ML20132E340
| ML20132E340 | |
| Person / Time | |
|---|---|
| Issue date: | 12/19/1996 |
| From: | Moulton J NRC (Affiliation Not Assigned) |
| To: | Firth D FRAMATOME |
| References | |
| PROJECT-683 NUDOCS 9612230231 | |
| Download: ML20132E340 (6) | |
Text
-
~.
December 19, 1996 Mr. David J. Firth Program Director Generic License Renewal; Program Framatome, Technologies, Inc.
i P.O. Bbxl10935 Lynchburg;jVirgini'aJ 24506-0535,,
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION IN THE CASE OF BAW-2251,
' " DEMONSTRATION OF. AGING EFFECTS FOR THE REACTOR VESSEL (RAI NOS. 8
, THROUGH 18),
Dear Mr. Firth:
The attached request for additional information (RAI) pertaining to the NRC staff's review of BAW-2251, " Demonstration of Aging Effects for the Reactor Vessel." This request contains RAls number 8 through 17.
RAIs I through 7 were forwarded in a letter to Mr. Don Cronenberger dated August 28, 1996.
Your prompt response to this RAI will ensure that these aspects of the staff's review can be completed.
If you have any questions concerning the attached RAI, please contact me at 415-1106.
i Sincerely, John P. Moulton, Project Manager original signed by License Renewal Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation Project No. 683
Attachment:
RAIs cc: See attached list DISTRIBUTION w/ attachments:
o Central File LLois y[YD J I I
PUBLIC JMoulton PDLR R/F PROT DISTRIBUTION w/ attachments via e-mail:
6 $3 JBirmingham (JLB4)
GBagchi (GXB1)
RJohnson (REJ)
LShao (LCXI)
EJordan (ELJ1)
HBrammer (HLB)
MCase (MJC)
Llois (LXL1)
RJones (RCJ)
JDavis (JAD)
TMartin (TTM)
.MBanic (MJB)
JFair (JRF)
BElliot (BJE)
JStrosnider (JRS2)
PDLR Staff DOCUMENT NAME: RAlt.JPN 0FFICE LA:PDLg PM:PDLR
{
SC:PDLR
(
(A)DE:Ait l
PD:PDLg V l
NAME
< "hL JMoulttinM 4 PTKuo 2 F Glainas U m g SNewbtrry l
3 DATE 11/ 7196 11A?J96 i 12f/f/96 1)/l(d96' 71/ W 96 l
tos0074 0FFICIAL RECORD COPY (V
@i/
yne TE CHiiB gp"
- n22n8,2=="
g.
683 PDR
p ub p*
~4 UNITED STATES
}
NUCLEAR REGULATORY COMMISSION o
t WASHINGTON, D.C. 2056M001
% *****/
December 19, 1996 Mr. David J. Firth Program Director Generic License Renewal Program Framatome Technologies, Inc.
P.O. Box 10935 l
Lynchburg, Virginia 24506-0935
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION IN THE CASE OF BAW-2251,
" DEMONSTRATION OF AGING EFFECTS FOR THE REACTOR VESSEL (RAI NOS. 8 THROUGH 17)
Dear Mr. Firth:
The attached request for additional information (RAI) pertaining to the NRC staff's review of BAW-2251, " Demonstration of Aging Effects for the Reactor Vessel." This request contains RAls number 8 through 17.
RAls I through 7 were forwarded in a letter to Mr. Don Cronenberger dated August 28, 1996.
Your prompt response to this RAI will ensure that these aspects of the staff's review can be completed.
If you have any questions concerning the attached RAI, please contact me at 415-1106.
Sincerely,
/,
?
d y' Jo)m P. Moul on, Project Manager ycense Renewal Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation Project No. 683
Attachment:
RAls cc:
See attached list l
i J
i
BAW-2251. Reauest for Additional Information Nos. 8 throuah 17 8.
Section 4.5.1 of the report addresses the management of fatigue. The report indicates that each of the participating utilities monitor occurrences of design transients and are thereby managing the potential for cracking resulting from fatigue. Describe how this monitoring is being performed.
Include a discussion of the instrumentation, if any, used to compare design transients with actual plant transients. Also provide documents which are the basis for your statement:
"It has been demonstrated that the existing fatigue usage factors... remain valid for the period of extended operation.."
9.
The report references-the staff recommendations for environmental assisted fatigue provided in SECY 95-245. The report interprets the staff recommendation to be an assessment of environmentally assisted fatigue is required only if the fatigue usage factor [without consider-ing environmental effects] exceeds 1.0 during the extended period of operation. This interpretation is incorrect. The recommendation in SECY 95-245 is that those components, evaluated as part of the staff's Fatigue Action Plan. sample, that had high usage factors [ calculated with fatigue curves that included environmental effects] be evaluated further for any extended period of operation.
Provide an additional discussion i
regarding environmentally assisted fatigue based on the above clarification of the staff recommendation.
10.
Identify the capsules.and " host" reactors in the Integrated B&WOG
- Surveillance program that will be used to monitor the effect of neutron radiation on the beltline reactor vessel materials for the participating plants in the B&WOG Generic License Renewal Program over the license renewal term.
Identify the materials in the capsules and the neutron fluence to be received by the capsules. Compare the neutron fluence of the capsules to the values specified in ASTM E-185-82 and explain the basis for the conclusion that the proposed. integrated surveillance program complies with the requirements of Appendix H to 10 CFR.Part 50 for the license renewal term.
1 11.
Provide B&WOG Integrated Surveillance program test results from welds fabricated using the same heats of weld wire as were used to fabricate the beltlines of the reactor pressure vessels at the participating plants in order to demonstrate that the embrittlement estimates in Appendix A, " Pressurized Thermal Shock," are conservative. Compare the 4
chemistry factors from surveillance data that are calculated using the methodology in section 2.1 of Regulatory Guide (RG) 1.99, Rev. 2 to the values used in Appendix A.
Provide a determination of whether or not the surveillance data meet the credibility criteria in RG 1.99, Rev. 2.
Identify the chemical composition (percentage amounts of copper and nickel) of the surveillance weld and the beltline weld materials.
If the copper and nickel content of the surveillance weld and beltline welds are different, adjust the surveillance data in accordance with the procedures in section 2.1 of RG 1.99, Rev. 2.
ATTACHMENT
12.
With respect to Appendix A, for " Pressurized Thermal Shock," in Topical Report BAW-2251, for all reactor pressure vessel (RPV) materials listed in Tables A A-5, provide the basis and references for the revised neutron fluence values at the inside RPV surfaces.
13.
In BAW-2251, the B&WOG proposed to manage age related cracking in partial penetration welded Alloy 600 vessel head penetration nozzles by performing ASME defined VT-2 visual examinations of 25% of the nozzles
(> 2 NPS).in the head. The B&WOG stated that the VT-2 examinations should be capable of detecting cracking the partial penetration nozzles.
The B&WOG has also proposed to use leakage detection and surveillance of-boric acid corrosion as acceptable methods for managing cracking of Alloy 600 components. However, the NRC issued a draft Generic Letter (GL) entitled " Primary Water Stress Corrosion Cracking of Control Rod Drive Mechanism and Other Vessel Head Penetration" in which the staff concludes that an integrated, long-term program, which includes periodic inspections and monitoring, is necessary to provide sufficient assurance of the structural integrity of these penetrations. The B&WOG should consider the information provided in the draft generic letter and will be required to provide an appropriate aging management program for managing PWSCC of Alloy 600 vessel head pernetrations, including <2" NPS penetrations.
14.
IGSCC of closure stud assembly made of 4340 steel.
According to the Summary of Technical Information and Agreements from NUMARC Industry Reports Addressing License Renewal (O. CHopra, D. Ma, and W. Shack, Argonne National Laboratory, November 1994), NUMARC/NRC agreements were that ASME XI examinations performed in accordance with Regulatory Guide (RG; 1.65, " Materials and Inspection for Reactor Vessel closure Studs," were acceptable management methods for managing the IGSCC of closure stud assemblies.
The B&WOG does not state in its i
repart that ASME inspections will be done using the guidance in RG 1.65.
Revise your topical report to commit to performing the inspections using the guidance in RG 1.65 or provide justification for not performing them.
15.
Page 4-5 of the report indicates that surveillance capsule testing will'-
be completed by year 2008.
Discuss what capsules, if any, will remain in the reactors after 2008 and their withdrawal schedule.
ATTACHMENT
16.
Page 4-10 of the report indicates loss of material on the mating surfaces between the closure flanges is managed by ASME Section XI Examination Category B-N-1.
It then discusses that B-N-1 requires a full visual examination of all accessible areas of the interior of the reactor vessel.
Clarify whether B-N-1, that is, examination of all accessible areas, or just a small portion of B-N-1, that is, examination of only the flange areas, is an inspection necessary for renewal.
17.
Page 4-10 of the report indicates that the fatigue usage factor for the Oconee Nuclear Station Units 1, 2, and 3 reactor vessel studs is l
projected to exceed 1.0 by the end of the renewal term and will be evaluated separately for renewal on a plant specific basis. However, because B&W plants are generally similar in design and operation, discuss why the studs at the other B&WOG GLRP plants are not similarly affected.
i i
ATTACHMENT
Project No. 683 Babcock & Wilcox Owners Group Gener-ic License Renewal Program cc:
Mr. Robert B. Borsum Regional Administrator, Region IV Framatome Technologies U.S. Nuclear Regulatory Commission 1700 Rockville Pike 611 Ryan Plaza Drive, Suite 1000 Suite 525 Arlington, Texas 76011 Rockville, Maryland 20852 Mr. James J. Fisicaro Michael Laggart Director, Licensing Manager, Corporate Licensing Entergy Operations, Inc.
GPU Nuclear Corporation Route 3, Box 137G One Upper Pond Road Russelville, Arkansas 72801 Parsippany, New Jersey 07054 Earnest L. Blaxe, Jr., Esq.
Chairman Shaw, Pittman, Potts Board of County Commissioners and Trowbridge of Dauphin County 2300 N. Street, NW Daughin County Courthouse Washington, D.C.
20037 Harrisburg, Pennsylvania 17120 Regional Administrator, Region I Mr. J. W. Hampton U.S. Nuclear Regulatory Commission Nuclear Generation Vice President 475 Allendale Road Duke Power Company King of Prussia, Pennsylvania 19406 Oconee Nuclear Station MC: ONO IVP B. Gutherman, Manager P.O. Box 1439 Licensing Seneca, South Carolina 29679 Florida Power Corporation (SA2A)
Crystal River Energy Complex Mr. John R. McGaha 15760 W. Powerline Street Vice President, Operations Support Crystal River, FL 34428-6708 Entergy Operations, Inc.
P.O. Box 31995 William Dornsife, Acting Director Jacksonville, Mississippi 39286 Bureau of Radiation Protection Pennsylvania Department of Regional Administrator, Region II Environmental Resources U.S. Nuclear Regulatory Commission P.O. Box 2063 101 Marietta St., N.W. Suite 2900 Harrisburg, Pennsylvania 17120 Atlanta, Georgia 30323 Chairman Mr. R. L. Gill Board of Supervisors GLRP Licensing Coordinator of Londonderry Township c/o Duke Power Company R.D. #1 Geyers Church Road EC-12R Middletown, Pennsylvania 17057 P.O. Box 1006 Charlotte, North Carolina 28201-Mr. J. E. Burchfield 1006 Compliance Duke Power Company Oconee Nuclear Site P.O. Box 1439 Seneca, South Carolina 29679 i