
Line 235: Line 284:


N-670  
N-670  
Use of Ductile Cast Iron Conforming to ASTM A 874/A 874M-98 or JIS  
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.  Therefore, it would not be appropriate to approve a Code Case that is an alternative to the Section III, Division 3, provisions.
G5504-1992 for Transport Containments, Section III, Division 3  
7/01/05
 
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.


7/1/05 N-673 Boron Containing Power Metallurgy Aluminum Alloy for Storage and Transportation of Spent Nuclear Fuel, Section III, Division 1 Code Case N-673 does not address the following:
RG 1.193, Page 9
(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 high-temperature 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 under 10 CFR 50.55a(b)(1)(iii).
5/21/03


RG 1.193, Rev. 6, Page 9 CODE CASE  
CODE CASE  
NUMBER  
NUMBER  
TABLE 1 UNACCEPTABLE SECTION III CODE CASES  
TABLE 1 UNACCEPTABLE SECTION III CODE CASES  
SUMMARY
DATE OR  
DATE OR  
SUPPLEMENT/  
SUPPLEMENT/  
EDITION  
EDITION  
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. Thus, it would not be appropriate to approve an Code Case that is an alternative to the provisions in Section III, Division 3.


11/2/04 N-717 Requirements for Construction of Storage Containments for Spent Nuclear Fuel and High Level Radioactive Waste and Material, Section III, Division  
SUMMARY
3 The NRC has not yet endorsed Section III, Division 3. Therefore, it would not be appropriate to approve an Code Case that is an alternative to the provisions in Section III, Division 3.
 
N-673 Boron Containing Power Metallurgy Aluminum Alloy for Storage and Transportation of Spent Nuclear Fuel, Section III, Division 1
8/7/03
 
The Code Case does not address the following:
(1)  Corrosion properties of this material in spent fuel pool chemistry and/or clean water.


5/4/04 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 Safety-Related 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.
(2)   Impact properties for use as a structural material.


The LRFD method is a probabilistic method developed to provide uniform practice in the design of steel structures for nuclear facilitie
(3)  Uniform distribution of boron carbide in the aluminum matrix.
 
(4)  Mechanical properties for the use of the material in high- temperature conditions.
 
N-693 Alternative Method to the Requirements of NB-3228.6 for Analyzing Piping Subjected to Reversing Dynamic Load, Section III, Division 1
5/21/03
 
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).
 
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
11/02/04
 
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.


====s. The LRFD ====
N-717 Requirements for Construction of Storage Containments for Spent Nuclear Fuel and High Level Radioactive Waste and Material,  Section III,
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.
Division 3
5/04/04


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-to-date. 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.
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.


9/9/08 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 The NRC has not yet endorsed Section III, Division 3. Therefore, it would not be appropriate to approve a Code Case that is an alternative to the provisions in Section III, Division 3.
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.


10/11/05 N-755 N-755-1 N-755-2 Use of Polyethylene (PE) Plastic Pipe, Section III, Division 1 and Division XI
The Code Case is listed in this guide pending the results of the NRC staff review.
The staff has raised issues on materials, fusion qualification requirements, nondestructive examination (NDE), crack growth, and lack of data to support operating experience.


3/22/07
RG 1.193, Page 10
7/15/11
1/13E


RG 1.193, Rev. 6, Page 10
CODE CASE  
CODE CASE  
NUMBER  
NUMBER  
TABLE 1 UNACCEPTABLE SECTION III CODE CASES  
TABLE 1 UNACCEPTABLE SECTION III CODE CASES  
SUMMARY
DATE OR  
DATE OR  
SUPPLEMENT/  
SUPPLEMENT/  
EDITION  
EDITION  
N-761 Fatigue Design Curves for Light Water Reactor (LWR) Environments, Section III, Division 1 Research has shown that the effect of the 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.


*
SUMMARY
The proposed curves in Code Case N-761 for carbon and low-alloy steels (as shown in Figure 2 and Table 1 of the Code Case) and the curves for austenitic stainless steels (as shown in Figure 3 and Table 2 of the Code Case) are not acceptable because sufficient technical basis has not been provided.
 
N-721 Alternative Rules for Linear Piping Supports, Section III, Division 1
9/09/08 Code Case N-721 allows the use of ANSI/AISC N690L-03, Load and Resistance Factor Design (LRFD) Specification for Safety-Related 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.
 
The LRFD method is a probabilistic method developed to provide uniform practice in the design of steel structures for nuclear facilitie


* 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. 12). The factor of 10 on cycles is technically inconsistent with the factor of 12 in NUREG/CR-6909. The proposed curves are nonconservative relative to the estimates based on the 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 a technical and regulatory perspective.
====s. 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 technical basis document for the proposed Code Case does not describe the step-by-step process, from beginning to end, on how final design curves for an LWR environment are obtained. The technical basis document does not provide the expression for the best-fit S-N
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-to-date.  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.
curve of the experimental data and the details of the mean stress correction for each curve and how the proposed design curves were obtained.


* The proposed Code Case contains five environmental fatigue curves for carbon and low-alloy steels and five for stainless steels (i.e., the air curve, 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 in the recent Electric Power Research Institute (EPRI)-sponsored component tests in Germany.
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
10/11/05 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.


* There is no information provided in the technical 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.
N-755 N-755-1 Use of Polyethylene (PE) Plastic Pipe, Section III, Division 1 and XI
3/22/07
6/10E
Issues have been raised concerning joining methods, qualification and testing as related to joining procedure, the effectiveness of non-destructive examination (NDE) to detect fabrication flaws, potential degradation processes and susceptibility, and the ability of NDE to detect service- related degradation. The NRC believes that additional technical studies are required to assess these issues and make a final regulatory decisio


* The technical basis document for the Code Case should address the effect of strain threshold and tensile hold time in fatigue evaluations.
====n.  NRC ====
staff continue to interact with the cognizant ASME committees, and the industry is working to provide additional data and information.


9/20/10
RG 1.193, Page 11
9/20/10


RG 1.193, Rev. 6, Page 11 CODE CASE  
CODE CASE  
NUMBER  
NUMBER  
TABLE 1 UNACCEPTABLE SECTION III CODE CASES  
TABLE 1 UNACCEPTABLE SECTION III CODE CASES  
SUMMARY
DATE OR  
DATE OR  
SUPPLEMENT/  
SUPPLEMENT/  
EDITION  
EDITION  
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. 13), could be used as a good model to develop definition and test methods for slip.


Concrete containments in nuclear power plants are important structures;
SUMMARY
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 and Commentary (Ref. 14). 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 leaktight 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.
 
N-761 Fatigue Design Curves for Light Water Reactor (LWR) Environments, Section III, Division 1
3/10E
Research has shown that the effect of environment on reactor components exposed to reactor water is not bounded by the current air fatigue curves.


Concrete and Commentary, Section 21.1.6.1, classifies mechanical splices as 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 spliced bar, as stated in Section 12.14.3.2 of the Code Case. 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 Case. The specified or actual tensile strength of the steel reinforcing bars is 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 Code Case N-791 is the equivalent criterion for Type 1 mechanical splices of ACI 318, Building Code Requirements for Structural Concrete and Commentary (Ref. 15), which is not an adequate criterion for qualifying mechanical splices for use in concrete containment structures.
Bounding curves and a series of other curves for known strain rates have been developed to account for the reduction of fatigue life.


Therefore, the Code Case should develop a more stringent strength criterion, and the same criterion should also be used for continuing splice performance tests in the field, as stated in Section 5 of the Code Case.
*  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.


9/20/10
*  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 (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.
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.
*  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.


Specifically, the six most significant differences between the Code Case and the latest NRC research are as follows:
*  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.
a.


Carbon and Low-Alloy Steel Fatigue Curve: Code Case, Figure-2100-1, Figure-2100-1M, and Table-2100-1, define the design
RG 1.193, Page 12
9/20/10
11/10E


RG 1.193, Rev. 6, Page 12 CODE CASE  
CODE CASE  
NUMBER  
NUMBER  
TABLE 1 UNACCEPTABLE SECTION III CODE CASES  
TABLE 1 UNACCEPTABLE SECTION III CODE CASES  
SUMMARY
DATE OR  
DATE OR  
SUPPLEMENT/  
SUPPLEMENT/  
EDITION  
EDITION  
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 Section III, Appendix I (2011 Addenda). The Code Case fatigue curve does not match the carbon or low-alloy steel design fatigue air curves from the initial revision of NUREG/CR-6909 (which are the same curves the NRC intends to use in NUREG/CR-6909, Revision 1)
because the Code Case fatigue curve uses 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 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 that was adjusted to account for the difference in the margin term used to develop the ASME and NRC design fatigue curves. This equation is different from the Fen expression recently developed by the NRC, and the equations for the transformed environmental parameters are different; therefore, the Fen equation may yield nonconservative values of Fen for the following reasons:
SUMMARY
(1) The use of average temperature with the Code Case Fen expression may be nonconservative (see item (f)).  
 
(2) The Code Case Fen expression was adjusted to account for the difference in the margin used to develop the design curve (i.e., the factor of 20 versus 12 discussed under item (a)
N-761 (contd)
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 nonconservative 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.
Fatigue Design Curves for Light Water Reactor (LWR) Environments, Section III, Division 1  
3/10E
*  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
4/10E
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.


(3) 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 because the constant is higher for low-alloy steel compared to carbon steel (0.702 versus 0.632).
Therefore, the code case needs to develop a slip criterion for mechanical splices.
(4) The Code Case Fen expression is nonconservative for some environmental conditions compared to the new NRC
expressions (i.e., for T less than 200 °C, strain rate equal to
0.001%/s, and dissolved oxygen values higher than 0.04 ppm).
c.


Stainless Steel Fatigue Curve: Code Case, Figure-2100-2, Figure-2100-2M, and Table-2100-2, define the design fatigue air curve for stainless steels. The Code Case fatigue curve matches the design fatigue air curve that is currently in Section III, Appendix I (2011 Addenda). The Code Case fatigue curve matches the stainless steel design fatigue air curve from the initial revision to NUREG/CR-6909 (which is the same curve the NRC intends to use in NUREG/CR-6909,  
RG 1.193, Page 13


RG 1.193, Rev. 6, Page 13 CODE CASE  
CODE CASE  
NUMBER  
NUMBER  
TABLE 1 UNACCEPTABLE SECTION III CODE CASES  
TABLE 1 UNACCEPTABLE SECTION III CODE CASES  
SUMMARY
DATE OR  
DATE OR  
SUPPLEMENT/  
SUPPLEMENT/  
EDITION  
EDITION  
Revision 1). 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 to NUREG/CR-6909. This equation is different from the Fen expression that the NRC recently developed, and the equations for the transformed environmental parameters are different; therefore, the Fen equation may yield nonconservative values of Fen in cases that use the average temperature (see item (f)).
SUMMARY
e.


Ni-Cr-Fe Steel: The same observations under item (c) applies for Ni-Cr-Fe steels because 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 to 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 nonconservative values of Fen in cases that use the average temperature (see item (f)).
N-791 (contd)
f.
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 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.


-2420 Determination of Transformed Temperature:
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.
(1) -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. The NRC disagrees with this approach because it is not consistent with the Fen methodology and because it can be nonconservative.


i.
N-792 Fatigue Evaluations Including Environmental Effects, Section III,   
Division 1
3/10E
Code Case N-792 provides guidance on the use of Fen factors to address the effect of reactor water environment on cyclic damage in Class 1 components.  Research results detailed in Welding Research Council Bulletin 487 and Argonne National Laboratory (ANL) report NUREG/CR-
6909 show that there is a possibility that reactor water environment may have an adverse effect on the fatigue damage for typical metals used in Class 1 components.  Since Section III does not provide specific guidance in the area of environmental fatigue effects, this Code Case has been developed to provide a Code approved method.  The Code Case uses the methodology and Fen equations suggested in NUREG/CR-6909.  One major change in the Code Case compared to NUREG/CR-6909 is the deletion of the strain threshold.


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.
However, based on industry comments that the Fen expressions give Fen values greater than 1.0 for situations when environmental effects have no impact, there are ongoing activities at NRC to modify Fen expressions.  The NRCs Office of Nuclear Regulatory Research (RES) with the assistance of experts at Argonne National Laboratory (ANL) is pursuing this effort.


ii.
N-793 Extruded Steel Sleeves With Parallel Threaded Ends, Section III, Division
2
4/10E
See comments for N-791.


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.
RG 1.193, Page 14


(2) -2422 defines the transformed temperature for carbon and low-alloy steels for temperatures up to 350 °C (660 °F). The NRCs updated research only includes data up to 325 °C
CODE CASE  
(615 °F); therefore, the updated Fen expression for carbon and low-alloy steels is only applicable for temperatures up to
325 °C.
 
(3) -2423 defines the transformed temperature for wrought and cast austenitic stainless steels for temperatures above 325 °C
(615 °F) as constant (T* = 1). The NRCs 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
 
RG 1.193, Rev. 6, Page 14 CODE CASE  
NUMBER  
NUMBER  
TABLE 1 UNACCEPTABLE SECTION III CODE CASES  
TABLE 1 UNACCEPTABLE SECTION III CODE CASES  
SUMMARY
DATE OR  
DATE OR  
SUPPLEMENT/  
SUPPLEMENT/  
EDITION  
EDITION  
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). The NRCs updated research only includes data up to 325 °C (615 °F), and the updated Fen expression for Ni-Cr-Fe steels does not plateau at temperatures above 325 °C. Therefore, the Code Case may provide nonconservative estimates of Fen for temperatures above 325 °C.
SUMMARY


The NRC recommends that Code Case N-792-1 be revised to reflect NUREG/CR-6909 Rev. 1 after it is published.
N-794 Swaged Splice With Threaded Ends, Section III


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
See comments for N-791.
sponsored research that is ongoing.


N-793 Extruded Steel Sleeves with Parallel Threaded Ends, Section III, Division 2 See comments for Code Case N-791.
N-796 Alternative Preheat Temperature for Austenitic Welds in P-No. 1 Material Without PWHT, Section III, Division 1


9/20/10
See comments for N-791.
N-794 Swaged Splice with Threaded Ends, Section III
See comments for Code Case N-791.


9/20/10
N-804 Alternative Preheat Temperature for Austenitic Welds in P-No. 1 Material Without PWHT, Section III, Division 1 
N-796 Threaded Sleeves with Parallel Threads Cut on Upsized Bar Ends, Section III, Division 2 See comments for Code Case N-791.
7/10E
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 values.  Hydrogen cracking of the welds could result in the absence of proper preheat.


10/18/10
N-807 Use of Grades 75 and 80 Reinforcement in Concrete Containments, Section III, Division 2
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 are insufficient to support a reduction in the Code-required preheat of 200 °F. Data for the welds in the production valve bodies tested indicate the presence of martensite, which results in unacceptably high hardness values. Hydrogen cracking of the welds could result in the absence of proper preheat.
7/10E
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.


10/14/11 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 will thereby reduce the overall ductility of the containment, which is undesirable.
N-812 Alternate Creep-Fatigue Damage Envelope for 9Cr-1Mo-V Steel, Section III, Division 1
7/10E
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.


4/20/11 N-812 N-812-1 Alternate Creep-Fatigue Damage Envelope for 9Cr-1Mo-V Steel, Section III, Division 5 Code Case N-812 utilizes Section III, Division, Subsection NH, Class 1 Components in Elevated Temperature Service. The NRC has not approved Subsection NH for use.
N-818 Alternative Requirements for Preservice Volumetric and Surface Examination, Section III, Division 1  
8/10E


8/5/11
RG 1.193, Page 15
1/13E
N-818 N-818-1  


Use of NDE and Fracture Mechanics for Acceptance of Full Penetration Butt Welds in Lieu of Weld Repair, Class 1 and 2, 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
CODE CASE  
12/6/11
7/13E
 
RG 1.193, Rev. 6, Page 15 CODE CASE  
NUMBER  
NUMBER  
TABLE 1 UNACCEPTABLE SECTION III CODE CASES  
TABLE 1 UNACCEPTABLE SECTION III CODE CASES  
SUMMARY
DATE OR  
DATE OR  
SUPPLEMENT/  
SUPPLEMENT/  
EDITION  
EDITION  
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 the following concerns regarding the provisions of UT
SUMMARY
and other issues in this code case.
 
N-818 (contd)
The NRC has been conducting research at Pacific Northwest National Laboratory on the examination of austenitic and ferritic welds.  The work has shown that performing a full volume examination for fabrication flaws is significantly different from an inservice examination.  For example, examination from two directions is necessary to detect certain circumferentially oriented fabrication flaws such as lack of fusion.  The work has also shown that the second leg of V-path can be applied to ferritic materials on a limited basis but will be difficult to apply to austenitic materials and dissimilar metal welds.  Another finding is that surface conditions are critical with respect to detecting and characterizing fabrication flaws.  Additionally, the PNNL research suggests that the ability to consistently and accurately characterize fabrication flaws by type (i.e., planar or volumetric) is difficult.  This capability is essential if acceptance criteria based on flaw type is to be applied.  In summary, the NRC believes that an analytical approach for the acceptance of certain fabrication flaws could be acceptable if appropriately justified and the scope limited to ferritic materials. The NRC believes that significant research will be required to demonstrate that full-volume examination for fabrication flaws is acceptable for austenitic and dissimilar metal welds.
 
N-820
Twisting of Horizontal Prestressing Tendons, Section III, Division 2
8/10E
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.


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 in Code Case N-831) for ferritic materials. The NRC staff has reviewed and approved relief requests for UT
N-828 Alternative Nonmetallic Material Manufacturers and Constituent Suppliers Quality System Program Requirements, Section III, NCA-3900,  
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.
2010 Edition, and Earlier Editions and Addenda, Section III, Divisions 1 and 2
10/10E
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.


2. Single side access in not acceptable for fabrication examinations because some flaws are only detectable from one direction.
DG-1298, Page 16
2.


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.
Unacceptable Section XI Code Cases


4. Surface preparation needs to be addressed. Welds must be conditioned without any gap more than 1/32-inch between transducer and weld.
The NRC determined that the following Section XI Code Cases are unacceptable for use by licensees in their Section XI inservice inspection programs.


5. Paragraph (g) of the code case seems to be the discussion of a calibration block, not a qualification block.
Table 2. Unacceptable Section XI Code Cases


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 service-induced surface connected flaws. The NRC staff does not find using this RMS error is appropriate for measurements of fabrication defects.
CODE CASE
NUMBER
TABLE 2 UNACCEPTABLE SECTION XI CODE CASES
DATE OR
SUPPLEMENT
/EDITION


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.
SUMMARY


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
N-465 N-465-1 Alternative Rules for Pump Testing, Section XI, Division 1
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 NB-3000 and NC-3000 are based on a component without flaws.
11/30/88 Annulled
2/14/03


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.
The draft standard referenced in the Code Case is outdated.


N-820
The requirements contained in the OM Code, Code for Operation and Maintenance of Nuclear Power Plants, should be used. Note that Revision 12 of Regulatory Guide 1.147 approved N-465 for use.
Twisting of Horizontal Prestressing Tendons, Section III, Division 2 New reactor designs will use 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
The disapproval of N-465 for use applies only to new users.


RG 1.193, Rev. 6, Page 16 CODE CASE
N-473 N-473-1 Alternative Rules for Valve Testing, Section XI, Division 1  
NUMBER
3/8/89 Annulled
TABLE 1 UNACCEPTABLE SECTION III CODE CASES
2/14/03
SUMMARY
DATE OR
SUPPLEMENT/  
EDITION
N-828 Alternative Nonmetallic Material Manufacturers and Constituent Suppliers Quality System Program Requirements, Section III, NCA-3900, 2010
Edition, and Earlier Editions and Addenda, Section III, Divisions 1 and 2 Code Case N-828 was developed to support new nuclear plant construction.


The NRC plans to address this Code Case in RG 1.136, Design Limits, Loading Combinations, Materials, Construction, and Testing of Concrete Containments.
The draft standard referenced in the Code Case is outdated.


4/27/12 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; therefore, it does not apply to U.S. nuclear facilities regulated by the NRC.
The requirements contained in the OM Code, Code for Operation and Maintenance of Nuclear Power Plants, should be used. Note that Revision 12 of Regulatory Guide 1.147 approved N-473 for use.


3/13E
The disapproval of N-473 for use applies only to new users.
N-846 Certificate Holder Ability to Supply Polyethylene Material, Section III,
Division 1 Code Case N-846 is not consistent with NRC position documented in NRC
Information Notice IN 86-21, "Recognition of ASME Accreditation program for N Stamp Holders," and creates issues for verifying the effective implementation of a suppliers QA program. IN 86-21, Supplement 2, stated that the NRC's recognition of the ASME Accreditation Program applied only to the programmatic aspects of the QA programs and that holders of operating licenses or construction permits, and their subcontractors, are still responsible for ensuring that the suppliers are effectively implementing their approved QA programs.


7/13E
N-480
Examination Requirements for Pipe Wall Thinning Due to Single Phase Erosion and Corrosion, Section XI, Division 1 Annulled 9/18/01


RG 1.193, Rev. 6, Page 17
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).
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.
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


Table 2. Unacceptable Section XI Code Cases
Code Case N-498-4 is conditionally approved in Revision 13 of Regulatory Guide 1.147. Those licensees choosing to implement this Code Case are to implement Revision 4, which is listed in Revision
15 of Regulatory Guide 1.147.


CODE CASE  
DG-1298, Page 17 CODE CASE  
NUMBER  
NUMBER  
TABLE 2 UNACCEPTABLE SECTION XI CODE CASES  
TABLE 2 UNACCEPTABLE SECTION XI CODE CASES  
SUMMARY
DATE OR  
DATE OR  
SUPPLEMENT/  
SUPPLEMENT
EDITION  
/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 should be used.


11/30/88 Annulled
SUMMARY
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 should be used.


3/8/89 Annulled
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,  
2/14/03 N-480
Division 1  
Examination Requirements for Pipe Wall Thinning Due to Single Phase Erosion and Corrosion, Section XI, Division 1 The Code Case has been superseded by Code Case N-597, Requirements for Analytical Evaluation of Pipe Wall Thinning, implemented in conjunction with EPRI Nuclear Safety Analysis Center 202L, Recommendations for an Effective Flow-Accelerated Corrosion Program (Ref. 16).
7/23/02
Annulled 9/18/01 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 NRC identified the following concerns during its review of the Code Case:
a.


The Code Case references new paragraph IWA-6350, which has not yet been incorporated into the Code.
The following concerns were identified during review of the Code Case:
(1) The Code Case references new paragraph IWA-6350, which has not yet been incorporated into the ASME Code.


b.
(2) NRC staff had difficulty reconciling Footnote 1 and Table 4 regarding the applicable edition and addenda.


The NRC staff had difficulty reconciling Footnote 1 and Table 4 in regard to 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.


c.
N-542 Alternative Requirements for Nozzle Inside Radius Section Length Sizing Performance Demonstration, Section XI, Division 1 Annulled 3/28/01


Submission of Form OAR-1, Owners Activity Report, is at the end of each inspection period rather than 90 days following the outage.
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.


7/23/02 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 Methods Qualification 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.
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 Annulled 5/20/01  


RG 1.193, Rev. 6, Page 18 CODE CASE
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)  
NUMBER
required the examination of CRD bolting material whenever the CRD  
TABLE 2 UNACCEPTABLE SECTION XI CODE CASES
housing is disassembled and the bolting material is to be reused.
SUMMARY
DATE OR
SUPPLEMENT/
EDITION
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) requires the examination of CRD bolting material whenever the CRD housing is disassembled and the bolting material is to be reuse


====d. Examination of CRD ====
Examination of CRD bolting material is required to verify that service- related 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.
bolting material is required to verify that service-related 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  
N-560-1 N-560-2 Alternative Examination Requirements for Class 1, Category B-J Piping Welds, Section XI, Division 1  
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),  
8/9/96
2/26/99
2/14/03
 
(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.
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.
(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
DG-1298, Page 18 CODE CASE
2/26/99
NUMBER
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.
TABLE 2 UNACCEPTABLE SECTION XI CODE CASES
DATE OR
SUPPLEMENT
/EDITION


Reinspection requirements are not provided to verify structural integrity because the root cause may not be mitigated.
SUMMARY


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
12/31/96  
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.
3/28/01  


Reinspection requirements are not provided to verify structural integrity because the root cause may not be mitigated.
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.


N-562 N-562-1 Alternative Requirements for Wall Thickness Restoration of Class 3 Moderate Energy Carbon Steel Piping, Section XI, Division 1
12/31/96  
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 Section XI,
3/28/01  
Appendix VIII, Performance Demonstration for Ultrasonic Examination Systems, the NRC staff noted that proficiency decreases over time. The data do not support recertification examinations at a frequency of every 5 years.
 
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.
 
N-574 NDE Personnel Recertification Frequency, Section XI, Division 1 Annulled 7/14/06
 
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.
 
N-575 Alternative Examination Requirements for Full Penetration Nozzle-to-Vessel Welds in Reactor Vessels with Set-On Type Nozzles, Section XI, Division 1
2/14/03
 
The supporting basis for the Code Case applies to the specific configuration of one plant and is not applicable on a generic basis.


Annulled 7/14/06 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.
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.


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  
2/14/03  


RG 1.193, Rev. 6, Page 19 CODE CASE  
(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.
 
DG-1298, Page 19 CODE CASE  
NUMBER  
NUMBER  
TABLE 2 UNACCEPTABLE SECTION XI CODE CASES  
TABLE 2 UNACCEPTABLE SECTION XI CODE CASES  
SUMMARY
DATE OR  
DATE OR  
SUPPLEMENT/  
SUPPLEMENT
EDITION  
/EDITION  
N-577 N-577-1 Risk-Informed Requirements for Class 1, 2, and 3 Piping, Method A,
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 that dominant risk contributors will not be created.
SUMMARY


9/2/97
N-578 N-578-1 Risk-Informed Requirements for Class 1, 2, and 3 Piping, Method B,  
2/14/03 N-578 N-578-1 Risk-Informed Requirements for Class 1, 2, and 3 Piping, Method B,  
Section XI, Division 1  
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 that dominant risk contributors will not be created.
9/2/97  
9/2/97  
2/14/03 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.
2/14/03  


Annulled
(1) The Code Case does not address inspection strategy for existing augmented and other inspection programs such as IGSCC, FAC,  
2/14/03 N-589 N-589-1 Class 3 Nonmetallic Cured-in-Place Piping, Section XI, Division 1
MIC, and pitting.
(1) The installation process provides insufficient controls on wall thickness measurements.


(2) There are no qualification requirements for installers and installation procedures such as those for welders and welding procedures.
(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.


(3) Fracture toughness properties of the fiberglass are such that the cured-in-place piping (CIPP) could crack during a seismic event.
N-587 Alternative NDE Requirements for Repair/Replacement Activities, Section XI, Division 1 Annulled
 
2/14/03
(4)
Equations 4 and 5 in the Code Case contain an i term (a stress intensification factor) that is derived from fatigue considerations.


However, stress intensification factors have not been developed for fiberglass materials.
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.


N-589 N-589-1 Class 3 Nonmetallic Cured-in-Place Piping, Section XI, Division 1
4/19/02  
4/19/02  
7/23/02 N-590
7/23/02  
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. Therefore, the Code Case is no longer needed and was annulled by the ASME.


Annulled 4/8/02
(1)
The installation process provides insufficient controls on wall thickness measurement.


RG 1.193, Rev. 6, Page 20
(2)
CODE CASE
There are no qualification requirements for installers and installation procedures such as those for welders and welding procedures.
NUMBER
TABLE 2 UNACCEPTABLE SECTION XI CODE CASES
SUMMARY
DATE OR
SUPPLEMENT/
EDITION
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-2500-7(a) through IWB-2500-7(d). The action is applicable from the 1974 Edition through the  
(3)  
2004 Edition with the 2005 Addenda. A similar action was taken in regard to Code Case N-619. The NRC did not take exception to Code Case N-619 because 10 CFR 50.55a(b)(2)(xxi)(A) requires 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 before the 1998 Edition, which do not have the appropriate figures.
Fracture toughness properties of the fiberglass are such that the cured-in-place piping (CIPP) could crack during a seismic event.


10/8/04 N-613 Ultrasonic Examination of Full Penetration Nozzles in Vessels, Examination Category B-D, Item Nos. B3.10 and B3.90, Reactor Vessel-To-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
(4)  
10 CFR 50.55a(b)(2)(xv)(K)(2)(i). A revision to the Code Case has been developed to address the concerns.
Equations 4 and 5 in the Code Case contain an i term [a stress intensification factor] that is derived from fatigue considerations.


7/30/98 N-615 Ultrasonic Examination as a Surface Examination Method for Category B-F
Stress intensification factors, however, have not been developed for fiberglass materials.
and B-J Piping Welds, Section XI, Division 1 The Code Case requires the ultrasonic technique used to 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, 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). However, the NRC staff determined that the Code Case did not apply to the review and approval process for transportation packages. The regulations in 10 CFR Part 71, Packaging and Transportation of Radioactive Material, address the use of RPVs as a transportation package (Ref. 17).
N-590
6/17/03 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. Separate Code Cases are addressing issues associated with supplements to Section XI, Appendix VIII, individually.
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 Annulled 4/8/02


Annulled on
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.
1/12/05


RG 1.193, Rev. 6, Page 21 CODE CASE  
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DATE OR  
DATE OR  
SUPPLEMENT/  
SUPPLEMENT
EDITION  
/EDITION  
N-653 Qualification Requirements for Full Structural Overlaid Wrought Austenitic Piping Welds, Section XI, Division 1 a.


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.
SUMMARY


Personnel qualification is necessary to validate the effectiveness of the procedure qualification.
N-591 Alternative to the Requirements of Subsection IWL, Requirements for Class CC Concrete Components of Light-Water Cooled Plants, Section XI, Division 1 Annulled 4/8/02


b.
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.


The minimum grading unit is 1.0 inch in the circumferential direction.
N-593-1 Examination Requirements for Steam Generator Nozzle-to-Vessel Welds, Section XI, Division 1
10/08/04 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-2500-7(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.


However, the acceptance tolerance is a 0.75-inch RMS error. Therefore, the length-sizing acceptance criteria do not adequately prevent the use of testmanship rather than skill to pass length-sizing tests.
N-613 Ultrasonic Examination of Full Penetration Nozzles in Vessels, Examination Category B-D, Item Nos. B3.10 and B3.90, Reactor Vessel- To-Nozzle Welds, Fig. IWB-2500-7(a), (b), and (c), Section XI, Division 1
7/30/98


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
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.
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 staffs 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 the achievement of a coverage requirement is impractical and when ASME
N-615 Ultrasonic Examination as a Surface Examination Method for Category B-F and B-J Piping Welds, Section XI, Division 1  
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 the NRCs responsibility for determining whether examinations are impractical and eliminates the NRCs ability to take exception to a licensees proposed action and to impose additional measures, where warranted, in accordance with 10 CFR 50.55a(g)(6)(i).
7/28/01
1/5/06 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, or N-659-2. Refer to the discussion on Code Case N-659-2 in Table 1 of this guide for more information.


11/10/08 N-716 Alternative Piping Classification and Examination Requirements, Section XI,
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.
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, Unit 1 (September 21, 2007, ML072430005); Donald C. Cook Nuclear Plant
4/10/06


RG 1.193, Rev. 6, Page 22 CODE CASE  
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NUMBER  
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/EDITION
SUMMARY  
SUMMARY  
DATE OR
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EDITION
(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 it has completed its review of the topical report.


N-722-2 Visual Examinations for PWR Pressure Retaining Welds in Class 1 Components Fabricated With Alloy 600/82/182 Materials, Section XI,  
N-618 Use of a Reactor Pressure Vessel as a Transportation Containment System, Section XI, Division 1  
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 under 10 CFR 50.55a and not through RG 1.147. The NRC has dispositioned Code Case N-722-2 as unacceptable.
6/17/03
 
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).
 
N-622 Ultrasonic Examination of RPV and Piping, Bolts, and Studs, Section XI, Division 1 Annulled on
1/12/05
 
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.


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 Welds, Section XI, Division 1 Code Case N-729 has been superseded by Revisions 1, 2, and 3 to the Code Case. Code Case N-729-1 is conditionally approved directly under 10 CFR
N-653 Qualification Requirements for Full Structural Overlaid Wrought Austenitic Piping Welds, Section XI, Division 1  
50.55a and not through RG 1.147. Code Case N-729-4 is addressed directly in 10 CFR 50.55a.
9/7/01


4/4/12
(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.
11/10E
N-740
N-740-1 N-740-2


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
(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.
or 1 in RG 1.147.


The ASME has approved and published Code Case N-740-2. Although Revision 2 addresses some of the NRC staffs concerns, significant issues remain. For example, the definition of nominal weld and base material appear to be inconsistent with the provisions of Section III. In addition, 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-654 Acceptance Criteria for Flaws in Ferritic Steel Components 4 in. and Greater in Thickness, Section XI, Division 1
4/17/02


10/12/06
Licensees intending to apply the rules of this Code Case must obtain NRC
12/25/09
approval of the specific application in accordance with 10 CFR 50.55a(z).
11/10/08


RG 1.193, Rev. 6, Page 23 CODE CASE  
DG-1298, Page 22 CODE CASE  
NUMBER  
NUMBER  
TABLE 2 UNACCEPTABLE SECTION XI CODE CASES  
TABLE 2 UNACCEPTABLE SECTION XI CODE CASES  
DATE OR
SUPPLEMENT
/EDITION
SUMMARY  
SUMMARY  
DATE OR
SUPPLEMENT/
EDITION
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 a.


Paragraph 1.(c)(1) of Code Case N-766 would potentially allow a 75%
N-691 Application of Risk-Informed Insights to Increase the Inspection Interval for Pressurized Water Reactor Vessels, Section XI, Division 1  
through-wall flaw to remain in service in the original Alloy 82/182 dissimilar metal weld, in accordance with Section XI, IWB-3600. The NRC staff finds it is unacceptable to allow such a large flaw to remain in service in Class 1 piping.
11/18/03


b.
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.


In 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 NDE techniques applied.
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
1/05/06


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%
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 NRCs responsibility for determining whether examinations are impractical, and eliminates the NRCs ability to take exception to a licensees proposed action and impose additional measures where warranted in accordance with 10 CFR 50.55a(g)(6)(i).  
through-wall flaw. A 360-degree, 10% through-wall flaw should be analyzed to determine the fatigue and stress-corrosion cracking degradation mechanisms.


c.
N-713 Ultrasonic Examination in Lieu of Radiography, Section XI, Division 1
11/10/08 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.


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 Section XI
N-716 Alternative Piping Classification and Examination Requirements, Section XI, Division 1  
12/20/10
4/10/06 The NRC has approved risk-informed inservice inspection (RI-ISI)  
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 Code Case N-770-2 examinations to be performed as an augmented inspection program under 10 CFR 50.55a(g)(6)(ii)(F). The latest version of Code Case N-770-2 approved by the NRC is incorporated by reference in 10 CFR 50.55a. Once the review of the topical report on the technical basis for peening is complete, the staff expects to review the latest Code-approved version of Code Case N-770 for incorporation directly in
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.
10 CFR 50.55a under 10 CFR 50.55a(g)(6)(ii)(F).  
1/13E
5/13E
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 the 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 NDE experts and laboratory testing. Although 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 to demonstrate reliability before the alternative can be approved in RG 1.147.


4/9/10
The purpose of the topical report, in part, is to provide guidance on 


RG 1.193, Rev. 6, Page 24 CODE CASE  
DG-1298, Page 23 CODE CASE  
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DATE OR
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/EDITION
SUMMARY  
SUMMARY  
DATE OR
SUPPLEMENT/
EDITION
N-784 Experience Credit for Ultrasonic Examiner Certification Code Case N-784 reduces the requirements for training and experience in regard to 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 that would be encountered. The impact of reduced training and experience has not been evaluated.


4/9/10
N-716 (contd)
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 NRCs concerns:
Alternative Piping Classification and Examination Requirements, Section XI, Division 1
(1) The rules applicable to determining corrosion rates that lead to the definition of the evaluation period and reexamination schedules are currently under development. Accordingly, the Code Case does not define the method for determining the wall loss rates, the time period for the length of the evaluation, and the reexamination period/frequency.
4/10/06 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.
 
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
7/10E
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.


(2) The Section XI, appendices used to calculate some of the important values are nonmandatory.
N-729-3 Alternative Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Nozzles, Section XI, Division 1
10/10E
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.


Licensees intending to use Code Case N-806 must submit a plant-specific request to the NRC staff for review and approval before its implementation.
N-740
N-740-1 N-740-2


6/22/12 N-813
Dissimilar Metal Weld Overlay for Repair of Class 1, 2, and 3 Items, Section XI, Division 1
10/12/06
12/25/09
11/10/08


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 Section XI, Paragraph IWB-3112, which 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. Under Section XI, Paragraph IWB-3112, any preservice flaw that exceeds the acceptance standards of Table IWB-3410-1 must be removed.
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.


Although 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 in regard to the proposed alternative in Code Case N-813:
Code Case N-740-2 has been approved and published by the ASME.
(1) The requirements of Section XI, 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 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 Section XI, Paragraphs IWB-3132.3, through IWB-3643, allows a flaw up to 75% through-wall to remain in service.
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.


10/24/11
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.


RG 1.193, Rev. 6, Page 25 CODE CASE  
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SUMMARY
DATE OR  
DATE OR  
SUPPLEMENT/  
SUPPLEMENT
EDITION  
/EDITION  
Larger flaws could grow to an unacceptable size between inspections, thus 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 it does for Class 1 piping.
SUMMARY


N-826 Ultrasonic Examination of Full Penetration Vessel Weld Joints in Fig. IWB-
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
2500-1 Through Fig. IWB-2500-6 Reduction of the inspection volume from 1/2 t to 1/2 inch conflicts with
4/10E
10 CFR 50.61a, Alternate Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events. Licensees implementing 10
(1) Paragraph 1.(c)(1) of Code Case N-766 would potentially allow a
CFR 50.61a must first examine the volume described in Section XI,  
75-percent through wall flaw to remain in service in the original Alloy 82/182 dissimilar metal weld, in accordance with IWB-3600.
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 that the flaw density requirements of 10 CFR 50.61a are met.


Although, under Code Case N-826, a licensee would have examined the full
The NRC staff finds it is unacceptable to allow such a large flaw to remain in service in Class 1 piping.
1/2-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 because of concerns about detection and sizing accuracy for smaller flaws using the current UT technolog


====y. Current UT ====
(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
technology cannot reliably detect and accurately size smaller flaws, which affects the validity of the comparison with the flaw density requirement of
360-degree, 10% through-wall flaw should be analyzed to determine the fatigue and stress corrosion cracking degradation mechanisms.
10 CFR 50.61a. In addition, recent experiences at operating plants involving missed defects during examinations that used qualified methods and were conducted in compliance with Section XI, Appendix VIII, have raised concerns about the reliability of ultrasonic examinations. Finally, the reduction from 1/2 t to 1/2 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 an examination of the full 1/2-t volume.


7/16/12 N-840
(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
Cladding Repair by Underwater Electrochemical Deposition in Class 1 and
and/or Appendix C to the ASME Code, Section XI.
2 Applications, Section XI, Division 1 Code Case N-840 was developed specifically to address erosion/corrosion concerns in a Korean nuclear facility where cladding damage in the RPV has exposed low-alloy steels. If this were to occur in a U.S. nuclear facility the NRC staff would want to review the particular circumstances on a case-by- case basis. Any licensee that wants to use Code Case N-840 should submit it to the NRC for review and approval in accordance with 10 CFR 50.55a(z).
4/13E


RG 1.193, Rev. 6, Page 26
N-780
3. Unacceptable OM Code Cases
Alternative Requirements for Upgrade, Substitution, or Reconfiguration of Examination Equipment When Using Appendix VIII Qualified Ultrasonic Examination Systems, Section XI, Division 1
 
1/10E
The NRC determined that the following OM Code Cases were unacceptable for use by licensees in their inservice testing programs. ASME issues OM Code Cases annually with the 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.
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.
 
Table 3. Unacceptable OM Code Cases


CODE CASE  
DG-1298, Page 25 CODE CASE  
NUMBER  
NUMBER  
TABLE 2 UNACCEPTABLE SECTION XI CODE CASES
DATE OR
SUPPLEMENT
/EDITION


TABLE 3 UNACCEPTABLE OM CODE CASES
SUMMARY


SUMMARY OF BASIS FOR EXCLUSION
N-784 Experience Credit for Ultrasonic Examiner Certification
1/10E
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.


EDITION/
N-806 Evaluation of Metal Loss in Class 2 and 3 Metallic Piping Buried in a Back-Filled Trench
ADDENDA
10/10E
OMN-10  
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 NRCs concerns:
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.
(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.


Note: The 2006 Addenda does not include Pages C-31 through C-34 of Code Case OMN-10.
(2) The ASME Section XI appendices used to calculate some of the important values are nonmandatory.


2000 Addenda Reaffirmed
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.
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 OMN-15 Requirements for Extending the Snubber Operational Readiness Testing Interval at LWR Power Plants The following list summarizes the issues that the NRC has 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.
N-813 Alternative Requirements for Preservice Volumetric and Surface Examination, Section XI, Division 1
8/10E
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 IWB-
3410-1 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:
(1) The requirements of paragraph IWB-3112 were developed to ensure that defective welds were not placed in servic


(3) The Code Case does not address the assignment of unacceptable snubbers in the failure mode group.
====e. A preservice flaw ====


(4) The Code Case does not address the treatment of isolated snubber failures.
DG-1298, Page 26 CODE CASE
NUMBER
TABLE 2 UNACCEPTABLE SECTION XI CODE CASES
DATE OR
SUPPLEMENT
/EDITION


(5) The Code Case does not address how unacceptable snubbers are accounted for during the extended test interval. For example, unacceptable snubbers could be identified during maintenance, service life monitoring, and visual examination activities conducted during the extended test interval.
SUMMARY


Note: Code Case OMN-15, Revision 2 (2017 Edition), is approved for use in RG
N-813 (contd)  
1.192, Revision 3.
Alternative Requirements for Preservice Volumetric and Surface Examination, Section XI, Division 1
8/10E
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.


2004 Edition Revised 2006 Addenda Reaffirmed
(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.
2009 Edition Reaffirmed
2012 Edition


RG 1.193, Rev. 6, Page 27
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.


==D. IMPLEMENTATION==
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.
The purpose of this section is to provide information on how applicants and licensees may use this guide and information regarding the NRCs plans for using this RG. This RG 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 RG. The request should address the NRCs concerns about the Code Case at issue.


RG 1.193, Rev. 6, Page 28
N-826 Ultrasonic Examination of Full Penetration Vessel Weld Joints in Fig.


REFERENCES1
IWB-2500-1 Through Fig. IWB-2500-6
10/10E
Reduction of the inspection volume from 1/2 t to 1/2 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 1/2 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 1/2 t to 1/2 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 1/2 t volume.


1.
DG-1298, Page 27


Code of Federal Regulations (CFR), Domestic Licensing of Production and Utilization Facilities, Part 50, Chapter 1, Title 10, Energy.
3.


2.
Unacceptable OM Code Cases


American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, Rules for Construction of Nuclear Power Plant Components, New York, NY.2
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.


3.
Table 3. Unacceptable OM Code Cases


ASME Code for the Operation and Maintenance of Nuclear Power Plants, New York, NY.2
CODE CASE
NUMBER


4.
TABLE 3 UNACCEPTABLE OM CODE CASES


ASME Boiler and Pressure Vessel Code, Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, New York, NY.2
EDITION/
ADDENDA


5.
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
2000 Addenda Reaffirmed 2001 Edition Reaffirmed 2003 Addenda Reaffirmed 2004 Edition Reaffirmed 2006 Addenda (see Note)
Reaffirmed 2009 Edition 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.


CFR, Licenses, Certifications, and Approvals for Nuclear Power Plants, Part 52, Chapter 1, Title 10, Energy.
Note: Pages C-31 through C-34 were not included in the 2006 Addenda.


6.
OMN-15 Requirements for Extending the Snubber Operational Readiness Testing Interval at LWR Power Plants
2004 Edition Revised 2006 Addenda Reaffirmed 2009 Edition Reaffirmed 2012 Edition


U.S. Nuclear Regulatory Commission (NRC), Regulatory Guide (RG) 1.84, Design, Fabrication, and Materials Code Case Acceptability, ASME Section III, Washington, DC.
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.


7.
(2)  The Code Case does not address snubber service life monitoring requirements when using the 1995 Edition of the OM Code.


NRC, RG 1.147, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1, Washington, DC.
(3)  The Code Case does not address the assignment of unacceptable snubbers in the Failure Mode Group.


8.
(4)  The Code Case does not address treatment of isolated snubber failures.


NRC, RG 1.192, Operation and Maintenance Code Case Acceptability, ASME OM Code, Washington, DC.
(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.


9.
DG-1298, Page 28


NRC, Spent Fuel Storage and Transportation (SFST) Interim Staff Guidance (ISG) SFST-ISG-4, Cask Closure Weld Inspections, Revision 1, Washington, DC (ADAMS Accession No. ML051520313).  
==D. IMPLEMENTATION==
The purpose of this section is to provide information to applicants and licensees regarding the NRC  
staffs 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 must address the NRCs concerns about the Code Case at issue.


10.
DG-1298, Page 29 REFERENCES1


NRC, SFST-ISG-18, The Design and Testing of Lid Welds on Austenitic Stainless Steel Canisters as Containment Boundary for Spent Fuel Storage, Revision 1, Washington, DC  
1. Code of Federal Regulations, Title 10, Energy, Part 50, Domestic Licensing of Production and Utilization Facilities (10 CFR Part 50), U.S. Nuclear Regulatory Commission, Washington, DC.
(ADAMS Accession No. ML031250620).  


11.
2. ASME Boiler and Pressure Vessel Code, Section III, Rules for Construction of Nuclear Power Plant Components, Division I; and Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, American Society of Mechanical Engineers, New York, NY.2


CFR, Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater Than Class C Waste, Part 72, Chapter 1, Title 10, Energy.
3. ASME Code for Operation and Maintenance of Nuclear Power Plants, American Society of Mechanical Engineers, New York, NY.2


12.
4. Regulatory Guide 1.84, Design, Fabrication, and Materials Code Case Acceptability, ASME
Section III, U.S. Nuclear Regulatory Commission, Washington, DC.


NRC, NUREG/CR-6909, Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials, Revision 1, Washington, DC, May 2018.
5. Regulatory Guide 1.147, Inservice Inspection Code Case Acceptability, ASME Section XI,  
Division 1, U.S. Nuclear Regulatory Commission, Washington, DC.


1 Publicly available NRC published documents are available electronically through the NRC Library on the NRCs public Web site at http://www.nrc.gov/reading-rm/doc-collections/ and through the NRCs 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 NRCs 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)
6. Regulatory Guide 1.192, Operation and Maintenance Code Case Acceptability, ASME OM Code, U.S. Nuclear Regulatory Commission, Washington, DC.
415-3548; or e-mail pdr.resource@nrc.gov.


2 Copies of American Society of Mechanical Engineers (ASME) standards may be purchased from ASME at American Society of Mechanical Engineers, Two Park Avenue, New York, NY 10016-5990; telephone (800) 843-2763.
7. NRC Spent Fuel Project Office Interim Staff Guidance No. 4 (ISG-4), Rev. 1, Cask Closure Weld Inspections, U.S. Nuclear Regulatory Commission, Washington, DC (ADAMS Accession No.


Purchase information is available through the ASME Web site store at http://www.asme.org/Codes/Publications/.  
ML051520313).  


RG 1.193, Rev. 6, Page 29
8. 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, U.S. Nuclear Regulatory Commission, Washington, D.C.


13.
9.


American Society for Testing and Materials International A 1034/A1034M-05b, Standard Test Methods for Testing Mechanical Splices for Steel Reinforcing Bars, West Conshohocken, PA.3
Code of Federal Regulations, 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), U.S. Nuclear Regulatory Commission, Washington, DC.


14.
10. NUREG/CR-6909, Revision 1, Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials.  U.S. Nuclear Regulatory Commission, Washington, DC.  ADAMS Accession No. ML13150A113.


American Concrete Institute (ACI) 349-06, Code Requirements for Nuclear Safety-Related Concrete Structures and Commentary, Farmington Hills, MI. 4
1 Publicly available NRC published documents are available electronically through the NRC Library on the NRCs public Web site at http://www.nrc.gov/reading-rm/doc-collections/ and through the NRCs 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 NRCs 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 e- mail pdr.resource@nrc.gov.


15.
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/.  


ACI 318, Building Code Requirements for Structural Concrete and Commentary, American Concrete Institute, Farmington Hills, MI.4
DG-1298, Page 30
11. ASTM A 1034/A1034M-05b, Standard Test Methods for Testing Mechanical Splices for Steel Reinforcing Bars, ASTM International, West Conshohocken, PA 3


16.
12. ACI 349-06, Code Requirements for Nuclear Safety-Related Concrete Structures and Commentary, American Concrete Institute, Farmington Hills, MI.4


Electric Power Research Institute, Nuclear Safety Analysis Center 202L, Recommendations for an Effective Flow-Accelerated Corrosion Program, Revision 3, Palo Alto, CA.5 
13. ACI 318, Building Code Requirements for Structural Concrete and Commentary, American Concrete Institute, Farmington Hills, MI.


17.
14. Recommendations for an Effective Flow-Accelerated Corrosion Program (NSAC-202L-R3),
Electric Power Research Institute, Palo Alto, CA.5


CFR, Packaging and Transportation of Radioactive Material, Part 71, Chapter 1, Title 10,  
15. Code of Federal Regulations, Title 10, Energy, Part 71, Packaging and Transportation of Radioactive Material (10 CFR Part 71), U.S. Nuclear Regulatory Commission, Washington, DC.
Energy.


3 The American Society for Testing and Materials (ASTM) is now known as ASTM International. Its standards may be purchased from ASTM International, 100 Barr Harbor Drive, P.O. Box C700, West Conshohocken, Pennsylvania  
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  
19428-2959; telephone (877) 909-2786.  Purchase information is available through the ASTM Web site at http://www.astm.org.
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 ACIs bookstore Web site (http://www.concrete.org/BookstoreNet/bookstore.htm), or by contacting the corporate office at American Concrete Institute, P.O. Box 9094, Farmington Hills, MI 48333; telephone (248) 848-3700; fax (248) 848-370
4 Document 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.


===1.     ===
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 EPRI at Electric Power Research Institute, 3420 Hillview Avenue, Palo Alto, CA 94304, telephone: 650-855-2000, or online at http://my.epri.com/portal/server.pt.}}
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.}}


{{RG-Nav}}
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Revision as of 19:04, 9 January 2025

ASME Code Cases Not Approved for Use
ML16321A338
Person / Time
Issue date: 08/31/2017
From: Giovanni Facco
Office of Nuclear Regulatory Research
To:
Giovanni Facco, 301-415-6337
Shared Package
ML16285A003 List:
References
NRC-2012-0059, RIN 3150-AJ13 RG-1.193, Rev 5
Download: ML16321A338 (30)
v • d • e


U.S. NUCLEAR REGULATORY COMMISSION

August 2017 OFFICE OF NUCLEAR REGULATORY RESEARCH

Revision 5 REGULATORY GUIDE

Technical Lead Giovanni Facco

Written suggestions regarding this guide or development of new guides may be submitted through the NRCs public Web site under the Regulatory Guides document collection of the NRC Library at http://www.nrc.gov/reading-rm/doc-collections/reg-guides/contactus.html.

Electronic copies of this regulatory guide, previous versions of this guide, and other recently issued guides are available through the NRCs public Web site under the Regulatory Guides document collection of the NRC Library at http://www.nrc.gov/reading-rm/doc-collections/.

The regulatory guide is also available through the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html, under ADAMS Accession No. ML16321A338. The regulatory basis for this guide is the regulatory analysis prepared for the amendment to 10 CFR 50.55a (ADAMS Accession No. ML16285A013. The staff responses to the public comments on DG-1298 may be found under ADAMS Accession No. ML16285A012.

REGULATORY GUIDE RG 1.193 (Draft was issued as DG-1298, dated March 2016)

ASME CODE CASES NOT APPROVED FOR USE

A. INTRODUCTION

Purpose

This regulatory guide 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.

Applicability

This RG applies to reactor licensees subject to 10 CFR Part 50, Section 50.55a, Codes and Standards.

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), Section 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 (Ref. 2), or equivalent quality standards.

Section 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) (Ref. 3), or equivalent quality standards.

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.

RG 1.193, Page 2

Related Guidance

Regulatory Guide 1.84, Design, Fabrication, and Materials Code Case Acceptability, ASME

Section III (Ref. 4), 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.

Regulatory Guide 1.147, Inservice Inspection Code Case Acceptability, ASME Section XI,

Division 1 (Ref. 5), 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.

Regulatory Guide 1.192, Operation and Maintenance Code Case Acceptability, ASME OM

Code (Ref. 6), lists the OM 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 regulatory guide 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 NRCs 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 regulatory guide is being issued for information only and does not contain any new or amended information collection requirements.

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.

RG 1.193, Page 3

B. DISCUSSION

Reason of Revision

Revision 5 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 2007 Edition and Supplements 0 through 10

to the 2010 Edition, and the OM Code Cases listed in the 2009 Edition through the 2012 Edition. This is an update to RG 1.193, Revision 4, that included information from Supplements 1 through 10 to the 2007 Edition (Sections III and XI), and the 2002 Addenda through the 2006 Addenda 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 2007 Edition through Supplement 10 to the 2010 Edition. Revision 37 of Regulatory Guide 1.84 and Revision

18 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 reviewed the OM Code Cases listed in the 2009 Edition through the 2012 Edition. Revision 2 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

RG 1.193, Page 4

C. STAFF REGULATORY GUIDANCE

Licensees may not implement Code Cases from the Section III and Section XI Codes listed in Supplement 11 to the 2010 Edition through Supplement 10 to the Edition, and the OM Code Cases listed in the 2009 Edition through the 2012 Edition that are listed in this guide without prior NRC approval.

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.

Table 1. Unacceptable Section III Code Cases

CODE CASE

NUMBER

TABLE 1 UNACCEPTABLE SECTION III CODE CASES

DATE OR

SUPPLEMENT/

EDITION

SUMMARY

N-201-6 Class CS Components in Elevated Temperature Service,Section III,

Division 1

4/10E

Code Case is applicable for high temperature applications beyond that of light water reactors.

N-284-1 Metal Containment Shell Buckling Design Methods,Section III,

Division 1, Class MC

5/9/03

RG 1.193, Page 5

CODE CASE

NUMBER

TABLE 1 UNACCEPTABLE SECTION III CODE CASES

DATE OR

SUPPLEMENT/

EDITION

SUMMARY

(1)

The following errata, misprints, recommendations, and errors have been identified:

Fig. 1511.1, The curve for L 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 il given for spherical shells in accordance with -1512.

-1521, (i) In (a) Axial Compression, G = L should be changed to G = L. (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 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 Ch = 0.92/(M - 0.636) should be changed to Ch = 0.92/(M - 0.636).

-1712.1.1-1, The leftmost curve should be labeled Ch.

-1712.2.2, (a) Axial Compression, (i) In the formula for ej, the denominator should be (m/Lj)2 t. (ii) The expressions for C

and C 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 t11/4.

should be corrected to t11/4.

RG 1.193, Page 6

CODE CASE

NUMBER

TABLE 1 UNACCEPTABLE SECTION III CODE CASES

DATE OR

SUPPLEMENT/

EDITION

SUMMARY

N-284-1 (contd)

Metal Containment Shell Buckling Design Methods,Section III,

Division 1, Class MC

5/9/03

-1713.1.1, (i) The equation a=el/FS should be changed to a=Lel/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=ra on the vertical axis should be changed to Ks=ha.

-1713.2.1, (i) The headings for (b) and (c) should include the words In-Plane. (ii) In (b) Axial Compression Plus Shear, should be changed to .

(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

5/7/99

2/25/02

The Code Case lacks sufficient detail to ensure that the supplied material is as represented by the Certified Material Test Report.

N-510

N-510-1 Borated Stainless Steel for Class CS Core Support Structures and Class 1 Component Supports,Section III, Division 1

12/9/93

8/14/01

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. 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 Annulled 2/3/03

Code Case is applicable to only one DOE aluminum vessel.

N-530

Provisions for Establishing Allowable Axial Compressive Membrane Stresses in the Cylindrical Walls of 0-15 Psi Storage Tanks, Classes 2 and 3

2/3/03

There are numerous errors in the equations. The errors must be corrected before the Code Case can be approved for use.

RG 1.193, Page 7

CODE CASE

NUMBER

TABLE 1 UNACCEPTABLE SECTION III CODE CASES

DATE OR

SUPPLEMENT/

EDITION

SUMMARY

N-565 Alternative Methods of Nozzle Attachment for Class 1 Vessels

12/3/99

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.

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

Requirements for Spent Fuel Storage Canisters,Section III, Division 1

2/26/99

9/24/99

12/8/00

04/08/02 Annulled 10/14/11

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 (Ref. 7), and Interim Staff Guidance No. 18 (ISG-

18), Rev. 1 (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.

N-645 N-645-1 Use of Rupture Disk Devices on Nuclear Fuel Storage Canisters, Class 1,Section III, Division 1

6/14/00

2/3/03 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

9/17/02

11/18/03

The NRC conditionally approved Code Case N-659 in Revision 34 of Regulatory Guide 1.84. The NRCs 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.

N-659-2 Use of Ultrasonic Examination in Lieu of Radiology for Weld Examination,Section III, Divisions 1 and 3

6/09/08

RG 1.193, Page 8

CODE CASE

NUMBER

TABLE 1 UNACCEPTABLE SECTION III CODE CASES

DATE OR

SUPPLEMENT/

EDITION

SUMMARY

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.

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 addend

a. 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 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

7/01/05

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.

RG 1.193, Page 9

CODE CASE

NUMBER

TABLE 1 UNACCEPTABLE SECTION III CODE CASES

DATE OR

SUPPLEMENT/

EDITION

SUMMARY

N-673 Boron Containing Power Metallurgy Aluminum Alloy for Storage and Transportation of Spent Nuclear Fuel,Section III, Division 1

8/7/03

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 high- temperature conditions.

N-693 Alternative Method to the Requirements of NB-3228.6 for Analyzing Piping Subjected to Reversing Dynamic Load,Section III, Division 1

5/21/03

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).

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

11/02/04

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-717 Requirements for Construction of Storage Containments for Spent Nuclear Fuel and High Level Radioactive Waste and Material, Section III,

Division 3

5/04/04

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.

RG 1.193, Page 10

CODE CASE

NUMBER

TABLE 1 UNACCEPTABLE SECTION III CODE CASES

DATE OR

SUPPLEMENT/

EDITION

SUMMARY

N-721 Alternative Rules for Linear Piping Supports,Section III, Division 1

9/09/08 Code Case N-721 allows the use of ANSI/AISC N690L-03, Load and Resistance Factor Design (LRFD) Specification for Safety-Related 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.

The LRFD method is a probabilistic method developed to provide uniform practice in the design of steel structures for nuclear facilitie

s. 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-to-date. 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

10/11/05 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 Use of Polyethylene (PE) Plastic Pipe,Section III, Division 1 and XI

3/22/07

6/10E

Issues have been raised concerning joining methods, qualification and testing as related to joining procedure, the effectiveness of non-destructive examination (NDE) to detect fabrication flaws, potential degradation processes and susceptibility, and the ability of NDE to detect service- related degradation. The NRC believes that additional technical studies are required to assess these issues and make a final regulatory decisio

n. NRC

staff continue to interact with the cognizant ASME committees, and the industry is working to provide additional data and information.

RG 1.193, Page 11

CODE CASE

NUMBER

TABLE 1 UNACCEPTABLE SECTION III CODE CASES

DATE OR

SUPPLEMENT/

EDITION

SUMMARY

N-761 Fatigue Design Curves for Light Water Reactor (LWR) Environments,Section III, Division 1

3/10E

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 (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.
  • 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.

RG 1.193, Page 12

CODE CASE

NUMBER

TABLE 1 UNACCEPTABLE SECTION III CODE CASES

DATE OR

SUPPLEMENT/

EDITION

SUMMARY

N-761 (contd)

Fatigue Design Curves for Light Water Reactor (LWR) Environments,Section III, Division 1

3/10E

  • 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

4/10E

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.

RG 1.193, Page 13

CODE CASE

NUMBER

TABLE 1 UNACCEPTABLE SECTION III CODE CASES

DATE OR

SUPPLEMENT/

EDITION

SUMMARY

N-791 (contd)

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 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 Fatigue Evaluations Including Environmental Effects,Section III,

Division 1

3/10E

Code Case N-792 provides guidance on the use of Fen factors to address the effect of reactor water environment on cyclic damage in Class 1 components. Research results detailed in Welding Research Council Bulletin 487 and Argonne National Laboratory (ANL) report NUREG/CR-

6909 show that there is a possibility that reactor water environment may have an adverse effect on the fatigue damage for typical metals used in Class 1 components. Since Section III does not provide specific guidance in the area of environmental fatigue effects, this Code Case has been developed to provide a Code approved method. The Code Case uses the methodology and Fen equations suggested in NUREG/CR-6909. One major change in the Code Case compared to NUREG/CR-6909 is the deletion of the strain threshold.

However, based on industry comments that the Fen expressions give Fen values greater than 1.0 for situations when environmental effects have no impact, there are ongoing activities at NRC to modify Fen expressions. The NRCs Office of Nuclear Regulatory Research (RES) with the assistance of experts at Argonne National Laboratory (ANL) is pursuing this effort.

N-793 Extruded Steel Sleeves With Parallel Threaded Ends,Section III, Division

2

4/10E

See comments for N-791.

RG 1.193, Page 14

CODE CASE

NUMBER

TABLE 1 UNACCEPTABLE SECTION III CODE CASES

DATE OR

SUPPLEMENT/

EDITION

SUMMARY

N-794 Swaged Splice With Threaded Ends,Section III

See comments for N-791.

N-796 Alternative Preheat Temperature for Austenitic Welds in P-No. 1 Material Without PWHT,Section III, Division 1

See comments for N-791.

N-804 Alternative Preheat Temperature for Austenitic Welds in P-No. 1 Material Without PWHT,Section III, Division 1

7/10E

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 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

7/10E

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.

N-812 Alternate Creep-Fatigue Damage Envelope for 9Cr-1Mo-V Steel,Section III, Division 1

7/10E

Code Case N-812 utilizesSection III, Division, Subsection NH, Class 1 Components in Elevated Temperature Service. Subsection NH is not approved for use by the NRC.

N-818 Alternative Requirements for Preservice Volumetric and Surface Examination,Section III, Division 1

8/10E

RG 1.193, Page 15

CODE CASE

NUMBER

TABLE 1 UNACCEPTABLE SECTION III CODE CASES

DATE OR

SUPPLEMENT/

EDITION

SUMMARY

N-818 (contd)

The NRC has been conducting research at Pacific Northwest National Laboratory on the examination of austenitic and ferritic welds. The work has shown that performing a full volume examination for fabrication flaws is significantly different from an inservice examination. For example, examination from two directions is necessary to detect certain circumferentially oriented fabrication flaws such as lack of fusion. The work has also shown that the second leg of V-path can be applied to ferritic materials on a limited basis but will be difficult to apply to austenitic materials and dissimilar metal welds. Another finding is that surface conditions are critical with respect to detecting and characterizing fabrication flaws. Additionally, the PNNL research suggests that the ability to consistently and accurately characterize fabrication flaws by type (i.e., planar or volumetric) is difficult. This capability is essential if acceptance criteria based on flaw type is to be applied. In summary, the NRC believes that an analytical approach for the acceptance of certain fabrication flaws could be acceptable if appropriately justified and the scope limited to ferritic materials. The NRC believes that significant research will be required to demonstrate that full-volume examination for fabrication flaws is acceptable for austenitic and dissimilar metal welds.

N-820

Twisting of Horizontal Prestressing Tendons,Section III, Division 2

8/10E

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.

N-828 Alternative Nonmetallic Material Manufacturers and Constituent Suppliers Quality System Program Requirements,Section III, NCA-3900,

2010 Edition, and Earlier Editions and Addenda,Section III, Divisions 1 and 2

10/10E

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.

DG-1298, Page 16

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.

Table 2. Unacceptable Section XI Code Cases

CODE CASE

NUMBER

TABLE 2 UNACCEPTABLE SECTION XI CODE CASES

DATE OR

SUPPLEMENT

/EDITION

SUMMARY

N-465 N-465-1 Alternative Rules for Pump Testing,Section XI, Division 1

11/30/88 Annulled

2/14/03

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. Note that Revision 12 of Regulatory Guide 1.147 approved N-465 for use.

The disapproval of N-465 for use applies only to new users.

N-473 N-473-1 Alternative Rules for Valve Testing,Section XI, Division 1

3/8/89 Annulled

2/14/03

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. Note that Revision 12 of Regulatory Guide 1.147 approved N-473 for use.

The disapproval of N-473 for use applies only to new users.

N-480

Examination Requirements for Pipe Wall Thinning Due to Single Phase Erosion and Corrosion,Section XI, Division 1 Annulled 9/18/01

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

Code Case N-498-4 is conditionally approved in Revision 13 of Regulatory Guide 1.147. Those licensees choosing to implement this Code Case are to implement Revision 4, which is listed in Revision

15 of Regulatory Guide 1.147.

DG-1298, Page 17 CODE CASE

NUMBER

TABLE 2 UNACCEPTABLE SECTION XI CODE CASES

DATE OR

SUPPLEMENT

/EDITION

SUMMARY

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

7/23/02

The following concerns were identified during review of the Code Case:

(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.

N-542 Alternative Requirements for Nozzle Inside Radius Section Length Sizing Performance Demonstration,Section XI, Division 1 Annulled 3/28/01

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.

N-547 Alternative Examination Requirements for Pressure Retaining Bolting of Control Rod Drive (CRD) Housings,Section XI, Division 1 Annulled 5/20/01

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 service- related 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.

N-560

N-560-1 N-560-2 Alternative Examination Requirements for Class 1, Category B-J Piping Welds,Section XI, Division 1

8/9/96

2/26/99

2/14/03

(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.

DG-1298, Page 18 CODE CASE

NUMBER

TABLE 2 UNACCEPTABLE SECTION XI CODE CASES

DATE OR

SUPPLEMENT

/EDITION

SUMMARY

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

12/31/96

3/28/01

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.

N-562 N-562-1 Alternative Requirements for Wall Thickness Restoration of Class 3 Moderate Energy Carbon Steel Piping,Section XI, Division 1

12/31/96

3/28/01

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.

N-574 NDE Personnel Recertification Frequency,Section XI, Division 1 Annulled 7/14/06

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.

N-575 Alternative Examination Requirements for Full Penetration Nozzle-to-Vessel Welds in Reactor Vessels with Set-On Type Nozzles,Section XI, Division 1

2/14/03

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.

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.

DG-1298, Page 19 CODE CASE

NUMBER

TABLE 2 UNACCEPTABLE SECTION XI CODE CASES

DATE OR

SUPPLEMENT

/EDITION

SUMMARY

N-578 N-578-1 Risk-Informed Requirements for Class 1, 2, and 3 Piping, Method B,

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.

N-587 Alternative NDE Requirements for Repair/Replacement Activities,Section XI, Division 1 Annulled

2/14/03

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.

N-589 N-589-1 Class 3 Nonmetallic Cured-in-Place Piping,Section XI, Division 1

4/19/02

7/23/02

(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.

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 Annulled 4/8/02

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.

DG-1298, Page 20

CODE CASE

NUMBER

TABLE 2 UNACCEPTABLE SECTION XI CODE CASES

DATE OR

SUPPLEMENT

/EDITION

SUMMARY

N-591 Alternative to the Requirements of Subsection IWL, Requirements for Class CC Concrete Components of Light-Water Cooled Plants,Section XI, Division 1 Annulled 4/8/02

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.

N-593-1 Examination Requirements for Steam Generator Nozzle-to-Vessel Welds,Section XI, Division 1

10/08/04 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-2500-7(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.

N-613 Ultrasonic Examination of Full Penetration Nozzles in Vessels, Examination Category B-D, Item Nos. B3.10 and B3.90, Reactor Vessel- To-Nozzle Welds, Fig. IWB-2500-7(a), (b), and (c),Section XI, Division 1

7/30/98

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.

N-615 Ultrasonic Examination as a Surface Examination Method for Category B-F and B-J Piping Welds,Section XI, Division 1

7/28/01

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.

DG-1298, Page 21 CODE CASE

NUMBER

TABLE 2 UNACCEPTABLE SECTION XI CODE CASES

DATE OR

SUPPLEMENT

/EDITION

SUMMARY

N-618 Use of a Reactor Pressure Vessel as a Transportation Containment System,Section XI, Division 1

6/17/03

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).

N-622 Ultrasonic Examination of RPV and Piping, Bolts, and Studs,Section XI, Division 1 Annulled on

1/12/05

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.

N-653 Qualification Requirements for Full Structural Overlaid Wrought Austenitic Piping Welds,Section XI, Division 1

9/7/01

(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.

N-654 Acceptance Criteria for Flaws in Ferritic Steel Components 4 in. and Greater in Thickness,Section XI, Division 1

4/17/02

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).

DG-1298, Page 22 CODE CASE

NUMBER

TABLE 2 UNACCEPTABLE SECTION XI CODE CASES

DATE OR

SUPPLEMENT

/EDITION

SUMMARY

N-691 Application of Risk-Informed Insights to Increase the Inspection Interval for Pressurized Water Reactor Vessels,Section XI, Division 1

11/18/03

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.

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

1/05/06

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 NRCs responsibility for determining whether examinations are impractical, and eliminates the NRCs ability to take exception to a licensees proposed action and impose additional measures where warranted in accordance with 10 CFR 50.55a(g)(6)(i).

N-713 Ultrasonic Examination in Lieu of Radiography,Section XI, Division 1

11/10/08 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.

N-716 Alternative Piping Classification and Examination Requirements,Section XI, Division 1

4/10/06 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

DG-1298, Page 23 CODE CASE

NUMBER

TABLE 2 UNACCEPTABLE SECTION XI CODE CASES

DATE OR

SUPPLEMENT

/EDITION

SUMMARY

N-716 (contd)

Alternative Piping Classification and Examination Requirements,Section XI, Division 1

4/10/06 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.

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

7/10E

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.

N-729-3 Alternative Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Nozzles,Section XI, Division 1

10/10E

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.

N-740

N-740-1 N-740-2

Dissimilar Metal Weld Overlay for Repair of Class 1, 2, and 3 Items,Section XI, Division 1

10/12/06

12/25/09

11/10/08

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.

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.

DG-1298, Page 24 CODE CASE

NUMBER

TABLE 2 UNACCEPTABLE SECTION XI CODE CASES

DATE OR

SUPPLEMENT

/EDITION

SUMMARY

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

4/10E

(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.

N-780

Alternative Requirements for Upgrade, Substitution, or Reconfiguration of Examination Equipment When Using Appendix VIII Qualified Ultrasonic Examination Systems,Section XI, Division 1

1/10E

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.

DG-1298, Page 25 CODE CASE

NUMBER

TABLE 2 UNACCEPTABLE SECTION XI CODE CASES

DATE OR

SUPPLEMENT

/EDITION

SUMMARY

N-784 Experience Credit for Ultrasonic Examiner Certification

1/10E

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.

N-806 Evaluation of Metal Loss in Class 2 and 3 Metallic Piping Buried in a Back-Filled Trench

10/10E

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 NRCs 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.

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.

N-813 Alternative Requirements for Preservice Volumetric and Surface Examination,Section XI, Division 1

8/10E

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 IWB-

3410-1 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:

(1) The requirements of paragraph IWB-3112 were developed to ensure that defective welds were not placed in servic

e. A preservice flaw

DG-1298, Page 26 CODE CASE

NUMBER

TABLE 2 UNACCEPTABLE SECTION XI CODE CASES

DATE OR

SUPPLEMENT

/EDITION

SUMMARY

N-813 (contd)

Alternative Requirements for Preservice Volumetric and Surface Examination,Section XI, Division 1

8/10E

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.

N-826 Ultrasonic Examination of Full Penetration Vessel Weld Joints in Fig.

IWB-2500-1 Through Fig. IWB-2500-6

10/10E

Reduction of the inspection volume from 1/2 t to 1/2 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 1/2 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 1/2 t to 1/2 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 1/2 t volume.

DG-1298, Page 27

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.

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

2000 Addenda Reaffirmed 2001 Edition Reaffirmed 2003 Addenda Reaffirmed 2004 Edition Reaffirmed 2006 Addenda (see Note)

Reaffirmed 2009 Edition 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.

OMN-15 Requirements for Extending the Snubber Operational Readiness Testing Interval at LWR Power Plants

2004 Edition Revised 2006 Addenda Reaffirmed 2009 Edition Reaffirmed 2012 Edition

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.

DG-1298, Page 28

D. IMPLEMENTATION

The purpose of this section is to provide information to applicants and licensees regarding the NRC

staffs 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 must address the NRCs concerns about the Code Case at issue.

DG-1298, Page 29 REFERENCES1

1. Code of Federal Regulations, Title 10, Energy, Part 50, Domestic Licensing of Production and Utilization Facilities (10 CFR Part 50), U.S. Nuclear Regulatory Commission, Washington, DC.

2. ASME Boiler and Pressure Vessel Code,Section III, Rules for Construction of Nuclear Power Plant Components, Division I; and Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, American Society of Mechanical Engineers, New York, NY.2

3. ASME Code for Operation and Maintenance of Nuclear Power Plants, American Society of Mechanical Engineers, New York, NY.2

4. Regulatory Guide 1.84, Design, Fabrication, and Materials Code Case Acceptability, ASME

Section III, U.S. Nuclear Regulatory Commission, Washington, DC.

5. Regulatory Guide 1.147, Inservice Inspection Code Case Acceptability, ASME Section XI,

Division 1, U.S. Nuclear Regulatory Commission, Washington, DC.

6. Regulatory Guide 1.192, Operation and Maintenance Code Case Acceptability, ASME OM Code, U.S. Nuclear Regulatory Commission, Washington, DC.

7. NRC Spent Fuel Project Office Interim Staff Guidance No. 4 (ISG-4), Rev. 1, Cask Closure Weld Inspections, U.S. Nuclear Regulatory Commission, Washington, DC (ADAMS Accession No.

ML051520313).

8. 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, U.S. Nuclear Regulatory Commission, Washington, D.C.

9.

Code of Federal Regulations, 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), U.S. Nuclear Regulatory Commission, Washington, DC.

10. NUREG/CR-6909, Revision 1, Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials. U.S. Nuclear Regulatory Commission, Washington, DC. ADAMS Accession No. ML13150A113.

1 Publicly available NRC published documents are available electronically through the NRC Library on the NRCs public Web site at http://www.nrc.gov/reading-rm/doc-collections/ and through the NRCs 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 NRCs 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 e- mail 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/.

DG-1298, Page 30

11. ASTM A 1034/A1034M-05b, Standard Test Methods for Testing Mechanical Splices for Steel Reinforcing Bars, ASTM International, West Conshohocken, PA 3

12. ACI 349-06, Code Requirements for Nuclear Safety-Related Concrete Structures and 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.

14. Recommendations for an Effective Flow-Accelerated Corrosion Program (NSAC-202L-R3),

Electric Power Research Institute, Palo Alto, CA.5

15. Code of Federal Regulations, Title 10, Energy, Part 71, Packaging and Transportation of Radioactive Material (10 CFR Part 71), U.S. Nuclear Regulatory Commission, Washington, DC.

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

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

4 Document 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.


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