ML20196G231
| ML20196G231 | |
| Person / Time | |
|---|---|
| Issue date: | 12/02/1998 |
| From: | Anand R NRC (Affiliation Not Assigned) |
| To: | Firth D BABCOCK & WILCOX OPERATING PLANTS OWNERS GROUP |
| References | |
| PROJECT-683 NUDOCS 9812070288 | |
| Download: ML20196G231 (8) | |
Text
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December 2, 1998 Mr. David J. Firth Progrcm Dircctor Generic License Renewal Program j
The B&W Owners Group 1700 Rockville Pike, Suite 525 i
Rockville, MD 20852
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE BABCOCK &
WILCOX OWNER'S GROUP GENERIC LICENSE RENEWAL PROGRAM TOPICAL REPORT ENTITLED," DEMONSTRATION OF THE MANAGEMENT OF AGING EFFECTS FOR THE REACTOR VESSEL INTERNALS," BAW-2248,
]
JULY 1997
Dear Mr. Firth:
By letter dated July 29,1997, the Babcock & Wilcox Owners Group (BWOG) submitted Topical Report BAW-2248, Demonstration of the Management of Aging Effects for the Reactor Vessel intemais," July 1997, requesting the U.S. Nuclear Regulatory Commission staff's review and issuance of a safety evaluation report.
i Based on the review of the information submitted, the staff has identified, in the enclosure, i
areas where additional information is necded to complete the review, Please provide a schedule for the submittal of your response within 30 days of the receipt of this letter. Additionally, the staff is willing to meet with the BWOG before you submit your i
response to clarify the staff's request for additional information.
Sincerely, l
l WWW&
i Raj K. Anand, Project Manager License Renewal Project Directorate Division of Reactor Program Management l
Office of Nuclear Reactor Regulation Project No. 683
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Enclosure:
As stated i
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Project No. 683 Bibcock & Wilcox Owners Group G:neric License R:n:wal Program Mr. Rick Edwards
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Framatone Technologies Mr. James J. Fisicaro 1700 Rockville Pike Director, Licensing Suite 525 Entergy Operations, Inc.
Rockville, Maryland 20852 Route 3, Box 137G Russelvitie, Arkansas 72801 Mr. Michael Laggart Manager, Corporate Licensing Eamest L. Blake, Jr., Esq GPU Nuclear Corporation Shaw, Pitman, Potts One Upper Pond Road and Trowbridge Parsippany, New Jersey 07054 2300 N. Street, N.W.
WatUngton, DC 20037 Chairman Board of County Commissioners M. S. Tuckman j
of Dauphin County Duke Energy Corporation Dauphin County Courthouse Mail Stop EC-07H Harrisburg, Pennsylvania 17120 P.O. Box 1006 Charlotte, North Carolina 28201-1006 Mr. J. W. Hampton Nuclear Generation Vice President B. Gutherman, Manager Duke Energy Corporation Florida Power Corporation, (GA2A)
Oconee Nuclear Station Crystal River Energy Complex MC: ONO 1VP 15760 W.. Power Street P. O. Box 1439 Crystal River, Florida 34428-6708 Seneca, South Carolina 29670 Mr. William Domsife, Acting Director Mr. John R. McGaha, Vice President Bureau of Radiation Protection Operations Support Licensing Pennsylvania Department of Entergy Operations, Inc.
Environmental Resources P. O. Box 31995 P. O. Box 2063 Jacksonville, Mississippi 39286 Harrisburg, Pennsylvania 17120 William R. McCollum, Jr Chairman t
Duke Energy Corporation Board of Supervisors P.O. Box 1439 of Londonderry Township Seneca, SC 29679 R.D. #1 Geyes Church Road Middletown, Pennsylvania 17057 Mr. R. L. Gill GLRP Licensing Coordinator Mr. J. E. Burchfield C/O Duke Energy Corporation Compliance Manager EC-12R Duke Energy Corporation l
P. O. Box 1006 Oconee Nuclear Site Charlotte, North Carolina 28201-1006 P. O. Box 1439 Seneca, South Carolina 29679 Gregory D. Robison Duke Energy Corporation Douglas J. Walters Mail Stop EC-12R Nuclear Energy Institute P.O. Box 1006 17761 Street, NW Charlotte, North Carolina 28201-1006 Suite 400 Washington, DC 20006-3708 DJW@NEl.ORG i
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. PUBLIC PDLR R/F RAnand M.El-Zeftawy ACRS T2E26 E-Mati R. Zimmerman I
M. Federline l
-T. Essig l
H. Brammer l
G. Holahan
- E. Sullivan B. Boger L
D. Martin T. Martin K. Manoly B. Elliot B. Sheron W. McDowell A. Murphy H. Conrad R. Weisman l
J. Roe l
G. Lainas l
D. LaBarge l
S. Newberry l
F. Grubelich
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D. Matthews J. Strosnider C. Ogle, Ril C. Gratton J. Fair C. Grimes l.
G. Bagchi l
R. Trojanowski(Rll State Liaison)
L. Spessard K. Wichman i
R. Correia R. Latta J. Moore M. Zobier J. Craig E. Hackett PDLR Staff r[v3 Mo
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i REQUEST FOR ADDITIONAL INFORMATION THE BABCOCK & WILCOX OWNER'S GROUP GENERIC LICENSE RENEWAL PROGRAM TOPICAL REPORT ENTITLED, " DEMONSTRATION OF THE i
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MANAGEMENT OF AGING EFFECTS FOR THE REACTOR VESSEL INTERNALS,"
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BAW-2248, JULY 1997 1)
Section 1.4 of BAW-2248 identifies the functions of the reactor vesselintemals (RVI). It does not include, " provide shielding for the RPV [ reactor pressure vessel)." Do the intended functions of the reactor vessel intemals " provide shielding for the reactor pressure vessel"?
2)
BAW-2248 addresses certain applicable aging effects for specific reactor vessel intemals j
components. Describe, in summary form, the bases for concluding that the following aging effects were not significant for the specific components: stress corrosion cracking (SCC), and irradiation-assisted stress corrosion cracking (IASCC) of the plenum cover and plenum cylinder; SCC, IASCC, wear, and thermal embrittlement of the rod control l
assembly (CRA) guide tubes; SCC and IASCC of the CRA guide tube bolts; SCC of the upper grid assembly; SCC and IASCC of the upper grid rib section, upper grid assembly i
bolts, and the upper intemals fuel guide pads; SCC, lASCC, and neutron irradiation embrittlement of the core support shield and the core support shield flange; IASCC and neutron irradiation embrittlement of the vent valve assemblies; SCC of the core barrel l
assembly; SCC, IASCC, creep, and neutron irradiation embrittlement of the baffle and former plates; SCC, and stress relaxation of the baffle-former bolts; SCC of the lower grid top rib section, the lower grid bottom rib weldment, and the lower grid assembly support posts; SCC, lASCC, and neutron irradiation embrittlement of the lower intemals fuel guide pads, the lower grid assembly bolts, and the lower grid assembly guide blocks.
3)
Section 54.21(a)(3) of 10 CFR states that the integrated plant assessment must
" demonstrate that the effects of aging will be adequately managed...for the period of l
extended operation." Section 4.6 of BAW-2248 describes a proposed Reactor Vessel i
intemals Aging Management Program (RVIAMP) that is intended to meet 10 CFR 54.21(a)(3), for the affected part and aging effect combinations listed in Table 4-1 of BAW-2248. For those items listed in Table 4-1 susceptible to a reduction of fracture toughness aging mechanism, submit a fracture mechanics analysis to determine the critical flaw size during normal operation and during emergency and faulted conditions.
Identify the inspection procedure and the capability of the inspection to detect flaws smaller in size than that of the critical flaw.
The fracture toughness of austenitic stainless steel can become degraded with high levels of neutron irradiation; for example, fluences greater than 1 x 102 n/cm (E > 1MeV).
2 Fracture toughness data for irradiated stainless steels at such high fluences are not plentiful. Two sources available in the public litereture are References 1 and 2.
The data in Reference 1 are for the initiation fracture toughness (i.e., at the initiation of crack growth), defined by:
\\
Enclosure
2 Kg=}J xE u
J. is defined as the J-integral value at the initiation of crack growth and E is the Young's modulus for the material. For Type 304 stainless steel plate irradiated to a fluence of
~5 x 10' n/cm (E > 1MeV) at ~280 *C and tested at 288 *C, the lowest reported value r
in Reference 1 of J. (~75 in.-lb/in.2) corresponds to a K, of ~50 ksi/in.
From Reference 2, J-integral resistance or J-R curve data are reported for two samples fabricated from core shroud material removed from an overseas boiling water reactor (BWR) (see figure attached). The fluence for these samples is reported in Reference 2 as 8 x 102" n/cm'.
Reconciliation of the J, from Reference 1 with the J-R curve trends from Reference 2 (through scaling of the J levels in the J-R curves) can provide one estimate of the fracture toughness of highly irradiated austenitic stainless steel.
Provide any other fracture toughness data used in this evaluation.
4)
Aging effects of many reactor vesselintemal components will be managed by the Reactor VesselIntemals Aging Management Program. The program elements are discussed in Section 4.6 of BAW-2248. Provide a plan and schedule for completing all the elements of the program.
5)
Table 4-1 of BAW-2248 indicates that management of reduction of fracture toughness in vent valve bodies and vent valve retaining rings will be accomplished principally by the American Society of Mechanical Engineers (ASME) Section Inservice inspection (ISI)
Program. This is in contrast to the treatment of other RVI components subject to loss of fracture toughness, for which the RVIAMP is set forth to manage the aging effects. Why are the vent valve components treated differently, and should they be included in the scope of the RVIAMP?
6)
Table 4-1 of BAW-2248 indicates that the vent valve retaining ring, vent valve bodies, and the locking devices on the modified vent valve assembly do not require a supplemental aging management program. Aging effects will be managed during the renewal term using ASME Boiling and Pressure Vessel Code (Code) inspection methods. Since functions of these components are affected by either a reduction of fracture toughness or stress corrosion cracking, will ASME Code VT-3 visual examination be adequate for discovering cracks that could lead to failure of the component? What examination methodology is required? Are the surfaces of the components accessible for detecting cracks that could lead to failure of the components?
7)
Page 3-5 of BAW-2248 states that "the ONS-1 CRGT assembly sectors required straightening after the first hot functional test (FHT)." Provide a summary of the evaluation indicating that the cracking mechanisms (SCC and IASCC) are not plausible in this case. If these guide tube sectors were to be degraded by a cracking mechanism, could such cracking impede the ability of the reactor vessel intemals to perform its function to " provide support, orientation, guidance, and protection of the control rod assemblies"?
a
3 8)
Examination Category B-N-3 of ASME Section XI requires a VT-3 visual examination of
" accessible surfaces" of removable core support structures. For assemblies and parts l
determined to be susceptible to no aging mechanisms, and hence to require no additional j
aging management, the VT-3 examination provides one measure of assurance of the i
structuralintegrity of the part. Which components not susceptible to an aging mechanism i
(and hence, no additional aging management) will receive a VT-3 examination that can j
serve as a sampling of nonsusceptible components?
9)
Section 3.3 of BAW-2248 indicates that crevice corrosion is not expected to be a concem, unless the intemals are exposed to a series of long outages that have stagnation and high impurity levels. What impurity levels and how much cumulative outage time are requiied before crevice corrosion becomes a concem? What components could be affected by crevice corrosion? How could crevice corrosion be prevented if there were a long outage?
- 10) Section 3.3 of BAW-2248 indicates that wear is not a concem for the modified vent valve locking devices. Explain the difference between the original design and the modified design that eliminated the concem for wear of the vent valve locking device. To what criteria are the modified vent valve locking devices being inspected to ensure that they are j
l not subject to wear? Summarize the results of these inspections; include the number and frequency of inspection. Will these inspections be continued into the license renewal term?
l
- 11) Section 3.3 of BAW-2248 indicates that Westinghouse has observed wear of incore guide tubes caused by flow induced vibration in regions directly exposed to reactor coolant system (RCS) flow. Wear of Babcox & Wilcox (B&W)-designed guide tube and spiders are not a concem; however, because of differences between the Westinghouse and B&W guide tube design and because the B&W-designed detectors are inserted and withdrawn once per fuel cycle. Explain the difference in design and operation of the Westinghouse and B&W incore guide tubes that indicates wear is not a concem for the B&W design. Are there any limits on the number of insertions and withdrawals of the incore monitors that could lead to a concem about wear of the guide tubes?
- 12) During its interaction meetings with the staff (referenced in the Section 4.3.11 of the Oconee Nuclear Station License Renewal-Technical Information New Programs and Activities Report, June 1998), the B&W Owners Group (B&WOG) described current and ongoing reactor intemals baffle bolt activities that included preparation for possible augmented baffle bolt inspection during the next 10-year ISI interval at Oconee 1 (2003 at the earliest). Describe baffle bolt inspections that will be conducted prior to the start of the extended license renewal period and indicate how these actions provide the basis for assuring the baffle bolt monitoring and inspection techniques that are planned during the period of extended operation are appropriate.
- 13) Describe the program that will be implemented as outlined in Section 4.6 of BAW-2248 with regard to the aging management of the reactor intemals baffle bolts. Describe the '
overall inspection program, including aspects such as, intervals, monitoring, and inspection techniques, i
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- 14) Describe the replacement bolts and redesigned RVI that are referred to in the fatigue analysis discussed in BAW-2248 Section 4.5.1. Are the replacernent bolts and redesigned RVI identified in the fatigue analysis related to the issue of the A-286 bolt cracking discovered in B&W RVl? Are the baffle bolts discussed in the BAW-2248 report included in the fatigue analysis? If not, what is the basis for not including the baffle bolts in the fatigue analysis? If the baffle bolts are included in the analysis, describe how baffle bolt cracking is taken into account and identify the analysis report.
REFERENCES
- 1. Loss, F. J., and Gray, Jr., R. A., "J-Integral Characterization of Irradiated Stainless Steels,' NRL Report 7565, Naval Research Laboratory, Washington, D. C., April 25, 1973.
- 2. EPRI TR-107079, "BWR Vessel and Internals Project, BWR Core Shroud Inspection and Flaw Evaluation Guideline, Revision 2 (BWRVIP-01)," October 1996, pp. 4-13.
Attachment:
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Figure J-R curve data for irradiated stainless steel for a fluence of 8 x 1020 2
n/cm (Ref. 2).
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