ML20138G303

Requests That CRGR Review & Endorse Proposed GL, Effectiveness of Ultrasonic Testing Sys in Inservice Insp Programs

ML20138G303
Person / Time
Issue date:	10/18/1996
From:	Thadani A NRC (Affiliation Not Assigned)
To:	Jordan E Committee To Review Generic Requirements
References
TAC-M95373, NUDOCS 9610240054
Download: ML20138G303 (19)

Text

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! s* j' UNITED STATES NUCLEAR REGULATORY COMMIS810N i

g p/ WASHINGTON, D.C. 20006-0001 Ococber 18, 1996

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I MEMORANDUM TO: Edward L. Jordan, Chairman ,

Committee To Review Generic Requirements  !

FROM:

Ashok C. Thadani, Acting Deputy Director

.A / M ' f Office of Nuclear Reactor Regulation .$ e ~

i

SUBJECT:

i REQUEST FOR REVIEW AND ENDORSEMENT OF PROPOSED GENERI t LETTER, " EFFECTIVENESS OF ULTRASONIC TESTING SYSTEMS IN' l

i INSERVICE INSPECTION PROGRAMS" (TAC NO. M95373) i The Office of Nuclear Reactor Regulation (NRR) requests that the Committee To

! Review Generic generic letter. Requirements (CRGR) review and endorse the subject proposed i Following endorsement, the proposed generic letter will be published'in the Federal Register for public comment.

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Attachment 1 is the generic letter as proposed by the staff. The NRC is

issuin~g this generic letter to (1) notify addressees about enhancements in

! ultrasonic testing (UT) systems and the significance of these enhancements in 4

plant-specific inservice inspection (ISI) programs, (2) request that all

addressees describe how they can ensure that flaws in the reactor vessel and i

safety-related piping are adequately detected, (3) request that all addressees describe the extent to which criteria in Appendix VIII to Section XI of the ASME Boiler and Pressure Vessel Code (ASME Code) are used to qualify ISI i ,activitles, and (4) require that all addressees send the NRC a written l

response to this generic letter relating to the requested information.

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Attachment 2 is the staff's response to the questions contained in Section

• IV.B of the CRGR Charter. The responses to these questions document the -

justification for the required responses regulated by 10 CFR 50.54(f).

) 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 UT equipment, procedures, and i examiners. The nuclear industry set about changing ASME Code requirements for -

i UT from the current minimum inspection standards endorsed in 10 CFR 50.55a to inspection standards with performance-based qualifications. The performance-

! based qualification criteria culminated with the publication of Appendix VIII i[

to time,Section XI ofVIII Appendix thehas ASME Code not been in the 1989 endorsed in 10EditionCFR 50.55a. with 1989 Addenda. At this I:

j CONTACT:~

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In 1991, the industry formed the Performance Demonstration Initiative (PDI) to ,

4 implement Appendix VIII with an alternative to the Intergranular Stress i Corrosion Cracking (IGSCC) Coordination Plan. The PDI has been used for ,

qualifying equipment, procedures, and examiners since 1994. The industry has  !

funded the PDI program for the past five years, and funding is authorized only j

to the end of 1996. Future funding'is dependent upon regulatory action. The issuance of the generic letter is expected to fill the regulatory void until

rulemaking is complete. Although the NRC is developing a rule to make

• j- Appendix VIII to Section XI a requirement, the final rule is not expected to be issued until July 1998. The proposed generic letter will encourage licensees to commit to important improvements in the effectiveness of ISI programs and will demonstrate N.RC support of the industry's current initiative .'

in this area. Since the NRC would like to receive the responses to the i requested information quickly, a 30-day public comment period is proposed. t

. The Office of the General Counsel (OGC) reviewed this generic letter and

raised no legal objections. It should be noted that the proposed GL as written, " requires" information from addressees rather than " requests"

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information. This is consistent with the memorandum from S. G. Burns to E. L.

Jordon, " Appropriate Language for 50.54(f) Letters," dated August 29, 1996.

j A decision is pending on the " rule" status of the proposed generic letter under the Small Business Regulatory Enforcement Fairness Act (5 U.S.C.,

i Chapter 8), enacted on March 29, 1996.

j The generic letter is sponsored by Brian W. Sheron, Director, Division of i

Engineering.

Attachments

1. Proposed Generic Letter, " Effectiveness of Ultrasonic Testing Techniques a

in Inservice Inspection Programs" l 2. Response to CRGR Charter Questions b DISTRIBUTION:

File Center EMCB RF/PF ACRS DOCUMENT NAME: G:\NAUJOCK\CRGR COV.ER

• See Previous Concurrence 4 To receive a copy of this document, indicate in the bax: "C" - Copy without attachment / enclosure "E" - Copy with attachment / enclosure "N" - No copy

'EMCB: LPM

• lE EMCB:SC* lE EMCB:BC* lE TechEd* IN j, DNau.iock:adi DTerao JRStrosnider RSanders l 8/21/96 8/22/96 8/28/96 9/03/96 DE:DD* l C- DE:D* lC OGC* DRPM:D* l 3 GClainas BWSheron LClark ITMartin i 09/04/96 10/01/96 ,, 09/26/96 10/03/96

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110/02/96 /t16/96 i OFFICIAL RECORD COPY l ,

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ATTACHMENT 1 i

UNITED STATES

NUCLEAR REGULATORY COMNISSION 0FFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C. 20555-0001 GENERIC LETTERINSPECTION 96-XX:PROGRAMS EFFECTIVENESS OF ULTRASONIC TESTIN Addressees ,-

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

Purcose i

The U.S. Nuclear Regulatory Commission (NRC) is issuing this generic letter to -

(1) notify addressees about enhancements in ultrasonic testing (UT) systems' and the significance of these enhancements in plant-specific inservice inspection (ISI) programs, (2) request that all addressees describe the extent to which their piping and reactor pressure vessel ISI activities are being qualified in the spirit of Appendix VIII to Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code); and (3) require that all addressees send to the NRC a written response to this generic i

letter relating to the required information.

Backaround

! 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. I 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 Dur'ing Preservice and Inservice

' Examinations," in February 1983. This regulatory guide was incorpor"ed into the technical specifications of many plants. '

As the nuclear industry gained more operating experience, the need for  !

l improving ISI capabilities became apparent. For example, in the late 1970s, thermal fatigue cracks were found on the inner-blend radius of nozzle-to-l vessel surfaces in boiling-water reactor (BWR) feedwater and control rod drive

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return line noizles. The NRC staff recommended, in NUREG-0619, "BWR Feedwater Nozzle and Contre) Rod

• Drive Return Line Nozzle Cracking," dated November 1980, that licensees develop ISI programs to search for cracks in the inner-4 blend radii using dye-penetrant, visual, and ultrasonic examinations. The NRC staff recognized the potential for improvements to UT systems, and stated in i

As used in this document, the term, "UT systems" refers to the equipment, procedures, or examiners involved in the ultrasonic examination.

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l GL 96-XX August XX, 1996 1 Page 2 of 9 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 staff 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 progrims, 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 currentqualifications.

based minimum inspection standards to inspection standards with performance-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 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.

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l Gl. 96-XX August XX, 1996  !

Page 3 of 9 Descriotion 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 i

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, - 1 procedures, and examinets be tested on flawed and notched materials and configurations similar to those' found in actual conditions. The nuclear industry created the Performance Demonstration Initiative manageimplempntationoftheperformancedemonstrationrequ(PDI)in1991to irements of Appendix VIII . '

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 i

t under the PDI program revealed that 22% of them failed to qualify for detection of flaws; 41% failed to qualify for length-sizing; 67% failed to qualify for depth measurement; and 49% failed to qualify for IGSCC. These

' percentages are based on a sampling that included retests. Th PDI tests ensure that the equipment must have adequate sensitivity, the procedures must have sufficient detail, and the individuals must be suffic'iently skilled in order to successfully qualify under,the PDI program.

The improvements in UT techniques performed using Appendix VIII requirements became apparent in 1993 during th'e 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, i

2 The PDI activities have been assessed by the NRC staff, as described i 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- l planned and effective program to test UT technicians on selected portions of l Appendix VIII. Accordingly, the NRC staff finds that UT procedures qualified '

under the PDI program using performance demonstration methods provide an i acceptable level of quality and safety. 1 1

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i GL 96-XX August XX, 1996 Page 4 of 9 and examiners were qualified in the " spirit of Appendix VIII" requirements3 .

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. -

l 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.

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Before the ex' amination, 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, i

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 l

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 ident'ified as nonmetallurgical or i 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.

I Reaulatory Reauirements Section 50.55a.to Part' 50 of Title 10 of the Code of Federal Regulations (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. It also requires that structures, systems, and components be designed, fabricated, erected, constructed, tested, and inspected to quality standards commensurate with the importance of the safety function to be performed.

Appendix B to 10 CFR Part 50 requires, in part, that a quality assurance 3'

The " spirit of Appendix VIII" means that the equipment, procedure, and -

examiners demonstrated the performance and conducted the inspection as closely as possible to Appendix VIII requirements.

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l I. GL 96-XX j August XX, 1996 .

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{ program shall take into account the need for special controls, processes, test 4 equipment, tools, and skills to attain the required quality and the need for i

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 i is achieved and maintained.

i Appendix B to 10 CFR Part 50 also requires that measures be established to i assure that nondestructive qualified testing is controlled and accomplished by qualified i personnel standards, spec using,ifications, procedures in accordance with applicable criteria, and other special requirements. It codes, i

further requires that material be examined, measured, or tested to assure quality.

j i 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

• 1 satisfying ASME Code,Section XI amplitude-based UT requirements are less -

I effective in identifying and characterizing certain types of flaws. The

experiences at Browns Ferry Unit 3 and Millstone Unit I highlight the '

i significant improvements in the effectiveness of UT systems when equipment, procedures, program. and examiners are qualified through a performance-demonstration i

i Therefore, a significant improvement is gained in the effectiveness i

of UT systems qualified through performance demonstrations (e.g., Appendix j VIII) over those satisfying conventional Section XI amplitude-based UT requirements.

The early and accurate detection of flaws in plants is essential for

j. maintaining the structural integrity and ensuring the safety function of i safety-related systems and components. 'As plants age, improved reliability in j' inspection methods, more flexibility in utilizing advanced technology, and a

better ISI programs.ability to detect new forms of degradation gain increased importance in In order to ensure that inservice inspections adequately assess

the structural integrity of safety-related components in operating nuclear power plants and to ensure that a sound basis exists for conducting fracture

[. mechanics analyses to evaluate flaws in safety-related components, the UT j

systems used for ISI must have high reliability for detection and characterization of flaws. .

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The nuclear industry recognizes Appendix VIII as an improvement over the l 3

current ISI requirements, and the NRC staff finds that Appendix VIII requirements, as implemented by the PDI program, provide UT results that j generally are superior to those of the 1989 (and earlier) Edition of Section ,

i XI of the ASME Code. The NRC staff finds that implementation of Appendix VIII j j,

criteria enhances the reliability of inspections and provides a significant improvement in the methods used to satisfy existing regulatory requirements

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l GL 96-XX i August XX, 1996 Page 6 of 9 ,

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i and assure plant safety. I

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Some licensees have already submitted requests to utilize Appendix VIII i

performance demonstrations as an alternative examination for selective ASME l Code,Section XI requirements. Licensees have also submitted requests to the  !

staff to use Appendix VIII criteria in lieu of criteria in Regulatory Guide I l.150. Some licensees are using Appendix VIII concepts in developing alternatives to the IGSCC Coordination Plan, and the NRC staff has already  ;

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. approved the use of either the Pp! program or the original IGSCC program for i IGSCC qualification of examiners l l

In con,clusion,.the NRC staff has determined that existing 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 safety concern when the possibility exists for exceeding the ASME Code,Section XI allowable flaw sizes or when cracks or flaws are sufficiently deep -

and are not reliably detected or sized. .Therefore, in order to assess whether the margins required by the ASME Code,Section XI are maintained and to verify -

compliance with regulatory requirements of Appendix'B to 10 CFR Part 50 as-they pertain to verifying quality by in~ spection and to using qualified procedures and personnel in nondestructive testing, the NRC has concluded that it is appropriate for licensees to submit the following information.

1 Reauired Information

• Addressees are required to submit the following information: l

1. In consideration of the concerns addressed above, discuss how your ISI program ensures that flaws in the reactor vessel and safety-related piping are adequately detected and sized.

2. The s.taff considers the use of Appendix VIII to be an acceptable basis to ensure that flaws are reliably detected and sized. If you are not using or do 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 discuss your plans for ensuring the effectiveness of your UT systems in detecting and sizing flaws in the reactor vessel and safety-related piping.

3. If you are using or plan to use Appendix VIII for the qualification of ISI ,

activities, then discuss the extent to which the equipment, procedures, and 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.  ;

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GL 96-XX August XX, 1996 Page 7 of 9 examiners piping are in your ISI program for the reactor vessel and safety-related performance-(or will be) qualified using Appendix VIII criteria or other based methods. . Include in your 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 in lieu of currently-required Section XI examination methods for inspecting the reactor vessel and safety-related piping.

Reauired Response .

All addressees shall submit in writing the information required above within 120 days from the date of this generic letter.

• Address the required written reports 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.

The NRC recognizes the potential difficulties (number and types of sources, age of records, proprietary data, etc.) that licensees may encounter while ascertaining whether they have all of the data pertinent to the requested assessme'nts. For this reason, 120 days are allowed for the responses.

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 Aust'enitic Stainless Steel Piping," January 25, 1988.

• Backfit Discussion This generic letter only requires information from the addressees under the i provisions of Section 182a of the Atomic Energy Act of 1954, as amended, and I 10 CFR 50.54(f); therefore, the staff has not performed a backfit analysis.

The staff is not establishing a new position for such compliance in this l

generic letter. Therefore, this generic letter does not constitute a backfit i and no documented evaluation or backfit analysis need be prepared.

Federa7 Reafster Notification (to be changed before issuing) =

A notice of opportunity for public comment will be published in the federal

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GL 96-XX August XX, 1996 Page 8 of 9 i

Register with a 30-day coment period since the NRC needs to receive the responses to the generic letter quickly.

Coments on the technical issue (s) addressedATTN:

Comission, by this generic letter may be sent to the U.S. Nuclear Regulatory Document Control Desk, Washington, D.C. 20555-0001.

Panerwork Reduction Act Statement This generic letter contains information collections that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). These information collections number' were approved by the Office of Management and. Budget, approval 3150-0011, which expires on July 31, 1997.

The public reporting burden for this collection of information is estimated to

average 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> per response, including the time for reviewing instructions, searching existing data sources, gathering and maintaining the data needed, and c'o mpleting and reviewing the collection of information. The U.S. Nuclear Regulatory Comission is seeking public coment on the potential impact of the collection issues

of information contained in the generic letter and on the following

' (1) Is the proposed collection of information necessary for the proper performance of the functions of the NRC,' including whether the information will have practical utility? '

(2) Is the estimate of burden accurate?

(3) Is there a way to enhance t'he quality, utility, and clarity of the 1

information to be collected?

(4) How.can the burden of the collection of information be minimized, e.g.,

including the use of automated collection techniques?

Send coments on any aspect of this collection of inNrmation, including suggestions for reducing this burden, to the Information and Records Management Branch, T-6F33, U.S. Nuclear Regulatory Comission, Washington, DC

' 20555-0001, and to the Desk Officer, Office of Information and Regulatory ,

Affairs,NE08-10202'(3150-0012), Office .of Management and Budget, Washington, l DC 20503. i The NRC may not conduct or sponsor, and a person is not required to respond to, a request for the collection of information unless it displays a currently valid OMB control number.

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GL 96-XX August XX, 1996 Page 9 of 9 If you have any questions about this matter, please contact the technical contact listed below or the appropriate NRR project manager.

Thomas T. Martin, Director Division of Reactor Program Management Office of Nuclear Reactor Regulation Technical contact: Donald Naujock, NRR -

(301) 415-1767 INTERNET:DGN9NRC. GOV Lead Project Manager: Karen Cotton, NRR (301) 415-1538 INTERNET:KRCONRC. GOV 9

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. ATTACHMENT 2 Response to CRGR Charter Question

! PROPOSED ACTION:

Issue a generic letter (GL on the effectiveness of  !

2 ultrasonic testing (UT) sys)tems used for inservice inspection (ISI) programs. This GL is a request for j information to verify compliance. The NRC staff is  ;

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requesting a 30-day public comment period.

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CATEGORY: 2 RESPONSE TO REOUIREMENTS FOR CONTENT OF PACKAGE SUBMITTED j (1) i The proposed generic requirement or staff position as it is proposed

' to be sent out to ifconsees. Where the objective or intended result

of a proposed generic requirement or staff position can be achieved by setting a readily quantifiable standard that has an unambiguous relationship to a readily measurable quantity and is enforceable, the i proposed requirement should merely specify the objective or result to

( be attained, rather than prescribing to the licensee how the objective or result is to be attained.

i i- Each holder of an operating license for,a boiling-water or l

pressurized-water reactor facility, except those licenses that have been amended to possession-only status is required (1 to discuss i

how its inservice inspection program en,sures that flaws) 'in the i .

reactor vessel and safety-related piping are adequately detected and

sized, (2)*to discuss its plans for ensuring the effectiveness of UT

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systems not qualified using pt formance-based methods in detecting

! and sizing flaws in the reactor <e'ssel and safety-related piping, and (3) to discuss the extent to which the equipment, procedures, and examiners in its ISI program are qualified using Appendix VIII of ASME Code, Se' c tion XI or other performance-based methods.

(ii) Draft staff papers or other underlying staff documents supporting the requirements or, staff positions. (A copy of all materials referenced i'

in the document shall be made available upon request to the CRGR staff.

Any Counittee member may request CRGR staff to obtain a copy of any reference material for his or her use.)

The following three references discuss problems that have occurred using current ASME Code requirements.

(1)

NRC Insnection Renort: 50-327. 50-328/92-09. March 23 Aoril 10. 1992 During March 1992, Sequoyah Uni.t I experienced a leak in the spool piece weld in the steam generator-to-feedwater transition area despite augmented ISI. The augmented inspections performed using UT techniques showed indications that revealed the cracks, but the licensee's examiners misinterpreted.the indications as resulting from geometric configurations of the pipe. After finding the leak, the licensee i

radiographed all feedwater nozzles of both units and found cracks in five of the eight nozzles. The five nozzles were repaired.

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-(2) Information Notice 93-20 f i

As a result of the Sequoyah incident, the staff issued Information i Notice 93-20, " Thermal- Fatigue Cracking of Feedwater Piping to Steam i Generators." In this information notice, the staff concluded that l

-inspection techniques specified in Section XI of the ASME Code did not  !

appear adequate for finding cracks.

In Information Notice 93-20, the staff told the industry about this type  !

of failure for the second time. The industry had been alerted in 1979 l to the potential for the cracking in feedwater nozzle-to-safe end welds  ;

through the issuance of IE Bulletin 79-13, " Cracking-in Feedwater System i Piping."

l (3a) Letter From G.C. Wrioht. NRC. to G.B. Slade. Consumers Power Company.  !

September 7. 1993. with Inspection Reoort: 50-255/93005(DRS) l (3b) Letter From Consumers Power to NRC. " Pressurizer Safe End Crack Renair f Action Plan." October 4.1993  !

In June 1993, the staff assigned to the NRC NDE mobile lab discovered a l crack in the dissimilar metal weld between the pressurizer power- i operated relief valve and nozzle safe end at Palisades. Palisades  !

personnel performed a supplemental ultrasonic examination on the  !

indications and characterized them as geometry. During the supplemental i ultrasonic examination, the minimum amplitude standard that existed in  !

1989 and earlier editions of Section XI of the ASME Code was not. i exceeded. The licensee considered this indication as a non-recordable '

indication. The licensee considered the subject weld acceptable for  !

continued service. During the September 1993 plant heatup, an  !

unisolable leak was discovered at this weld, j

i The next two ASME Code references recognize 'the value of Appendix VIII.

(4) American Society of Mechanical .Enaineers. Boiler and Pressure Vessel Code. of Section XI. 1989 Edition. Anoendix III Mandatory Sunclement 4 In this supplement, ASME recommends that examiners and procedures be qualified using welded samples and simulated or actual flaws, or both, located in positions in which geometry may make flaws more difficult to detect. ASME does not specify the method or acceptance criteria to be used for qualification, but states that the requirements for the qualifications of examiners and procedures are being prepared. Without a uniform method or acceptance criteria, or both, a large variation in effectiveness can be expected in the industry.

(5) American Society of Mechanical Enaineers. Boiler and Pressure Vessel Code. of Section XI. 1989 Edition With 1989 Addenda. Accendix VII Mandatory To become an ASME Code, UT Level II examiner, Appendix VII requires a practical examination that consists of a sample set with at least five l

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flaws that do not exceed the standards of Subarticle IWB-3500, actual i

8 flaws, or a mixture of both. Alternately, successful completion of a UT i i

[ performance this practicaldemonstration examination.in accordance with Appendix VIII may serve as  :

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The next reference shows the deficiencies of current ASME-qualified Level II UT examiners demonstrating their proficiency. ,

(6) American Society oF Mechanical Enaineers. Boiler and Pressure Vossel Code. of Section XM. 1989 Edition With 1989 Addenda. Annendix VLII i

Currently, Appendix VIII is.not a requirement. Appendix VIII as .

implemented by the Performance Demonstration Initiative j i

i qualification program for equipment, procedures, and exam (PDI) iners.provides The a -

program uses material with flaws and configurations similar to those found in nuclear power plants. A random sampling of 27 ASME Code, UT j

• Level II examiners tested in the different piping examinations under the 4 PDI program revealed that' 22% of the examiners failed to qualify for detection of flaws; 41% failed to qualify for length-sizing; 67% failed

to qualify for depth measurement; and 49% failed to qualify for IGSCC.

These results suggest an industry-wide need for improving the

! effectiveness flaws. of equipment, procedures, and examiners to. find reportable 1 .

The next twodemonstrations.

performance references show the benefits from using Appendix VIII-type 1

j (7) i Letter From P. Salaa to NRC. " Browns Ferry Nuclear Plant. Unit 3 Reactor Pressure Vessel Shetl Welds Auamented Examination and Inservice Insoection relief Reauest 3 ISI-17." March 6. 1995 4

j In the submittal from Browns Ferry Nuclear Power Plant t thc licensee described its augmented reactor vessel exam (BFN), inationUnit using 3, UT techniques that followed the spirit of-Appendix VIII req ~uirements. The examination found 15 flaws that did not meet the ASME Code,Section XI, 3

Subarticle IWB-3500 acceptance criteria and required further evaluation.

i Of the 15 flaws, only 3 would have been recordable using conventional Section XI minimum inspection standards and RG I.150 criteria, and only l

• 2 of these 3 flaws would have required an analytical evaluation in i

'accordance with Section XI, Subarticle IWB-3600. This experience I

indicates that flaws which require analytical evaluation might not be i

detected when amplitude-based UT techniques are used.

(8)

Millstone 029-00. and Nuclour FoT Power Station. Unit 1. Licensee Event Reoort LER 95-owun Phone Calls i

i Millstone Unit 1 inspectors performed an augmented ultrasonic  !

examination for IGSCC in the welds in reactor system piping. The i j licensee used a newly developed UT transducer technology to supplement the original examinations. Before the examination, UT examiners from  ;

i Millstone who had been qualified according to the IGSCC Coordination j Plan (NUREG-0313), demonstrated the adequacy of the new transducer 3

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' technology by the PDI program. During the augmented examination, the Millstone inspectors examined 264 of 411 pipe welds and found that 35 i welds had cracks. A review of examination records from 1984 through l 1994 revealed 211 indications that were previously considered by Level III inspectors to be nonmetallurgical or geometric indications. During l l

l the 1995 inspection, 14 of the previously identified nenmetallurgical or i

geometric indications were identified as flaws; 3 of these flaws developed through-wall leaks when they were mechanically buffed in preparation for repair by NRC-approved weld overlay process. The Appendix VIII qualification, by Millstone inspectors using the normal IGSCC UT procedures, improved the licensee's reliability in. detecting IGSCC. The additional demonstrated capability of the new transducer i' technology under the PDI-administered program clearly incieased the level of confidence in the new transducer technology used to uncover  !

previous errors made in flaw disposition.

(iii) Each proposed requirement or staff position shall contain the sponsoring I office's position as to whether the proposal would. increase requirements or staff positions, implement existing requirements or staff positions, ,

i

, or would relax or reduce existing requirements or staff positions.

Not applicable. The propos.ed generic letter is for information only. I (iv) The proposed method of implementation with the concurrence (and any  !

comments) of 0GC on the method proposed. The concurrence of affected  !

program offices or an explanation of any nonconcurrences.

See attached concurrence page.  :

i (v) Regulatory analyses conforming to the ' directives and guidance of NUREG/BR-0058 and NUREG (This does not apply for backfits that ensure compliance or e/CR-3568. nsure, define, or redefine adequate protection. In these  ;

cases a documented evaluation is required as discussed in IV.B.(ix).)

Not applicable. .

(vi) Identification of the category of reactor plants to which the generic requirement or staff position is to apply (that is, whether it is to apply to new plants only, new OLs only, OLs after a certain date, OLs before a certain date, all OLs, all plants under construction, all plants, all water reactor , all PWRs only, some vendor tips, some vintage types such as BWR 6 and 4, jet pump and nonjet pump plants, etc.).

All holders of operating licenses or construction permits for nuclear power

• reactors, except those licenses that have been amended to possession-only status. .

(vii) For backfits other than compliance or adequate protection backfits, a backfit analysis as. defined,in 10 CFR 50.109. The backfit analysis shall include, for each category of reactor plants, an evaluation which demonstrates how the action should be prioritized and scheduled in light

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of other ongoing regulatory activities. The backfit analysis shall  !

document for consideration information available concerning any of the following factors as may be appropriate and any other information  !

relevant and material to the proposed action:

(a) Statement of the specific' objectives that the proposed action is designed to achieve; -

(b) General description of the activity that would be required by the licensee or applicant in order to complete the action; (c) Potential chan release of radfe in thematerial; onctive risk to the public from the accidental l (d) Potential impact on radiological exposure of facility employees and i other onsite workers; (e) Installation and continuing costs associated with the action, including the cost of facility downtime or the cost of construction i

delay; l

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i (f) The potential safety impact of changes in plant or operational complexity, including the relationship of proposed and existing regulatory requirements and staff positions;

! (g) The citimated resource burden on the NRC associated with the proposed l

action and the availability of resources;-

(h) The potential impact of differences in facility type, design, or age .

on the relevancy and practicality of the proposed action; e

(i) Whether the proposed action is interim or final, and if interim, the 1

justification for imposing the proposed action on an interin basis;

(j) How the action should be prioritized and scheduled in light of other j

ongoing regulatory activities. The following information may be appropriate in this regard:

4

1. The proposed priority or schedule, 4

2. A summary of the current backlog of exis, ting requirements awaiting i implementation, -

) 3. An assessment of whether implementation of existing requirements i

should be deferred as a result,.and

! 4. Any other information that may be considered appropriate with regard

{ to priority, schedule, or cumulative impact. For example, could implementation be delayed pending public comment?

Not applicable.

i (viii) For each backfit analyzed pursuant to 10 CFR 50.109(a)(2) (i.e., not adequate protection backfits and not compliance backfits), the i proposing Office Director's~ determination, together with the rational for the determination based on the consideration of paragraph (1) and

{ , (vii) above, that: .-

• i (a) There is a substantial increase in the overall protection of public health and safety or the common defense and security to be derived from the proposal; and (b) The direct and indirect co::ts of implementation, for the facilities affected, are justified in view of this increased protection.

Not applicable. .

(ix) For adequate protection or compliance backfits evaluated pursuant to 10 CFR 50.109(a)(4)(ii)

(a) a documented evaluation consisting of:

(1) the objectives (2) the reasons for (3) the basis for invoking the compliance or adequate protection exemption.

l

(b) in addition, for actions that were immediately effective (and therefore issued without prior CRGR review as discussed in III.C) the evaluation shall document the safety si of the action taken and ' if applicable)gnificance consideration and appropriateness of how costs contributed to selecting (the solution among various acceptable l alternatives. .

i Not applicable. The information requested by the proposed generic letter l is under 10 CFR 50.54(f).  !

(x) For each evaluation conducted for proposed relaxations or decreases in '

current requirements or staff positions, the proposing Office Director's determination, together with tie rationale for the determination based on the considerations or paragraphs (1) through (vii) above, that:

(a) The public health and safety and the cosmon defense and security j would be adequately protected if the proposed reduction in 1 i

requirements or positions were implemented, and .

(b) The cost savings attributed l

i to justify taking the action,to the action would be substantial enough Not applicable because information requested by the proposed generic letter '

j under 10 CFR 50.54(f)'will be used to evaluate compliance with existing regulations or conformance with written commitments.

j i

(xi) For each request for information under.10 CFR 50.54(f) (which is not subject to exception as discussed in III.A) an evaluation that includes i at least the following elements:

(a),

j A problem statement that describes the need for the information in terms of potential safety benefit.

j (b) The licensee actions required and the cost to develop a response to the information request.,

(c) An anticipated schedule for NRC use of the information.

2 (d) A statement affirming that the request does not' impose new requirements on the licensee, other than Br the requested

information.

i'

The NRC staff considers cracks and flaws in the reactor vessel and other

safety-related components to be a safety concern when the possibility i exists for exceeding the ASME Code,Section XI allowable flaw sizes or when sufficiently deep cracks or flaws are not reliably detected or sized.

• J Therefore, in order to assess that the' margins required by the ASME Code,

Section XI are maintained and to assess compliance with the regulatory

' requirements of Appendix B to 10 CFR Part 50 as they pe'rtain to the i

effectiveness of inspection and to using qualified procedures and personnel in nondestructive testing, the staff requires the addressees to prepare a

] rritten response assessing the UT systems being used in'their piping and b

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reactor pressure vessel ISI programs. Since all addressees are participating in the PD1 program, the staff believes that the collection and submittal expense to them. of the information in a docketed form will impose minimal The information requirement does not impose new requirements on licensees, except for the requirement to submit the information identified in the draft generic letter.

(xti) An assessment of how the proposed action relates to the Commission's Safety Goal Policy Statement.

l Not applicable. The proposed generic letter is for information only.

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