ML20198F153
| ML20198F153 | |
| Person / Time | |
|---|---|
| Issue date: | 01/08/1998 |
| From: | Shapaker J NRC (Affiliation Not Assigned) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-97-06, GL-97-6, SECY-97-289-R, TAC-M95290, NUDOCS 9801090307 | |
| Download: ML20198F153 (7) | |
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UNITED STATES y
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PlUCLEAR REGULATORY COMMISSION 1.,*****){
WASHINGTON, D.C. 3006H001 a
January 8, 1998 a
MEMORANDUM 10:
Document Control Desk Information and Records Management Branch Information Management Division Office of the Chief Information Officer FROM:
James W. Shapaker V
i Events Assessment. Generic C nications and Spe:ial Inspections Branch Division of Reactor Program Management Office of Nuclear Reactor Regulation
SUBJECT:
DOCUMENTS ASSOCIATED WITH NRC GENERIC LETTER 97-06,
' DEGRADATION OF STEAM GENERATOR INTERNALS (TAC NO. M95290)
The Materials and Chemical Engineering Branch (EMCB) in the Divisison of Engineering (DE) prepared the subject generic letter, which was issued on December 30, 1997, and given accession number 9712180168.
There is material related to the subject generic letter that should be placed in the NRC Public Document Room and made cvailable to the public. Therefore, by copy of this memorandum, I am providing the following documents to the NRC Public Document Room:
(1) a copy of the published version of the subject generic letter. (2)
-a copy of the information paper (SECY-97-289) that was sent to the Commission, (3) a copy of each letter. received in response to the notice of oportunity for public comment on the proposed generic letter that was published in the FederaT Register on December 31, 1996. (4) a copy of the summary and resolution of public comments that were received, and (5) a copy of the CRGR review package.
I request that you provide me with the Nuclear Documents System accession number for this memorandum. This information may be provided by telephone
-(415-1151)-or by e-mail (Jws).
In addition, please modify the appropriate
'NUDOCS entries to reflect the fact that /.a documents identified herein are related to Generic Letter 97-06.
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i UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C. 20555 December 30,1997 NRC GENERIC LETTER 97-06: DEGRADATION OF STEAM GENERATOR INTERNALS MdInten All holders of operating licenses for pressurized water reactors (PWRs), except those who have permanently ceased operations and have certified that fuel has been permanently removed from the reactor vessel.
Purnose The U.S. Nuclear Regulatory Commission (NRC) is issuing this generic letter to (1) again alert addressees 'o the proviously communicated findings of damage to steam generator internals, namely, tube support plates and tube bundle wrappers, at foreign PWR facilities; (2) alert addressees to recent findings of damage to steam generator tube support plates at a U.S. PWR facility; (3) emphasize to addressees the importance of performing comprehensive examinations of steam generator internals to ensure steam generator tube structuralintegrity is maintained ir accordance with the requirements of Appendix B to 10 CFR Part 50; and (4) require all addressees to submit information that will enable the NRC staff to verify whether addressees' steam generator internals comply with and conform to the current licer. sing bases for their respective facilities.
Background
The NRC issued information Notice (lN) 96-09 and IN 96-09, Supplement i to alert atessees to findings of damage to steam generator internals at foreign PWR facilities.
Queriotion of Circumstances Foreign authorities have reported various steam generator tube support plate damage mechanisms. The affected steam generators are similar, but not identical, to Westinghouse Model 51 steam generators. As previously documented in IN 96-09 and IN 96-09, Supple-ment 1, one damage mechanir,m involved the wastage of the uppermost support plate caused by the misapplication of a chemical cleaning process. A second damage mechanism involved broken tube support plate ligaments at the uppermost, and sometimes at the next lower, tube support plates. The support plate ligaments broke near a radial seismic restraint and near an antirotation ksy; the damage apparently dates back to initial startup of the affected plants.
According to foreign authorities, the ligaments may have broken becaue of excessive stress
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during the final thermal treatment of the monobloc steam generators, which in turn was caused by inadequate clearance for differential thermal ex, ansion lietween the support plates, n
wrapper, and seismic restraints.
As previously documented in IN 90-09, Supplement 1, a third damage mechanism involved wastage not associated with chemical cleaning and affected tube support plates at various elevations. This damage mechanism is active (progressive) and apparently involves a corrosion or erosion-corrosion mechanism of undetermined origin.
The staffs of potentially affected foreign reactors are currently inspec;ing steam generators for evidence of the various damage mechanisms, both visually and with eddy current testing.
Tubes without adequate lateral support are being plugged.
IN 06 09, Supplement 1, also documented that cooling transients involving the injection of large quantities of auxiliary feedwater may have been a key factor in the steam generator wrapper drop phenomenon observed at a foreign PWR facility. These cooling transients are believed to have been particularly severe for two units as a result cf the use of a special operating procedure to accelerate the transition from hot to cold shutdown. The weight of the wrapper assembly and support plates is borne by six tenons mounted on the steam generator shell. The wrapper is nominally free to expand axially relative to the shell. However, it is postulated that an interference fit developed between the wrapper and the seismic restraints (mounted to the shell) as c result of differential thermal expansion associated with the cooling transients at the seventh support plate elevation. This interference fit prevented axial expansion of the wrapper, which led to excessive vertical bearing loads at the tenon supports, thus causing localized wrapper failure at this location and downward displacement of the wrapper (20 millimeters maximum). Poor quality wrapper support welds may also have contributed to this failure.
Repairs have been made at the affected foreign PWR facility, Wrapper dropping is being monitored in all steam generators of similar design. The monitoring is performed through both online instrumentation and visualinspections during outages. In addition to the wrapper dropping problem, cracking of the wrapper above the original upper support was discovered at the same fore!gn unit. The cause of the cracking is not yet known.
At a U.S. PWR facility, degradation of steam generator tube eggcrate supports was discovered through secondary side visualinspections performed during the spring 1997 refueling outage.
The licensee identified erosion corrosion as the damage mechanism; the cause is not yet known. The damage appears to be confined to the periphery and the ur. tubed staywell region of the tube bundle. The oggerate degradation at the periphery extends inward to the first one or two rows of tubes. The degradation at the staywell region primarily affects the support structures within the untubed section. Damage to the eggerate supports was found in both steam generators on the hot and cold leg sides although the damage was more extensive on the hot leg side. No degradation of eggerate supports was identliied within the tube bundle.
DiscusslCD The reported foreign and U.S. experience highlights the potential for degradation mechanisms that may lead to tube suoport plate and tube bundle wrapper damage. The steam generator tube support plate. support the tubes against lateral displacement and vibration and minimize
GL 97 06 December 30,1997 Page 3 of 5 bending moments in the tubes in the event of an accident. Support plate damage can impair the support plate's ability to perform this function and, thus, could potentially lead to the impairment of tube integrity. Vibration induced fatigue could present a potential problem if tube support plates lose integrity, particularly in areas of high secondary side cross flows. As previously noted in IN 06 09, tube support plate signal anomalies found during eddy current testing of the steam generator tubes may be indicative of support plate damage or ligament cracking. Certain visual and video camera inspections on the secondary side of a steam generator may also provide usefulinformation concoming the degradation of steam generator intemals. The NRC staff will continue to monitor information on tube support plate and tube bundle wrapper damage as it become, available.
This letter also alerts addressees to the importance of performing comprehensive examinations of steam generator internals to ensure steam generator tube structuralintegrity is maintained in accordance with the requirements of Appendix B to 10 CFR Part 50. More specifically, Critorion XVI of Appendix D, Corrective Action," requires, in part, that measures be established to assure that conditions adverse to quality are promptly identified and corrected.
Requitad Informatl0D Within 90 days of the date of this generic letter, each addressee is required to provide a written report that includes the following information for its facility:
(1)
Discussion of any program in place to detect degradation of steam generator internals and a description of the inspection plans, including the inspection scope, frequency, methods, and equipment.
The discussion should include the following information:
(a)
Whether inspection records at the facility have been reviewed for indications of tube support plate signal anomalies from eddy current testing of the steam generator tubes that may be indicative of support plato damage or ligament cracking. If the addressee has performed such a review, include a discussion of the findings.
(b)
Whether visual or video camera inspections on the secondary side of the steam generators have been performed at the facility to gain information on the condition of steam generator internals (e.g., support plates, tube bundle wrappers, or other cominnents), if the addrcssee has performed such inspections, include a discussion of the findings.
(c)
Whether dec adation of steam generator internals has been detected at the facility, and how the degradation was assessed and dispositioned.
(2)
If the addressee currently has no program in place to detect degradation of steam generator internals, include a discussion and justification of the plans and schedule for establishing such a program, or why no program is needed.
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GL G7 06 December 30,1997 Page 4 of 5 Addressees are encouraged to work closely with industry groups on the coordination of Inspections, evaluations, and repair options for all types of steam generator degradation that may be found.
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The NRC is aware that the industry has developed generic guidance on performing steam generator inspections, and that this guidance is continually being updated. If an addressee intends to follow the guidance developed by the industry for this issue, reference to the relevant generic guidance documents is acceptable, and encouraged, as part of the response, as long as the referenced documents have been officially submitted to the NRC. However, additional plant specific information will be needed.
NRC staff will review the responses to this generic letter and if concerns are identified, affected addressees will be notified.
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).
DEMtLDhcuulon Under the provisions of Section 182a of the Atomic Energy Act of 1954, as amorded, and 10 CFR 50.54(f), this generic letter transmits an information request for the purpose of verifying compliance with applicable existing regulatory requirements. Specifically, the requested information will enable the NRC staff to determine whether the condition of the addressees' steam generator internals comply with and conform to the current licensing bases for their respective facilities. In particular, the information would help the staff to ascertain whether the regulatory requirements pursuant to Appendix B to 10 CFR Part 50 are met.
No backfit is either intended or approved in the context of issuance of this generic letter.
Therefore, the staff has not performed a backfit analysis.
EDderalRegister Notificall0D A notice of opportunity for public comment was published in the Federal Regisfer on December 31,1990 (61 FR 69116).
Panerwork Reduction ActStatement This genenc letter contains information collections that are subject to the Paperwork Reduction Act of 1995 (22 U.S.C. 3501 et seq.). These information collections were approved by the Office of Management and Budget, approval number 3150-0011, which expires on September 30,2000, The public reporting burden for this collection of information is estimated to average 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> per response, including the time for reviewing instructions, searching existing data sources, gathering and maintaining the data needed, and completing and reviewing the collection of l
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GL 97 06 December 30,1997 Page 5 of 5 t
information. The U.S. Nuclear Regulatory Commission is seeking public comment on the potentialimpact of the collection of information contained in the generic letter and on the following issues:
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(1) is the proposed collection of information necessary for the proper performance of the i
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 the quality, utility, and clarity of the information to be collected? -
(4) ' How can the burden of the collection of information be minimized, including the use of
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automated collection techniques?
Send comments on any aspect of this collection of information, including suggestions for
- reducing this burden, to the Information and Records Management Branch, T 6F33, U.S. Nuclear Regulatory Commission, Washington, DC 20555 0001, and to the Desk Officer, Office of information and Regulatory Affairs, NEOB 10202 (3150-0011), Office of Management and Budget, Washington, DC 20503.
The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of information unless it displays a currently valid OMB control number.
If you have any questions about this matter, please contact the technical contact listed below or the appropriate Office of Nuclear Reactor Regulation (NRR) project manager, i
A Jack W. Roe, Acting Director Division of Reactor Program Management Office of Nuclear Reactor Regulation Technical contact: Stephanie M. Coffin, NRR 301-415 2778 E mail: smci@nrc. gov Lead Project Manager: George F, Wunder, NRR 301-415 1494 E mail:- gfw@nrc, gov -
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Attachment:
List of Recently issued NRC Generic Letters
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Attachment GL 97-06 December 30,1997 Page 1 of 1 LISLQf_ REC E NTLY ISSUED 3ENERIC_1.EIIERS GENERIC DATE OF LEIIER -
SUBJEC1_
ISSUANCE ISSUED TO 97 05 Steam Generator Tube 12/17/97 All holders of OLs for inspection Techniques pressurized-water reactors, except those who have permanently ceased operations and have enrtified that fuel has been perman-ently removed from the reactor vessel 90 00, Assurance of Equipment 11/13/97 All holders of OLs for nuclear Sup. 1 Operability and Containment power reactors except those integrity During Design Basis who have permtnently Accident Conditions ceased operations and have certified that fuel has been permanently removed from the reactor vessel 91 18, Information to Licensees 10/08/97 All holders of OLs for nuclear Rev,1 Regarding NRC Inspection power and NPRs, including Manual Section on Resolution those power reactor of Degraded and Nonconform-licensees who have per-ing Conditions manently ceased operations, and all holders of NPR licenses whose license no longer authorizes operation 97-04 Assurance of Sufficient Nel 10/07/97 All holders of OLs for nuclear Positive Suction Head for power plants, except those Emergency Core Cooling who have permanently and Containment Heat ceased operations and have Removal Pumps certified that fuel has been permanently removed from the reactor vessel 97-03 Annual Financial Surety 07/09/97 Uranium recovery licensees Update Requirements for and state officials Uranium Recovery Licensees OP = Operating License CP = Construction Permit NPR = Nuclear Power Reactors
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POLICY ISSUE (Information)
December 15, 1997 SECY-97-289 EDB:
The Commissioners EROM:
L. Joseph Callan Executive Director for Operrstions
SUBJECT:
PROPOSED NRC GENERIC LETTER, " DEGRADATION OF STEAM GENERATOR INTERNALS" PURPOSE:
To inform the Comndssion, in accordance with the guide-o in a memorandum dated December 20,1991, from Samuel J. Chilk to James M. Taylor regarding SECY 91-172,
- Regulatory impact Survey Report Final," of the staff's intent to issus the attached generic l
letter. The generic letter discusses findings of damage to steam generator intemals at foreign and U.S. pressurized water reactor (PWR) facilities and requires addressees to submit information that will enable the NRC staff to verify whether steam generator internals comply with and conform to the current licensing bases for their respective facilities.
QlSCUSSION:
The generic letter alerts addressees to reported damage to steam generator internals at foreign and U.S. PWR facilities. These reports highlight the potential for degradation mechanisms that i
mry lead to steam ger'erator tube support plate and tube bundle wrapper damage. The steam generator tube support plates support the tubes against lateral displacement and vibration and minimize bending moments in the tubes in the event of an accident. Support plate damage can impair the support plate's ability to perform this function and, thus, could potentially lead to the impairment of tube integrity. Vibration-induced fatigue could present a potential problem if tube support plates lose integrity, particularly in areas of high riecondary side crossflows.
Implications of a complete tall of the steam generator tube bundle wrapper have been assessed by a foreign utility and include the potential for loss of feedwiter, damage to the largest radius tube U-bends, loose parts, and tube rupture.
The generic letter requires that addressees submit information that will enable the staff to determine whether the condition of the addressees' steam generator intemals comply with and SECY NOTE:
To be made publicly CONTACT: Stephanie M* Coffin' NRR available in 5 working days from 415-2778 the date of this paper l
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conform to the current licensing bases for theit respective facilities, in particular, the information wou '1 help the staff to ascertain whether the regulatory requirements pursuant to Appendix B to 1,-CFR Part 50 are met.
On December 31,1996, the staff placed a notice in the Federal Register (61 FR 69116) publishing for comment the draft generic letter, " Degradation of Steam Generator Intemals."
At the end of the % day comment period, the staff P'
- received three public comments.
Two comments wt a from industry organizations, and the third was from a licensee, The primary concem of both industry organizations was that the generic letter had the effect of imposing actions on addressees to inspect steam generator intemais, in response to this comment, the staff reiterated that the detection of steam generator intemals' degradation through inspection will help to ensure that steam generator tube structural integrity is maintained in accordance with the existing requirements of Appendix B to 10 CFR Part 50; the generic letter does not impose new or revised Stati positions or requirements. The staff also noted that the generic letter does not require addressees to inspect steam generator intemals. The generic letter requests information regarding past inspection practices, plans for future inspections, and/or justification as to why inspections are not warranted, given foreign and U.S. experience with steam generator intemal degradation. Other comments trom the industry organizations and from the licensee resulted in both minor clarifying changes to the generic letter and a more practicalimplementation schedule. The comments did not result in a change or modification of the original Intent of the generic letter. Copies of the comment letters received are available in the Public Document Room (POR). A copy of the staff's evaluation of the cornments is available in the NRC Central Files and will be made available in the PDR when the generic letter is issued.
The Committee to Review Generic Requirements endorsed the generic letter without formal review. The draft Generic Letter was discussed with the Advisory Committee on Reactor Safeguards (ACRS) on April 15,1997. The ACRS decided not to review the final version.
The Office of the General Counsel reviewed the generic letter and has no legal objections.
The staff intends to issue this generic letter approximately 5 working days after the date of this information paper.
L. o'p Ex tive Director for Operations
Attachment:
Generic Letter," Degradation of Steam Generator Intemals" CONTACT: Stephanie M. Coffin, NRR 415-2778
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Comments on Proposed Generic Communication; Degradation of Steam Generator Inkrnalg. 61 Fed. Reg. 69.116 (December 31.1996)
A1TN: Chief, Rules Review and Directives 13 ranch On December 31,1996, the Nuclear Regu!atory Commission (NRC) issued the above captioned proposed generic communication for public comment. On February 6,1997, the comment period for the proposed generic communication was extended to hiarch 15,1997.F Provided below nre the comments of the Nuclear Utility Backfitting and Refbmi Group (NUDARG).F These comments concern the backlitting implications of the proposed generic communication.
'lhe proposed Generic Letter addresses the potential efTects of degradation of steam genemtor intemals on the structural integrity of steam gei,erator tubes. The regulatory requirements for steam generator tube inservice inspections are given in 10 C.F.R. f 50.55a(b)(2)(iii) which says that if the facility's Technical Specifications include surveillance requirements for steam generators differen. than those in Article IWB-2000,F 5t 7 inservice inspection program for steam generator tubing is govemed by the requirements in Technical Specifications. It is out understanding that F
Srs 62 Fed. Reg. 5,656 (Febmary 6,1997).
F' NUBARO is a consortium of 15 utilities fonned in the early 1980s, which participated actively in the development of the NRC's backlitting rule (10 C.F.R. {50.109) in 1985, and which has closely monitored the NRC's application of the rule since that time.
R Article IWB 2000 is contained in Section XI of the American Society of hicchanical Engineers Boiler and Pressure Vessel Cois which is incorporated by reference, editions
- through the 1989 Edition, in paragraph (b) of Section 50.55a.
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.w WINSTON & STI(AWN U. S. Nuclear Regulatory Commission March 14,1997 Page 2 most, if not all, plants do have specific Technical Specifications surveillance requirements governing htcam generator tube integrity.
The purposes of the proposed Generic Letter would be to (1) re-alert licensees to previously communicated findings of damage to steam generator internals, namely, tube suppon plates and tube bundle wrappers, at foreign pressurized water reactor facilities; (2) emphasize the importance of perfonning comprehensive examinations of steam generator internals to ensure steam generator tube structural integrity is maintained in accordance with the requirements of Appendix 11 to 10 C.F.R. Part 50; and (3) request infonnation from licensees that "will enable the NRC stafT to verify whether or not the condition of addressees' steam generator internals comply and confbnn with the current licensing basis for their respective facilities." The "Backfit Discussion" of the proposed Generic Letter states:
Under the provisions of Section 182a of the Atomic Energy Act of 1954, as amended, and 10 CFR 50.54(f), this generic letter transmits an infbnnation request for the purpose of verifying compliance with applicable existing regulatory requirements. Specifically, the requested infonnation will enable the NRC stafTto detennine whether or not the conditions of the addressees' steam generator intemals comply and confbnn with the current licensing basis fbr their respective facilities.
In particular, it would help ascertain whether or not the regulatory requirements nursuant to Appendix B to 10 CFR Part 50 are met, namely,(1) Criterion XI," Test Control," conceming the establishment of effective test programs for systems, structures and components, and (2) Criterion XVI, " Corrective Action," which 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.
Additionally, no backfit is either intended or approved in the context ofissuance of this generic letter. Thercibre, the stafThas not perfbnned a backlit analysis.
The proposed Generic Letter thus implies that licensees should be inspecting steam generator intemals to meet certain Appendix B criteria which relate to Quality Assurance prognun requirements for various activities. These criteria are general and provide process control requirements rather than specify the requirement for the process itself (sa, inspection). The specific requirements for inservice inspections of steam generators would be found in Section 50.55a or the plant Technical Specifications. The infomtation request, therefore, appears to put licensees in a position of having to justify the adequacy of existing inspection criteria in plant Technical Specifications, or other existing requirements or procedures, for steam generator tube integrity. 'lhe proposed Generic Letter appears to impose a requirement to perfonn certain inspections that may not be required under current regulations or plant Technical Specifications by citing Appendix B, Criterion XI," Test Control," as the basis for the compliance exception to the backfitting provisions of Section 50.109 'lhis is more than a mere "infbmiation request" and should be closely scrutinized.
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WINSTON & STHAWN i
i U. S. Nuclear Regulatory Commission l
March 14,1997 l
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In its "Backfitting Guidelines," the NRC Staff recognized that "[s]ome infonnation -
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requests promulgate new or revised stalTpositions and request that licensees, in their responses, state whether they will adopt the new positions. Even though these actions do not impose backfits, as a matter of internal stafrpractice they are identified as backfits and justified accoroingly before they i
are issued, as required by NRC procedures.... this is often the case with generic letters and bulletins."N Retefore, to the extent that the proposed Generic Letter would impose new Staff positions for the inspection of steam generator internals, a backfitting analysis would be required.
We recommend that the Staffnot issue the proposed Generic Letter until a backfitting analysis has been completed, justifying the need for the infbmiation and any new positions on I
inspection requiremen'.s. Licensees have already been informed of the NRC's coneems by In crmation Notices 95 09 and its Supplement 1, which described findings of damage to steam r
geterator intemals at farcign facilities. If the Staff believes that it has additional infonnation useful-to licensees, a Supplement 2 to infonnation Notice 96 09 could be issued rather than the proposed -
Generic Letter.
Sincerely, Y
Daniel F. Stenger Patricia L Campbell i
Counsel for Nuclear Utility llackfitting and Refonn Group i
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NUREG 1409,"Backfitting Guidelines," at 5 (July 1990).
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(C,,. taartTonto wstawG s JMCLE AR GE M RAT ON hfnreh 14,1997 hir. David L. hieyer, Chief Itules lleview and Directives Ilranch U.S. Nuclear Itegulatory Commission Washington, D.C. 20555 0001
SUBJECT:
Proposed Generic Communication - Degradation of Steam Generator Internals (61 Eed. Eng. 69110 - December 31,1996)
No1Lcr of Onno.rltwity for Public Comment
Dear hir. hieyer:
The Nuclear Energy Institute (NEI) submits these comments on behalf of the nuclear energy industry. These industry comments are in response to the December 31,1996, Federal Register notice concerning the proposed Generic Letter," Degradation of Steam Generator Internals."
The stated purposes of the proposed generic letter are to:
ite alert whiressees to previously communicated findings of e
damage to steam generator (SG) internals; Emphasize to addressees the importance of performing e
comprehensive examinations; and Itequest all addressees to submit information to NRC staff to e
verify whether or not the condition of addressees' steam generator internals comply and conform with the current licensing basis for their respective facilities.
I NEl in the organiration responsible for establishing untfied nuclear industry policy on matters affecting the nuclear energy industry. including regulatory aspects of generie operational and technical issues. NE!
members include all utihties licenned to cperate co.mercial nuclear power plants in the United States. nuclear plant designers. major architect /enoncenng firms, fuel fabncation facihties. materials licenseca. and other organiratmne and indmduals involved in the nuclear energy ndustry, w.,
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l'b Mr. David L. Meyer March 14,1997 Page 2 The first two purposes have already been adequately addressed. The NRC staff have communicated findings of damage to SG internals in French units, and the industry is well aware of the importance of comprehensive examinations. NRC Information Notice (IN) 90 09 and IN 90 09, Supplement 1, highlighted the past findings, their significance, and what steps licensnes should consider to preclude problems at their facilities.
Examinations performed during the normal course of a steam generator inspection outage are sufficient to detect significant degradation of SG tube support structures, (TSS). All utilities inspect a significant percentage of SG tubes during overy outage with a bobbin coil eddy current examination. If sections of TSS are missing (such as was discovered in some Electricit6 de France (EdF) plants), this condition in most instances would be obvious due to lack of an eddy current response at that TSS h> cation.
Many plants (representing each major SG design) also conduct secondary side visual examinations during perfortnance of tube pulls, chemical cleanings, loose-part retrieval efforts, and other maintenance activities. These inspections have not identified any significant degradation of SG internals. It can be reasonably assured that the extent of degradation reported in some EdF plants is not currently present in U.S. plants. Therefore, SG secondary sido degradation does not constitute an immediate threat to the safe operation of U.S. steam generators, and can be better resolved with a well planned, systematically executed approach.
The proposed generic letter request for information has the effect ofimposing actions for licensees to perform extensive secondary side in service inspections (ISI).
Inspections of this nature exceed those required by current regulations and plants technical specifications.
The industry does share the desire of NRC staff to gair. additional insight into the significance of the French experience relative to domestic facilities. Following the suggestion in the proposed generic letar to coordinate inspections, evaluations, and repair options, NEl has formed a task force to consider the need for an industry program. The intent of the program would be to evaluate the degradation found at the French and domestic plants, assess the susceptibility ofinstalled domestic SG designs to internals degradation, and provide recommendations for future inspections as may be appropriate.
An industry program, as described, will provide the NRC with information being sought by the proposed generic letter in a manner that would make more officient use of utility and NRC resources. Also, an industry program will provide guidance
4 Mr. David L. Meyer March 14,1997 Page 3 and direction for determining whether degradation of secondary side components pone a threat to tube integrity. Further, an industry program should negate the need to issue a generic letter regarding SG internals degradation.
If the NRC believes that a generic letter is necessary, then we offer the following comments to the proposed generic letter:
More time should be allocated for plant specific response to review the appropriate ISI data and 120 days is recommended; and--
- The NRC should delay issuing the proposed generic letter until after meeting with industry representatives to achieve a better understanding of an industry action plan.
We cppreciate tise opportunity to submit comments on the issue of degradation of steam generator internals. Please direct any questions on our comments to Clive Callaway at (202) 739 8114.
i Sincerely, ei htlhkhl$
luy David J. Modeon
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c:. Mr. Ous C. Lainas, NRC
- Mr. Stewart L. Magruder, Jr., NRC Mr. Edmund J. (Ted) Sullivan, Jr., NRC i
Ms. Stephanie M. Coffin, NRC -
Public Document Room (Project Number 689) i i
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Al l Foyetteville Street Mall Raleigh NC 27602 Serial: PE& RAS97-020 March 14,1997 Chief, Rules Review and Directives Branch U.S. Nuclear Regulatory Commission Mail Stop T-6D-69 Washington, DC 20555 0001
Subject:
Comments on Two Procosed NRC Generic Letters:
Degradation of Steam denerator Intemals (61 FR 69116), and [
Steam Generator Tube inspection Techniquer (61 FR 69118)
Dear Sir or Madam:
Carolina Power & Light Company (CP&L) offers the following specific comments on the proposed generic letter on degradation of steam generator internals:
The proposed generic letter requests the licensee to provide a "... [ discussion] of the program in place, if any, to detect degradation af steam generator internals and a description of the inspection plans, including the inspection scope, frequency, methods, equipmens and criteria, andplansfor corrective action in the event degradati(>n isfound." (emphasis added] While CP&L certainly has no specific objection to providing a general description ofits inspection program, we question whether it is possible to specify a plan for corrective action prior to assessing any damage found during these inspections. CP&L recommends that this aspect of the request be clarified.
CP&L offers the following specific comments on the proposed generic letter on steam generator tube inspection techniques:
The requested brormation, as expressed in the proposed generic letter, is unclear. For example, the tee of the word " defect" used in the Requested Information section of the proposed generic letter is inconsistent with *he definitions in the Technical Specifications at CP&L plants. CP&L's Technical Specifications for the Robinson Nuclear Plant and the Harris Nuclear Plant define " defect" as "... an imperfection of such severity that it exceeds the plugging limit." The word " defect" should be
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Chief, Rules Review and Directives Branch 2
March 14,1997 replaced with the word " indication" as used in EPRI Guideline NP-6201, "PWR Steam Generator Tube Examination Guidelines," Rev,4.
The proposed generic le:ter requests licensees to inform the NRC:
(1) Whether it is their practice to leave steam generato: tubes with defects in service, based on sizing, and (2) if the response to item (1) is affirmative, those licensees are requested to submit a wraten report that includes, for each 2
type of steam generator degradation mechanism, e &cription of1.e associated nondestructive examination method t =i ; ted and the technical basis for the acceptability of the technique used."
CP& L recommends for clarity that " degradation mechanism" be replaced with " type ofindication," e.g., circumferential cracking, axial cracking, large volumdric indication (wastage), small volumetric indicatio,. (pitting), denting, primary versus secondary side indication.
If you have any questions regarding these comments, please contact me at (919) 546-6901, or Mike Murdock at (919) 546-3193.
Sincerely, tb T.D. Walt Manager, Performance Evaluation
& Regulatory Affairs MLM/
cc:
Mr. L A. Reyes, Regional Administrator - Region 11 Mr. J. B. Brady, USNRC Resident inspector - IINP, Unit 1 Mr. B. B. Desai, USNRC Resident inspector - IIBRSEP, Unit 2 Mr. N. B. L.e, USNRC Project Manager - IINP, Unit 1 Ms. B. L Mozafari. USNRC Project Manager - IIBRSEP, Unit 2 r
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k' Manelma, Utility Emel<fittina and Rafntm Grann fMllRAstG1
- The proposed generic letter imposes a new staff requirement to perform inspectiot. of the steam generator internals and thus a backfitting analysis to justify the proposed generic
-letter is needed.
Inspections of the steam generator internals at foreign and U.S. PWR facilities lett to the
- identification of tube support plate and bundle wrapper degradation that may affect the integrity of the steam generator tubes, The damage mechanisms responsible for the degradation do not appear to be unique to the facilities; thus they may be active at other U.S. PWR facilities. As stated in the proposed generic letter, detection of steam generator internals degradation through inspections will help to ensure the steam generator tube structuralIntegrity is maintained in accordance with the requirements of Appendices A and 8 to.10 CFR Part 50. These criteria have always existed; the generic letter does not impose new or revised staff positions or requirements. The staff also notes the proposed generic letter does not require addressees to perform inspections of the steam generator internals.
The proposed generic letter requests information regarding past inspection practices, plans for future inspections and/or justification as to why inspections are not warranted given the foreign and U.S. experience with steam generator internal degradation.
Nucimar EnergyJontituta (NED (1)
The generic letter has the effect of imposing actions for addressees to perform inspection of steam generator intemals.
(2)
An industry program to address th y issues raised by the proposed generic letter negates the need for the proposed generic letter..
(3)
The response time for the requested informatten should be increased from 60 days to 120 days.
. (4)
Delay issuing the proposed generic letter until after meeting with industry representatives to achieve a better understanding of the industry action plan.
(1)
The staff response to the first comment from NElis identical to the response to
.NUBARG discussed above.
(2)
- The proposed generic letter is an appropriate vehicle for documenting the staff's position concerning the foreign and U.S. experience with degradation of steam generator internals and for documenting U.S. industry response to this issue.
(3)
. Since no actions on the part of the addressess are requested, the staff considered 60 days to be adequate. However, the staff understands there will be an industry plan developed in response to the generic letter and thus an increase in the response time to 90 days is reasonable to allow more time for coordination of industry offorts.
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' The staff is willing to meet with industry representatives to discuss industry initiatives; however, the industry action plan would not change the purpose or content of the proposed generic letter and thus the staff will not delay issuing the letter based on such a meeting taking place.
Caratina Pawar & Ilant camaanyJCP&t )
CP&l's comments concerned the 'RequestedInformation'section where the staff requested addressees discuss their program in place, if any, to detect degradation of steam generator internals, including plans for corrective action in the event degradation is found.
CP&L questioned how one could specify a plan for corrective action prior to assessing any damage found.
The staff agrees the wording' of the " Requested Information" section was misleading. This section of the proposed generic letter was modified.
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CBGR REVIEW PACKAGE PROPOSED ACTION:
Issue a generic letter to (1) again alert addressees to the previously communicated findings of damage to steam generator internals, namely, tube support plates and tube bundle wrappers, at foreign PWR facilities; (2) alcrt add *essees to recent findings of damage to steam generator tube support plates at a U.S. PWR facility; (3) emphasize to aodressees the importance of performing comprehensive examinations of steam generator internals to ensure steam generator tube structural integrity is maintained in accordance with the requirements of Appendices A and B to 10 CFR Part 50; and (4) require all addressees to submit information that will enable the NRC staff to verify whether or not addressees' steam generator internals comply with and conform to the current licensing basis for their respective facilities.
CATEGORY:
2 RESEQNSE TO RFOUIREMENTS FOR COhlTFhrt OF PACKAGF RURMITTED FOR CRGR RFVIEW (i)
The proposed generic requirement or staff position as it is proposed to be sent out to licensees.
Within 90 days of the date of this generic letter, each addressee is required to provide a written report that includes the following information for its facility:
(1)
Discussion of any program in place to detect degradation of steam generator internals and a description of the inspection plans, including the inspection scope, frequency, methods, and equipment.
The discussion should include the following information:
(a)
Whether inspection records at the facility have been reviewed for indications of tube support plate signal anomalies from eddy current testing of the steam generator tubes that may be indicative of support plate damage or ligament cracking. If the addressee has performed such a review, include a discussion of the findings.
(b)
Whether visual or video camera inspections on the secondary side of the steam generators have been performed at the facility to provide information on the condition of steam generator internals (e.g.,
support plates, tube bundle wrappers, or other components). If the addressee has performed such inspections, include a discussion of the findings.
2 (c)
Whether degradation of steam generator internals has been detected at the facility, and how the degradation was assessed and dispositioned.
(2)
If the addressee currently has no program in place to detect degradation of steam generator internals, include a discussion and justification of the plans and schedule for establishing such a program.
(3)
If the addressee has no program in place and no plans for establishing a program to detect degradation of steam generator internals, include a justification of why no program is needed.
(ii)
Draft staff papers or other undertying staff documents supporting the requirements or staff positions. (A copy of all materials referenced in the document shall be made available upon request to the CRGR staff. Any Committee member may request CRGR staff to obtain a copy of any reference rnatorial for his or her use.)
1.
NRC Information Notice 96-09, " Damage in Foreign Steam Generato Internals," February 12,1996 2.
NRC Information Notice 96-09, Supplement 1, " Damage in Foreign Steam Generator Intemals," July 10,1996 (iii)
Each proposed requirement or staff position shall contain the sponsoring office's position as to whether the proposal would increase requirements or staff positions, implement existing requirements or staff positions, or would relax or reduce existing requirements or staff positions.
The generic letter will permit verification that addressees are in compliance with existing regulations.
(iv)
The proposed method of implementation with the concurrence (and any comments) of OGC on the method proposed. The concurrence of affected program offices or an explanation of any nonconcurrences.
The method of implementation will be the generic letter (Attachment 1). OGC has no legal objections to the generic letter.
(v)
Regulatory analyses conforming to the directives and guidance of NUREG/BR 0058 and NUREGICR 3568. (This does not apply for backfits that ensure compliance or ensure, define, or redefine adequate protection. in these cases a documented evaluation is required as discussed in IV.8.(ix).)
The generic letter requests information be submitted so the NRC can verify addressees are in compliance with existing regulatory requirements; therefore, no value/ impact analysis was performed.
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3 (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 reactors, all PWRs only, some vendor types, some vintage types such as BWR 6 and 4, jet pump and nonjet pump plants, etc.).
The generic letter applies to all holders of operating licenses for pressurized water reactors (PWRs), except those who have permanently ceased operations and have certified that fuel has been permanently removed from the reactor vessel.
(vill For backftts other than compliance or adequate protection backfits, a backfit analysis as defined in 10 CFR 50.109. The backfit analysis shall include, fer each category of reactor plants, an evaluation which demonstrates how the action should be prioritized and scheduled in light of other ongoing regulatory activities. The backfit analysis shall document for consideration information available concoming any of the following factors as may be appropriate and any other information relevant and material to the proposed action:
This item is not applicable to the generic letter.
(viii)
For each backfit analyzed pursuant to 10 CFR 50.109(a)(2) (i.e., not adequate protection backfits and not compliance backfits), the proposing Office Director's determination, together with the rational for the determination based on the consideration of paragraph (i) and (vii) above, that:
(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 costs of implementation, for the facilities affected, are justified in view of this increased protection.
This item is not applicable to the generic letter.
(ix)
For adequate protection or compliance backfits evaluated pursuant to 10 CFR 50.10Ma)(4)
(a) a documented evaluation consisting of:
(1) the objectives of the modification (2) the reasons fu the modification (3)-
the basis for invoking the compliance or adequate protection exemption.
(b) in addition, for actions that were immediately effective (and therefore issued without prior CRGR review as discussed in lil.C) the evaluation shall document the safety significance and appropriateness of the action taken and
-(if applicable) consideration of how costs entributed to selecting the solution among various acceptable altamatives.
This item is not applicable to the generic letter, i
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-4 (x)
For each evaluation conducted for proposed relaxations or decreae" in current requirements or staff positions, the proposing Office Director's deter.
i, together with the rationale for the determination based on the considerations or paragraphs (i) through (vil) above, that:
(a)
The public hashh and safety and the common defense and security would be adequately protected if the proposed reduction in requirements or positions were implemented, and (b)
The cost savings attributed to the action wow!d be substantial enough to Justify taking the action.
This item is not applicable to the generic letter since there is no relaxation or decrease in the current requirements.
(xi)
For each request for information under 10 CFR 50.54(f) (which is not schlect to aception as discussed in Ill.Al an evaluation that includes at least the following elements:
(a)
A problem statement that describes the need for the information in terms of potential safety benefit.
(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.
(d)
A statement affirming that the request does not impote new requirements on the licensee, other than for the requested information.
This item is not applicable under 10 CFR 50.54(f) berause the requested information contained in this generic letter is to verify compliance with existing requirements.
(xii)
An assessment of how the proposed action relates to the Commission's Safety Goal Policy Statement.
There is no impact on the Commission's Safety ( n Policy Statement since the requested information is considered necessary to verify compliance with existing regulations.
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