ML060550419
| ML060550419 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 03/20/2006 |
| From: | Ashley D NRC/NRR/ADRO/DLR/RLRA |
| To: | Swenson C AmerGen Energy Co |
| Ashley D, NRR/DLR/RLRA, 415-3191 | |
| References | |
| %dam200604, TAC MC7624 | |
| Download: ML060550419 (13) | |
Text
March 20, 2006 C. N. Swenson Site Vice President
AmerGen Energy Company, LLC
P.O. Box 388
Forked River, NJ 08731-0388
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE OYSTER CREEK NUCLEAR GENERATING STATION, LICENSE RENEWAL
APPLICATION (TAC NO. MC7624)
Dear Mr. Swenson:
By letter dated July 22, 2005, AmerGen Energy Com pany, LLC (AmerGen or the applicant) submitted to the U.S. Nuclear Regulatory Commission (NRC or the staff) an application
pursuant to Title 10 of the Code of Federal Regulations Part 54 (10 CFR Part 54), to renew the operating license for Oyster Creek Nuclear Generating Station. The NRC staff is reviewing the
information contained in the license renewal application and has identified, in the enclosure, areas where additional information is needed to complete the review.
These questions were discussed with members of your staff during a conference call on February 2, 2006. A mutually agreeable date for a response is within 30 days from the date of
this letter. If you have any questions, please contact me at 301-415-3191 or via e-mail at
DJA1@nrc.gov
.Sincerely,/RA/Donnie J. Ashley, Project Manager License Renewal Branch A
Division of License Renewal
Office of Nuclear Reactor Regulation Docket No. 50-219
Enclosure:
As stated cc w/encl: See next page
ML060550419 DOCUMENT NAME: C:\MyFiles\Copies\Oyster Creek RAI - CVIB - Ganesh.wpdOFFICEPM:RLRALA:DLROGC NLOBC:RLRANAMEDAshleyYEdmondsAHodgonLLund DATE03/ 7 /0603/ 6 /063/ 8 /0603/ 20 /06
cc: Site Vice President - Oyster Creek Nuclear Generating Station
AmerGen Energy Company, LLC
P.O. Box 388
Forked River, NJ 08731 Senior Vice President of Operations
AmerGen Energy Company, LLC
200 Exelon Way, KSA 3-N
Kennett Square, PA 19348 Kathryn M. Sutton, Esquire Morgan, Lewis, & Bockius LLP
1111 Pennsylvania Avenue, NW
Washington, DC 20004 Kent Tosch, Chief New Jersey Department of
Environmental Protection
Bureau of Nuclear Engineering
CN 415 Trenton, NJ 08625 Vice President - Licensing and Regulatory Affairs
AmerGen Energy Company, LLC
4300 Winfield Road
Warrenville, IL 60555 Regional Administrator, Region I U.S. Nuclear Regulatory Commission
475 Allendale Road
King of Prussia, PA 19406-1415 Mayor of Lacey Township 818 West Lacey Road
Forked River, NJ 08731 Senior Resident Inspector U.S. Nuclear Regulatory Commission
P.O. Box 445
Forked River, NJ 08731 Director - Licensing and Regulatory Affairs AmerGen Energy Company, LLC Correspondence Control P.O. Box 160
Kennett Square, PA 19348 Manager Licensing - Oyster Creek Exelon Generation Company, LLC
Correspondence Control
P.O. Box 160
Kennett Square, PA 19348 Regulatory Assurance Manager Oyster Creek
AmerGen Energy Company, LLC
P.O. Box 388
Forked River, NJ 08731 Assistant General Counsel AmerGen Energy Company, LLC
200 Exelon Way
Kennett Square, PA 19348 Ron Bellamy, Region I U.S. Nuclear Regulatory Commission
475 Allendale Road
King of Prussia, PA 19406-1415 Correspondence Control Desk AmerGen Energy Company, LLC
200 Exelon Way, KSA 1--1
Kennett Square, PA 19348 Oyster Creek Nuclear Generating Station Plant Manager
AmerGen Energy Company, LLC
P.O. Box 388
Forked River, NJ 08731 License Renewal Manager Exelon Generation Company, LLC
200 Exelon Way, Suite 230
Kennett Square, PA 19348 Oyster Creek Nuclear Generating Station cc:
Mr. James Ross Nuclear Energy Institute
1776 I Street, NW, Suite 400
Washington, DC 20006-3708 Mr. Michael P. Gallagher Vice President License Renewal
Exelon Generation Company, LLC
200 Exelon Way, Suite 230
Kennett Square, PA 19348 Mr. Christopher M. Crane President and Chief Nuclear Officer
AmerGen Energy Company, LLC
4300 Winfield Road
Warrenville, IL 60555 Ltr. to C.N. Swenson from Donnie Ashley dated:
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE OYSTER CREEK NUCLEAR GENERATING STATION, LICENSE RENEWAL
APPLICATION (TAC NO. MC7624)Adams Accession No.:
ML060550419 HARD COPY DLR R/F DISTRIBUTION
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RidsNrrDorl Enclosure OYSTER CREEK NUCLEAR GENERATING STATION LICENSE RENEWAL APPLICATION (LRA)
REQUEST FOR ADDITIONAL INFORMATION (RAI)
RAI 3.1.1-1 (A) LRA Section 3.1.2.2.2 states that aging effects due to loss of material due to pitting and crevice corrosion in the isolation condenser will be managed by additional augmented
inspections (i.e., eddy current testing (ET) of the stainless steel tubes and ultrasonic testing (UT) or visual testing (VT) of the tube sheet and channel head). During the telephone
conference dated February 2, 2006, the applicant indicated that thus far no augmented
inspections were performed on components associated with isolation condenser and that
the proposed augmented inspections will be applicable as a part of an aging management
program (AMP) during the extended period of operation. The staff requests the applicant to
provide the following information so that an assessment can be made as to the
effectiveness of the future augmented inspection program of the isolation condenser and its
components.
(1) Previous experience related to the frequency of occurrence of pitting and crevice corrosion in the isolation condenser and its components.
(2) Previous inspection methods and the inspection frequency that were implemented prior to the replacement of some of the isolation condenser components.
(3) Criteria for establishing future augmented inspection frequency.(B) LRA Section 3.1.2.2.4(3) states that aging effects due to stress corrosion cracking (SCC) and intergranular stress corrosion cracking (IGSCC) will be managed by additional
augmented inspections (i.e., ET of the stainless steel tubes and UT or VT of the tube sheet
and channel head). During the telephone conference dated February 2, 2006, the applicant
indicated that thus far no augmented inspections were performed on components
associated with isolation condenser and that the proposed augmented inspections will be
applicable as a part of an AMP during the extended period of operation. The staff requests
the applicant to provide the following information so that an assessment can be made as to
the effectiveness of the future augmented inspection program of the isolation condenser and
its components.
(1) Previous experience related to the frequency of occurrence of SCC and IGSCC in the isolation condenser and its components.
(2) Previous inspection methods and the inspection frequency that were implemented prior to the replacement of some of the isolation condenser components.
(3) Criteria for establishing future augmented inspection frequency.
RAI 3.1.1-2 Item 3.1.1-33 in LRA Table 3.1.1 indicates that the AMP for the RPV inside diameter (ID) attachment welds comply with the recommendations specified in AMP-B.1.4, "BWR Inside
Diameter Attachment Welds Program." LRA AMP-B.1.4 states that the frequency and
the method of inspection specified in the BWRVIP-48, "Vessel ID Attachment Weld Inspection
and Flaw Evaluation Guidelines" report will be implemented for the attachment welds. These
guidelines apply to core spray piping bracket attachments, steam dryer support and hold down
brackets, feedwater spargers, guide rod and surveillance sample holder. According to the
BWRVIP-48 report Section 2.2.3, furnace-sensitized stainless steel vessel ID attachment welds
are highly susceptible to IGSCC. The staff requests the applicant to identify whether there are
any furnace-sensitized stainless steel attachm ent welds at the OCN unit, and explain what type of AMP is implemented for any existing furnace-sensitized stainless steel attachment welds.
The staff also request the applicant to provide details on any additional augmented inspection
program that is implemented for any existing furnace-sensitized stainless steel attachment welds at the OCN unit.
RAI 3.1.1-3 (A) Item 3.1.1-36 in LRA Table 3.1.1 indicates that augmented inspection for the CRD return line weld is required in accordance with NUREG-0619, "BWR Feedwater Nozzle and Control
Rod Drive Return Line Nozzle Cracking" at the OCN unit. NUREG-0619 recommends a
periodic liquid penetrant test (PT) to evaluate the aging effect due to IGSCC in the CRD
return line weld. The applicant in AMP B.1.6, "BWR Control Rod Drive Return Line Nozzle,"
states that it obtained approval from the staff to substitute UT for PT as a part of the
augmented inspection program and this approval is valid only for the current in-service
inspection (ISI) interval. Therefore, the staff requests the applicant to provide justification for
continuing UT inspections in lieu of PT for the subject weld during the extended period of
operation. (B) The staff requests the applicant to provide information whether the CRD return line nozzle has been capped at the OCN unit. If the CRD return line nozzle has been capped, the staff
requests the applicant to provide the following information regarding the cap and the weld:(1)Describe the configuration, location and material of construction of the capped nozzle. This should include the existing base material for the nozzle, piping (if piping
remnants exist) and cap material, and any welds.(2) Describe how this weld and cap is managed in accordance with the guidelines of BWRVIP-75, "BWR Vessel and Internals Project (BWRVIP), Technical Basis for
Revisions to Generic Letter 88-01 Inspection Schedule." (3) Discuss whether the event at Pilgrim (leaking weld at capped nozzle, September 30, 2003) is applicable to OCN unit. The staff issued Information Notice 2004-08, "Reactor Coolant Pressure Boundary Leakage Attributable to Propagation
of Cracking in Reactor Vessel Nozzle Welds" dated April 22, 2004, which states that
the cracking occurred in an 182 weld that was previously repaired extensively.
Discuss any plant experience with previous leakage at the capped nozzle. Include in
your discussion the past inspection techniques applied, the results obtained, and
mitigative strategies imposed. Provide information as to how the plant-specific
experience related to this aging effect impacts the attributes specified in AMP-B.1.6.
RAI 3.1.1-4 (A) AMP B.1.5, references GE Report GE-NE-523-A71-0594, "Alternate BWR Feedwater Nozzle Inspection Requirements," which is not the Nuclear Regulatory Commission (NRC)
approved version of the report. The staff requests the applicant to confirm if OCN will
implement the recommendations of Revision 1, Version A of the report (GE-NE-523-A71-
0594-A, Revision 1) which is approved by the staff.(B) The staff requests the applicant to identify whether the dissimilar metal welds of RPV nozzles, safe end components and piping have previously experienced cracking due to SCC, IGSCC or cyclic loading and the extent of cracking. The applicant should provide
information regarding the extent of mitigative techniques [i.e., structural overlay, mechanical
stress improvement (MSIP)] that were im plemented to mitigate crack propagation due to IGSCC in the dissimilar metal welds between RPV nozzles and safe ends, and welds
between safe ends and piping. In addition, the applicant should provide information on the
inspection methods, sample size, and the frequency of inspections that were used thus far
in these welds and the inspection results. The applicant should provide its basis for using
the current inspection program as an effective AMP in monitoring the aging effect due to
IGSCC in the aforementioned welds.
RAI 3.1.1-5 Item 3.1.1-43 in LRA Table 3.1.1 indicates that cast austenitic stainless steel (CASS) components are used for orificed fuel support components. The staff requests the applicant to
provide the following information on this component so that assessment can be made as to its
susceptibility to thermal and neutron irradiation embrittlement.(a) Information on type of casting (i.e., centrifugal or static)
(b) The composition of CASS (i.e., molybdenum content and delta ferrite values)
(c) Previous plant-specific experience regarding the cracked components and type and extent of subsequent inspection of CASS orificed fuel support components due to neutron and thermal embrittlement. The fluence values should be based on the end
of the extended period of operation.
RAI 3.1.2.1-1 (A) In LRA Table 3.1.2.1.5, the applicant states that it will implement ASME Section XI, ISI program to monitor carbon steel (SA 105 Grade II) cracking in the following RPV
components:
(1) Bottom head drain nozzle; (2) Feedwater and main steam nozzles and safe ends;
(3) Core spray nozzle;
(4) Isolation condenser nozzle;
(5) Top head nozzles;
(6) Top head flange;
(7) Bottom head flange;
(8) RPV shell welds and, (9) Reactor head cooling. The staff requests the applicant to provide the following information related to the subject
aging effect in the aforementioned carbon steel components: (a) Previous plant experience related to cracking in carbon steel RPV components when exposed to treated water.(b) Established mechanism of the cracking in carbon steel RPV components.
(c) The scope and the techniques of the past inspections, the results obtained, applied mitigative methods, repairs, frequency of the inspections and any other relevant
information related to the identification of the subject aging effect.(B) The staff requests the applicant to address whet her there was any previous plant experience related to cracking (not due to SCC or IGSCC) in carbon steel valve bodies of the reactor
head cooling system, when exposed to treated water (Table 3.1.2.1.3).
RAI 3.1.2.1-2 Table IV.B1, Item IV.B1-15, of the GALL Report, Volume 2, Revision 1, recommendsimplementation of AMP B.1.1, "ASME Section XI, Inservice Inspection Subsections IWB, IWC, and IWD," and AMP B.1.2, "Water Chemistry," to manage aging effects due to loss of material, pitting and crevice corrosion in stainless steel and nickel-alloy materials in the reactor vessel
internal (RVI) components. Since this is not included in LRA Table 3.1.2.1.4, the staff requests
the applicant to address these aging effects in LRA Table 3.1.2.1.4.
RAI B.1.9-1 RAI-AMP B.1.9-1(A) - In the final safety analysis report (FSAR) Supplement A.1.9, "BWR Vessel Internals," the applicant states that the BWR vessel internals program is consistent with
the BWRVIP-94, "BWR Vessels and Internals Project, Program Implementation Guideline"
report. The staff requests the applicant to revise AMP B.1.9 to reference the BWRVIP-94
report, and include the following issues related to the scope of the implementation of the
BWRVIP-94 guidelines in AMP B.1.9:
(1) The applicant shall inform the staff of any decision to not fully implement a BWRVIP guideline approved by the staff within 45 days of the report.
(2) The applicant shall notify the staff if changes are made to the RPV and its internals' programs that affect the implementation of the BWRVIP guidelines.
(3) The applicant shall submit any deviation from the existing flaw evaluation guidelines that are specified in the BWRVIP report.
RAI B.1.9-2 The BWRVIP-76, "BWR Core Shroud Inspection and Flaw Evaluation Guidelines" and BWRVIP-104, "Evaluation and Recommendations to Address Shroud Support Cracking in BWRs" reports are currently being reviewed by the staff. The staff requests the applicant to make a commitment that it will comply with all the recommendations that will be specified in the staff's
final safety evaluations (SEs) of these reports, and that it will complete all the license renewal
action items in the final SEs when they are issued.
RAI B.1.9-3 The applicant states that two leaking CRD stub tubes were repaired by using a roll expansion method which was approved by the staff on November 16, 2000, for one refuel cycle only. The
applicant further states that this repair was submitted to the ASME Code in the form of a draft ASME Section XI Code Case N-730, "Roll-Expansion of Class 1 Control Rod Drive Bottom
Head," for review and approval. The applicant intends to apply this repair on a permanent basis
at the OCN unit when Code Case N-730 is approved by the ASME Code and the NRC. If Code
Case N-730 is not approved, the applicant shall submit a permanent repair plan to the staff for
review and approval two years prior to the co mmencement of the extended period of operation.
After the implementation of an approved permanent roll repair, if there is a leak in CRD stub
tubes, the applicant shall commit to immediately repair any leaking CRD stub tubes during the
extended period of operation by implementing a permanent weld repair per the approved ASMESection XI Code Cases with staff conditions, if any.
The staff requests the applicant to revise AMP B.1.9 and FSAR supplement A.1.9 to commit to implementing a staff-approved permanent repair as stated above, which will result in no leakage of the CRD of the stub tubes during the extended period of operation.
RAI B.1.9-4 The applicant states that recent inspections of core spray spargers and core spray piping welds indicated that the mitigation techniques have been proven to be effective in minimizing the crack
growth rates due to IGSCC in the subject components. The staff requests the applicant to
provide further information on its future inspection plans which include the type and frequency of
inspections, inspection methods, sample size, and inspection frequency for the repaired and
non-repaired core spray components during the extended period of operation.
RAI B.1.9-5 The applicant states that recent inspections of the core shroud repair tie rods indicated that the repair techniques are effective in minimizing the crack growth rates in the subject component.
The staff requests the applicant to provide information on its future inspection plans such as
type and frequency of inspections and percentage of the core shroud tie rods that are currently
being inspected. If the inspection sample size is not consistent with the BWRVIP-76 guidelines, the applicant should provide an explanation for this inconsistency. The staff also requests that
the applicant provide its plans regarding the inspection plans (i.e., inspection methods, sample
size, and inspection frequency) of non-repaired core shroud welds during the extended period
of operation.
RAI B.1.9-6 The staff requests the applicant to provide information regarding the type of plugs (i.e., spring-loaded plugs or welded plugs) that were used for plugging the core plate holes at the OCN unit. If spring-loaded core plate plugs were used at the OCN unit, the applicant should provide the type of AMP that is implemented to ensure their integrity.
RAI B.1.9-7 The staff requests the applicant to provide information whether any noble metal chemical application (NMCA) is applied at the OCN unit. Confirm the method of controlling hydrogen
water chemistry and any NMCA as a mitigative method to reduce the IGSCC susceptibility in
the RVI components. Provide details on the methods for determining the effectiveness of
hydrogen water chemistry and/or NMCA by using the following parameters:
(1) Electro Chemical Potential (ECP)
(3) Main steam oxygen content
(4) Hydrogen/oxygen molar ratio.
RAI B.1.9-8 The staff requests the applicant to address how it will use AMP B.1.9 to monitor aging due to loss of material due to pitting and crevice corrosion and aging degradation due to SCC and
IGSCC in non-safety related RVI components (i.e., steam dryer, core shroud heads and
separators, internal feedwater spargers, and RPV surveillance capsule holders).
RAI B.1.9-9 The NRC staff has approved the applicable BWRVIP reports and attached the following required license renewal applicant action items, in accordance with 10 CFR Part 54, when
incorporating the reports in a license renewal application.
The applicant is to verify that its plant is bounded by cited BWRVIP reports. Further, the applicant is to commit to programs described as necessary in the BWRVIP reports to manage
the effects of aging during the period of extended operation. Applicants will be responsible for
describing any such commitments and identifying how such commitments will be controlled.
Any deviations from the AMPs within these BW RVIP reports described as necessary to manage the effects of aging during the period of extended operation and to maintain the functionality of
the components or other information presented in the reports, such as materials of construction, will have to be identified by the applicant and evaluated on a plant-specific basis in accordance
with 10 CFR 54.21(a)(3) and (c)(1).
Section 54.21(d) of the 10 CFR requires that an FSAR supplement for the facility contain a summary description of the programs and activities for managing the effects of aging and the
evaluation of TLAAs for the period of extended operation. Those applicants referencing the
applicable BWRVIP reports shall ensure that the programs and activities specified as necessary
in the applicable BWRVIP reports are summarily described in the FSAR supplement.
Section 54.22 of the 10 CFR requires that each application include any technical specification changes (and the justification for the changes) or additions necessary to manage the effects of
aging during the period of extended operation as part of the renewal application. The applicable
BWRVIP reports may state that there are no generic changes or additions to technical specifications associated with the reports as a result of the applicant's aging management review and that the applicant will provide the justification for plant-specific changes or additions.
Those referencing the applicable BWRVIP reports shall ensure that the inspection strategy
described in the reports does not conflict with, or result in, any changes to their technical
specifications. If technical specifications changes do result, then the applicant must ensure that
those changes are included in its application for license renewal.
If required by the applicable BWRVIP report, the applicant referencing a particular report for license renewal should identify and evaluate any potential TLAA issues and/or commitments to
perform future inspections when inspection tooling is made available.
The staff requests the applicant to provide the necessary commitments, information and changes as described above for each of the following applicable BWRVIP reports:
BWRVIP-05, "Reactor Vessel Shell Weld Inspection Guidelines."
BWRVIP-18, "BWR Core Spray Internals Inspection and Flaw Evaluation Guidelines."
BWRVIP-25, "BWR Core Plate Inspection and Flaw Evaluation Guidelines."
BWRVIP-26, "BWR Top Guide Inspection and Flaw Evaluation Guidelines."
BWRVIP-27-A, "BWR Standby Liquid Control System/Core Plate P Inspection and Flaw Evaluation Guidelines."
BWRVIP-38, "BWR Shroud Support Inspection and Flaw Evaluation Guidelines."
BWRVIP-47, "BWR Lower Plenum Inspection and Flaw Evaluation Guidelines."
BWRVIP-48, "Vessel ID Attachment Weld Inspection and Flaw Evaluation Guidelines."
BWRVIP-49, "Instrument Penetration Inspection and Flaw Evaluation Guidelines."
BWRVIP-74-A, "BWR Reactor Pressure Vessel Inspection and Flaw Evaluation Guidelines."
BWRVIP-75, "BWR Vessel and Internals Project (BWRVIP), Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedule."
BWRVIP-76, "BWR Core Shroud Inspection and Flaw Evaluation Guidelines."
BWRVIP-78 , "BWR Integrated Surveillance Program (ISP) Plan."
BWRVIP-86, "BWR Vessel and Internals Project, BWR Integrated Surveillance Program Implementation."
Other reports applicable to license renewal.
RAI B.1.23-1 The staff requests the applicant to include the following statement (shown below) in the FSAR Section A.1.23, "Reactor Vessel Surveillance" of the LRA.
The applicant states that it will implement the BWRVIP integrated surveillance program (ISP) as specified in BWRVIP-116, "BWR Vessel Internals Project Integrated Surveillance Program
Implementation for License Renewal" at the OCN unit. The staff is currently reviewing the
BWRVIP-116 report, and if this report is not approved by the staff, the applicant must submit a
plant-specific surveillance program for the O CN unit, two years prior to the commencement of the extended period of operation. The following commitment shall be included in the FSAR
Section A.1.23, "Reactor Vessel Surveillance" of the LRA.
BWRVIP ISP as specified in BWRVIP-116, "BWR Vessel Internals Project Integrated Surveillance Program Implementation for License Renewal" and approved by the staff will be implemented, or if the ISP is not approved two years
prior to the commencement of the extended period of operation, a plant-specific surveillance program for the OCN unit will be submitted.
RAI B.1.23-2 The staff requests the applicant to include the following statement (shown below) in the FSAR Section A.1.23, "Reactor Vessel Surveillance" of the LRA.
10 CFR Part 50, Appendix H, requires that an integrated surveillance program (ISP) used as a basis for a licensee implemented reactor vessel surveillance program be reviewed and
approved by the NRC staff. The ISP to be used by the applicant is a program that was
developed by the BWRVIP. The applicant will apply the BWRVIP ISP as the method by which
the OCN unit will comply with the requirements of 10 CFR Part 50, Appendix H. The BWRVIP
ISP identifies capsules that must be tested to monitor neutron radiation embrittlement for all
licensees participating in the ISP and identifies capsules that need not be tested (standby
capsules). Tables 2-3 and 2-4 of the BWRVIP-116 report indicate that the capsules from OCN
unit are not tested. These untested capsules were originally part of the applicant's plant-specific
surveillance program and have received signi ficant amounts of neutron radiation.
The following commitment shall be included in the FSAR Section A.1.23, "Reactor Vessel
Surveillance" of the LRA.If the OCN standby capsule is removed from the RPV without the intent to test it,the capsule will be stored in a manner which maintains it in a condition which would permit its future use, including during the period of extended operation, if necessary.