[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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