[Federal Register Volume 61, Number 252 (Tuesday, December 31, 1996)]
[Notices]
[Pages 69120-69124]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 96-33249]


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NUCLEAR REGULATORY COMMISSION

Proposed Generic Communication; Effectiveness of Ultrasonic 
Testing Systems in Inservice Inspection Programs

AGENCY: Nuclear Regulatory Commission.

ACTION: Notice of opportunity for public comment.

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SUMMARY: The Nuclear Regulatory Commission (NRC) is proposing to issue 
a generic letter to determine if addressees are taking appropriate 
action to qualify future ultrasonic testing (UT) examinations. The 
purpose of the proposed generic letter is to (1) alert addressees to 
the importance of using equipment, procedures, and examiners (UT 
systems) capable of reliably detecting and sizing flaws in the 
performance of comprehensive examinations of reactor vessels and 
piping, (2) notify addressees about enhancements in UT systems and the 
significance of these enhancements in plant-specific inservice 
inspection (ISI) programs, (3) request that all addressees describe the 
extent to which their piping and reactor pressure vessel ISI activities 
are being qualified consistent with the objectives of Appendix VIII to 
Section XI of the American Society of Mechanical Engineers Boiler and 
Pressure Vessel Code (ASME Code), and (4) require that all addressees 
send to the NRC a written response to this generic letter relating to 
the actions and information requested in this letter. The

[[Page 69121]]

NRC is seeking comment from interested parties regarding both the 
technical and regulatory aspects of the proposed generic letter 
presented under the SUPPLEMENTARY INFORMATION heading.
    The proposed generic letter was endorsed by the Committee to Review 
Generic Requirements (CRGR) on December 19, 1996. The relevant 
information that was sent to the CRGR will be placed in the NRC Public 
Document Room. The NRC will consider comments received from interested 
parties in the final evaluation of the proposed generic letter. The 
NRC's final evaluation will include a review of the technical position 
and, as appropriate, an analysis of the value/impact on licensees. 
Should this generic letter be issued by the NRC, it will become 
available for public inspection in the NRC Public Document Room.

DATES: Comment period expires January 30, 1997. Comments submitted 
after this date will be considered if it is practical to do so, but 
assurance of consideration cannot be given except for comments received 
on or before this date.

ADDRESSES: Submit written comments to Chief, Rules Review and 
Directives Branch, U.S. Nuclear Regulatory Commission, Mail Stop T-6D-
69, Washington, DC 20555-0001. Written comments may also be delivered 
to 11545 Rockville Pike, Rockville, Maryland, from 7:30 am to 4:15 pm, 
Federal workdays. Copies of written comments received may be examined 
at the NRC Public Document Room, 2120 L Street, N.W. (Lower Level), 
Washington, DC.

FOR FURTHER INFORMATION CONTACT: Donald G. Naujock (301) 415-2767.

SUPPLEMENTARY INFORMATION:

Generic Letter 96-XX: Effectiveness of Ultrasonic Testing Systems In 
Inservice Inspection Programs

Addressees

    All holders of operating licenses or construction permits for 
nuclear power reactors, except those licenses that have been amended to 
possession-only status.

Purpose

    The U.S. Nuclear Regulatory Commission (NRC) is issuing this 
generic letter to (1) Alert addressees to the importance of using 
equipment, procedures, and examiners capable of reliably detecting and 
sizing flaws in the performance of comprehensive examinations of 
reactor vessels and piping, (2) notify addressees about enhancements in 
ultrasonic testing (UT) systems (Note: As used in this document, ``UT 
systems'' refers to the equipment, procedures, or examiners involved in 
the ultrasonic examination) and the significance of these enhancements 
in plant-specific inservice inspection (ISI) programs, (3) request that 
all addressees describe the extent to which their piping and reactor 
pressure vessel ISI activities are being qualified consistent with the 
objectives of Appendix VIII (Note: ``Consistent with the objectives of 
Appendix VIII'' means in close conformance with Appendix VIII criteria, 
even though the Appendix has not been formally incorporated into the 
regulations as a requirement.) To Section XI of the American Society of 
Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), and 
(4) require that all addressees send to the NRC a written response to 
this generic letter relating to the actions and information requested 
in this letter.

Background

    In the 1970s, operating experience and industry tests indicated a 
need for improving UT procedures to consistently and reliably detect 
and characterize flaws during ISI of reactor vessel welds. Also noted 
was the need for more definitive reporting of results and for more 
descriptive requirements for essential variables associated with 
ultrasonic examinations. That need was satisfied with the issuance of 
Regulatory Guide (RG) 1.150, Revision 1, ``Ultrasonic Testing of 
Reactor Vessel Welds During Preservice and Inservice Examinations,'' in 
February 1983. RG 1.150 was incorporated into the technical 
specifications of many plants.
    As the nuclear industry gained more operating experience, the need 
for improving ISI capabilities became apparent. For example, in the 
late 1970s, thermal fatigue cracks were found on the inner-blend radius 
of nozzle-to-vessel surfaces in boiling-water reactor (BWR) feedwater 
and control rod drive return line nozzles. The NRC staff recommended, 
in NUREG-0619, ``BWR Feedwater Nozzle and Control Rod Drive Return Line 
Nozzle Cracking,'' dated November 1980, that licensees develop ISI 
programs to search for cracks in the inner-blend radii using dye-
penetrant, visual, and ultrasonic examinations. The NRC staff 
recognized the potential for improvements to UT systems, and stated in 
NUREG-0619 that demonstrated improvements could be used as the basis 
for modifying the inspection criteria.
    Also in the late 1970s, intergranular stress corrosion cracking 
(IGSCC) was identified in austenitic stainless steel piping. The NRC 
staff recommended in NUREG-0313, ``Technical Report on Material 
Selection and Processing Guidelines for BWR Coolant Pressure Boundary 
Piping,'' dated July 1977, and in subsequent revisions published in 
July 1980 and January 1988, that a program be established to conduct 
formal IGSCC performance demonstration testing for UT examiners.
    The regulatory guide and NUREG reports were issued as guidance in 
detecting flaws and in preventing the conditions that could lead to 
unacceptable flaws.
    The need for additional guidance related to performing UT in ISI 
programs, that were based on requirements in Section XI of the ASME 
Code, prompted a reexamination of the effectiveness of UT as it was 
being applied through the ASME Code. The conventional (amplitude-based) 
UT requirements in the ASME Code establish minimum acceptable 
inspection standards. In the 1970s and 1980s, the nuclear industry 
tested UT systems extensively to identify the critical aspects of an 
effective UT inspection program that would provide a high reliability 
for detection and characterization of flaws. In the mid-1980s, the NRC 
and the nuclear industry recognized that the reliability of UT in ISI 
programs could be significantly improved through performance-
demonstration qualification of nondestructive examination equipment, 
procedures, and examiners.
    In 1984, the NRC entered into an agreement, known as the IGSCC 
Coordination Plan, with the Boiling Water Reactor Owners' Group (BWROG) 
and the Electric Power Research Institute (EPRI) to coordinate selected 
activities in regard to training and qualification of personnel using 
UT to examine piping weldments. As part of the IGSCC Coordination Plan, 
EPRI administered IGSCC performance demonstration tests to personnel 
seeking UT qualifications in IGSCC detection and characterization in 
piping systems.
    The nuclear industry set about changing ASME Code requirements for 
UT from the current minimum inspection standards to inspection 
standards with performance-based qualifications. The performance-based 
qualifications would also produce uniform acceptance criteria for 
evaluating new technology and addressing new forms of degradation. The 
efforts of the industry to develop performance-based qualification 
criteria culminated with the publication of Appendix VIII to Section XI 
of the ASME Code, which was published in

[[Page 69122]]

the 1989 Addenda. Appendix VIII, ``Performance Demonstration for 
Ultrasonic Examination Systems,'' contains detailed requirements for UT 
performance demonstrations that include statistically based acceptance 
criteria to detect and size flaws.
    The NRC has initiated rulemaking to amend 10 CFR 50.55a to 
reference Section XI of the ASME Code up to and including the 1995 
Edition. After completion of rulemaking, Appendix VIII to Section XI 
will become a requirement for all licensees. The final rule 
incorporating Appendix VIII is expected to be issued around July 1998.

Description of Circumstances

    Appendix VIII is based on the qualification of equipment, 
procedures, and examiners using performance demonstrations; whereas, 
existing requirements in the 1989 (and earlier) Edition of Section XI 
of the ASME Code are prescriptive, minimum inspection standards. A 
performance-based qualification program encourages development of 
improved methods for detecting and characterizing flaws, and 
facilitates implementing the methods with a defined testing curriculum. 
The performance demonstrations require that equipment, procedures, and 
examiners be tested on flawed and notched materials and configurations 
similar to those found in actual conditions. The nuclear industry 
created the Performance Demonstration Initiative (PDI) in 1991 to 
manage implementation of the performance demonstration criteria of 
Appendix VIII (Note: The PDI activities have been assessed by the NRC 
staff, as described in the letter from J. Strosnider (NRC) to B. 
Sheffel (PDI) dated March 6, 1996, and have been found to provide a 
significantly improved method for qualification of equipment, 
procedures, and examiners. Overall, the NRC staff found that PDI has 
established and is in the process of executing a well-planned and 
effective program to test UT technicians on selected portions of 
Appendix VIII. Accordingly, the NRC staff finds that UT procedures 
qualified under the PDI program using performance demonstration methods 
provide an acceptable level of quality and safety.)
    Because performance demonstrations test the ability of equipment, 
procedures, and examiners to detect and size flaws, the demonstrations 
raise the performance threshold for examiners conducting ultrasonic 
inspections. For example, a sampling of individuals tested in the 
different piping examinations under the PDI program revealed that 22% 
of them did not satisfy the screening criteria for detection of flaws; 
41% did not satisfy the screening criteria for length-sizing; 67% did 
not satisfy the screening criteria for depth measurement; and 49% did 
not satisfy the screening criteria for IGSCC. These percentages are 
based on a sampling that included retests. The PDI tests ensure that 
the equipment must have adequate sensitivity, the procedures must have 
sufficient detail, and the individuals must be sufficiently skilled in 
order to successfully qualify under the PDI program.
    The improvements in UT techniques performed using Appendix VIII 
criteria became apparent in 1993 during the reactor pressure vessel 
shell weld augmented examination at the Browns Ferry Nuclear Power 
Plant, Unit 3, and in 1995 during the inspection of piping systems for 
IGSCC at the Millstone Nuclear Power Station, Unit 1. At Browns Ferry, 
the equipment, procedures, and examiners were qualified consistent with 
the objectives of Appendix VIII. The examination revealed 15 flaws that 
did not meet the ASME Code, Section XI,
    Subarticle IWB-3500 acceptance criteria and that required further 
evaluation. Of the 15 flaws, only 3 would have been recordable using 
conventional Section XI minimum inspection standards and RG 1.150 
criteria, and only 2 of the 3 flaws would have required an analytical 
evaluation in accordance with Section XI, Subarticle IWB-3600. This 
experience indicates that flaws large enough to require analytical 
evaluation might not be detected using current UT standards.
    Millstone Unit 1 inspectors performed an augmented UT examination 
for IGSCC in the welds in reactor system piping. The licensee used a 
newly developed ultrasonic transducer technology to supplement the 
original examinations. Before the examination, UT examiners from 
Millstone who were qualified under the IGSCC Coordination Plan 
demonstrated the adequacy of the new transducer technology by 
successfully passing the Appendix VIII performance demonstration test 
administered through the PDI program. During the augmented examination, 
the UT inspection personnel examined 264 of the 411 pipe welds and 
found that 35 welds had cracks. A review of examination records from 
1984 through 1994 revealed 211 indications that were previously 
considered by Level III inspectors to be nonmetallurgical or geometric 
indications. During the 1995 inspection, 14 of the indications 
previously identified as nonmetallurgical or geometric were identified 
as flaws; 3 of these flaws developed through-wall leaks when they were 
mechanically buffed in preparation for repair by the NRC-approved 
overlay process. The Appendix VIII qualification by Millstone 
inspectors using normal IGSCC UT procedures increased the licensee's 
reliability in detection of IGSCC. The additionally demonstrated 
capability of the new transducer technology under the PDI-administered 
program clearly increased the level of confidence in the new transducer 
technology used to identify previous errors made in flaw disposition.
    Although, the above experiences clearly depict the need for 
improvement by using performance demonstration methods in performing UT 
examinations of reactor vessels and piping, it should be noted that a 
safety concern does not exist which would warrant immediate backfitting 
of Appendix VIII in advance of the rulemaking that has been initiated. 
The staff has reached this conclusion based on consideration of 
defense-in-depth measures, Code margins in component design, and 
leakage monitoring systems. In addition, the staff has been requiring 
for some time now that selected inspections be performed using 
performance-based qualified techniques (e.g., IGSCC piping 
inspections).

Regulatory Requirements

    10 CFR 50.55a requires that systems and components of boiling-water 
and pressurized-water reactors conform to the requirements of the ASME 
Code, Sections III and XI.
    Appendix A to 10 CFR Part 50 Criterion 14 requires that the reactor 
coolant pressure boundary shall be designed, fabricated, erected, and 
tested so as to have an extremely low probability of abnormal leakage, 
of rapidly propagating failure, and of gross rupture.
    Criterion XVI of Appendix B to 10 CFR Part 50 requires that 
measures shall be established to assure that conditions adverse to 
quality, such as failures, malfunctions, deficiencies, deviations, 
defective material and equipment, and nonconformances are promptly 
identified and corrected. In the case of significant conditions adverse 
to quality, the measures shall assure that the cause of the condition 
is determined and corrective action taken to preclude repetition. The 
identification of the significant condition adverse to quality, the 
cause of the condition, and the corrective action taken shall be 
documented and reported to appropriate levels of management.

[[Page 69123]]

    Criterion II of Appendix B to 10 CFR Part 50 requires, in part, 
that a quality assurance program shall take into account the need for 
special controls, processes, test equipment, tools, and skills to 
attain the required quality and the need for verification of quality by 
inspection and test. It also requires that the program provide for 
indoctrination and training of personnel performing activities 
affecting quality, as necessary to assure that suitable proficiency is 
achieved and maintained.

Discussion

    The qualification statistics from PDI discussed above and the 
issuance of the regulatory guide and staff reports highlight the fact 
that some UT systems satisfying ASME Code, Section XI amplitude-based 
UT requirements are less effective in identifying and characterizing 
certain types of flaws. The experiences at Browns Ferry Unit 3 and 
Millstone Unit 1 highlight the significant improvements in the 
effectiveness of UT systems when equipment, procedures, and examiners 
are qualified through a performance-demonstration program. Therefore, a 
significant improvement is gained in the effectiveness of UT systems 
qualified through performance demonstrations (e.g., Appendix VIII) over 
those satisfying conventional Section XI amplitude-based UT 
requirements.
    The early and accurate detection of flaws in plants is important 
for maintaining the structural integrity and ensuring the safety 
function of safety-related systems and components. As plants age, 
improved reliability in inspection methods, more flexibility in 
utilizing advanced technology, and a better ability to detect new forms 
of degradation gain increased importance in ISI programs. The nuclear 
industry recognizes Appendix VIII as an improvement over the current 
ISI requirements, and the NRC staff finds that Appendix VIII criteria, 
as implemented by the PDI program, provide UT results that generally 
are superior to those of the 1989 (and earlier) Edition of Section XI 
of the ASME Code. The NRC staff finds that implementation of Appendix 
VIII criteria enhances the reliability of inspections and provides a 
significant improvement in the methods used to satisfy existing 
regulatory requirements and assure plant safety.
    Some licensees have already submitted requests to utilize Appendix 
VIII performance demonstrations as an alternative examination for 
selective ASME Code, Section XI requirements. Licensees have also 
submitted requests to the staff to use Appendix VIII criteria in lieu 
of criteria in Regulatory Guide 1.150. Some licensees are using 
Appendix VIII concepts in developing alternatives to the IGSCC 
Coordination Plan, and the NRC staff has already approved the use of 
either the PDI program or the original IGSCC program for IGSCC 
qualification of examiners

(Note: Letter from W. T. Russell (NRC) to K. P. Donovan (Chairman, 
Boiling Water Reactor Owners' Group), ``Transition From the IGSCC 
Qualification Program to the Performance Demonstration Initiative 
Program,'' March 1, 1996.)

    In conclusion, the NRC staff has determined that using only 
existing ISI requirements for performing UT examinations might not 
provide reasonable assurance that flaws can be reliably detected and 
sized in certain areas. The staff considers cracks and flaws in the 
reactor vessel and other safety-related components to be a concern when 
the possibility exists for flaws exceeding the ASME Code, Section XI 
allowable flaw sizes not being reliably detected or sized. Adequate 
safety exists through defense-in-depth measures, leakage monitoring 
systems, and Code margins in component design; however, significant 
improvement in the ability to reliably detect and size flaws in reactor 
vessels and piping can be achieved using performance demonstration 
methods. In order to assess whether the margins required by the ASME 
Code, Section XI are adequately maintained and to ensure compliance 
with the applicable existing requirements identified above, the NRC has 
concluded that it is appropriate to request certain actions and 
information from the addressees, as indicated below.

Requested Actions

    In consideration of the information and concerns addressed above, 
each addressee is requested to perform an evaluation to determine 
whether its current ISI program ensures that flaws in the reactor 
vessel and safety-related piping are reliably detected and sized.
    If it is determined that flaws in the reactor vessel and safety-
related piping cannot be reliably detected and sized, each addressee is 
expected to take appropriate corrective action in future inspections, 
in accordance with the requirements of Criteria II and XVI of Appendix 
B to 10 CFR Part 50, to improve the capability to reliably detect and 
size flaws.

Requested Information

    Within 90 days of the date of this generic letter, addressees are 
requested to submit a written summary report that includes the 
following:
    1. A brief description of the addressee's evaluation of its ISI 
program, its determination regarding the capability of its current 
program to reliably detect and size flaws, and corrective actions taken 
(if any) in response to the requested actions above.
    2. If the addressee is not using and does not plan to use the 
criteria in Appendix VIII of the ASME Code Section XI or other 
performance-based methods for the qualification of ISI activities, then 
provide a discussion of any plans for ensuring the effectiveness of 
current UT systems in detecting and sizing flaws in the reactor vessel 
and safety-related piping.
    3. If the addressee is using or plans to use Appendix VIII for the 
qualification of ISI activities, then discuss the extent to which the 
equipment, procedures, and examiners in your ISI program for the 
reactor vessel and safety-related piping are (or will be) qualified 
using Appendix VIII criteria or other performance-based methods. 
Include in this discussion a description of any alternate examination 
methods (i.e., IWA-2240 of ASME Code Section XI) in your ISI program 
that use Appendix VIII or other performance-based examination methods 
as allowed in applicable sections of 10 CFR 50.55a for inspecting the 
reactor vessel and safety-related piping.

Required Response

    Within 30 days of the date of this generic letter, addressees are 
required to submit a written response indicating: (1) Whether or not 
the requested actions will be completed, (2) whether or not the 
requested information will be submitted, and (3) whether or not the 
requested information will be submitted within the requested time 
period.
    Addressees who choose not to complete the requested actions, or 
choose not to submit the requested information, or are unable to 
satisfy the requested completion date, must describe in their response 
any alternative course of action that is proposed to be taken, 
including the basis for establishing the acceptability of the proposed 
alternative course of action. [For addressees that fail to have or 
implement appropriate qualification methods for future UT examinations 
where subsequent inspections find previously unidentified or improperly 
dispositioned flaws, the staff will consider whether such circumstances 
(a) are the result of failing to adequately take into account the need 
for special controls, skills and training needed to ensure suitable 
proficiency in the conduct of UT examinations contrary to the 
requirements of Criterion II, Quality

[[Page 69124]]

Assurance Program, of Appendix B ``Quality Assurance Criteria for 
Nuclear Power Plants and Fuel Reprocessing Plants,'' of 10 CFR Part 50; 
and/or (b) represent inadequate corrective action for known 
inadequacies contrary to the requirements of Criterion XVI, Corrective 
Action, of Appendix B, of 10 CFR Part 50.]
    Address the required written responses to the U.S. Nuclear 
Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 
20555-0001, under oath or affirmation under the provisions of Section 
182a, Atomic Energy Act of 1954, as amended, and 10 CFR 50.54(f). In 
addition, send a copy to the appropriate regional administrator.

Related Generic Communications

    (1) Information Notice 96-32, ``Implementation of 10 CFR 
50.55a(g)(6)(ii)(A), Augmented Examination of Reactor Vessel,'' June 5, 
1996.
    (2) Information Notice 93-20, ``Thermal Fatigue Cracking of 
Feedwater Piping to Steam Generators,'' March 24, 1993.
    (3) Generic Letter 88-01, ``NRC Position on IGSCC in BWR Austenitic 
Stainless Steel Piping,'' January 25, 1988.

Backfit Discussion

    This generic letter transmits an information request pursuant to 
the provisions of Section 182a of the Atomic Energy Act of 1954, as 
amended, and 10 CFR 50.54(f) to determine if licensees are taking 
appropriate action to qualify future UT examinations. To the extent 
that the actions requested in this letter may result in corrective 
actions taken by addressees that are considered backfits, the backfits 
are justified under the compliance exception of the backfit rule, i.e., 
10 CFR 50.109 (a)(4)(i).

    Dated at Rockville, Maryland, this 23rd day of December, 1996.

    For the Nuclear Regulatory Commission.
David B. Matthews,
Acting Director, Division of Reactor Program Management, Office of 
Nuclear Reactor Regulation.
[FR Doc. 96-33249 Filed 12-30-96; 8:45 am]
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