ML050110235
| ML050110235 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 01/03/2005 |
| From: | Spina J Constellation Energy Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NMP1L 1909, TAC MC3272, TAC MC3273 | |
| Download: ML050110235 (24) | |
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Or'Z Constellation Energy Nine Mile Point Nuclear Station P.O. Box 63 Lycoming, New York 13093 January 3, 2005 NMP1L 1909 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001
SUBJECT:
Nine Mile Point Units 1 and 2 Docket Nos. 50-220 and 50-410 Facility Operating License Nos. DPR-63 and NPF-69 License Renewal Application - Responses to NRC Requests for Additional Information Regarding the Reactor Recirculation Systems (TAC Nos. MC3272 and MC3273)
Gentlemen:
By letter dated May 26, 2004, Nine Mile Point Nuclear Station, LLC (NMPNS) submitted an application to renew the operating licenses for Nine Mile Point Units 1 and 2.
In a letter dated December 3, 2004, the NRC requested additional information regarding the Reactor Recirculation Systems that are described in Sections 3.1.2.A.4 and 3.1.2.B.4 of the License Renewal Application (LRA). The NMPNS responses to these requests for additional information are provided in Attachment 1. Attachment 2 provides a list of the regulatory commitments associated with this submittal.
If you have any questions about this submittal, please contact Peter Mazzaferro, NMPNS License Renewal Project Manager, at (315) 349-1019.
Very truly yours, PreiZ P Jaws A. Spina Maice President Nine Mile Point JAS/DEV/jm at Dq
Page 2 NMP1L 1909 STATE OF NEW YORK TO WIT:
COUNTY OF OSWEGO I, James A. Spina, being duly sworn, state that I am Vice President Nine Mile Point, and that I am duly authorized to execute and file this supplemental information on behalf of Nine Mile Point Nuclear Station, LLC. To the best of my knowledge and belief, the statements contained in this submittal are true and correct. To the extent that these statements are not based on my personal knowledge, they are based upon information provided by other Nine Mile Point employees and/or consultants. Such information has been reviewed in accordance with company practice and I believe it to be reliable.
Subscribed and sworn before me, a Notary Public in and for e State of New York and County of Oswego, this 3 day of S 2005.
WITNESS my Hand and Notarial Seal:
Notary Public My Commission Expires:
/-3i-oS Date SANDRA A. OSWALD Notary Public. State of New York Cualilied In Oswego CpunN Commission Expires
/
Attachments:
- 1. Responses to NRC Requests for Additional Information (RAI) Regarding the Reactor Recirculation Systems Described in Sections 3.1.2.A.4 and 3.1.2.B.4 of the License Renewal Application
- 2. List of Regulatory Commitments cc:
Mr. S. J. Collins, NRC Regional Administrator, Region I Mr. G. K. Hunegs, NRC Senior Resident Inspector Mr. P. S. Tam, Senior Project Manager, NRR Mr. N. B. Le, License Renewal Project Manager, NRR Mr. J. P. Spath, NYSERDA
ATTACHMENT 1 Nine Mile Point Nuclear Station Responses to NRC Requests for Additional Information (RAI)
Regarding the Reactor Recirculation Svstems Described in Sections 3.1.2.A.4 and 3.1.2.B.4 of the License Renewal Application This attachment provides Nine Mile Point Nuclear Station, LLC (NMPNS) responses to the requests for additional information contained in the NRC letter dated December 3, 2004, regarding the Reactor Recirculation Systems. Each NRC RAI is repeated, followed by the NMPNS response for Nine Mile Point Unit 1 (NMPl) and/or Nine Mile Point Unit 2 (NMP2), as applicable. Revisions to the License Renewal Application (LRA) are described where appropriate. The revisions are highlighted by shading unless otherwise noted.
Note: Section, table, and page numbers cited for LRA Section 3.1 refer to the revised version of LRA Section 3.1 that was submitted by NMPNS letter NMP1L 1892, dated December 6, 2004.
(1) The following RAIs are related to NMPI:
RAI 3.1.2.A.4-1 In Section 3.1.2.A.4 of the LRA, the applicant lists the followving environments which the NMPJ Reactor Recirculation System components are exposed:
o Closure boltingfor non-borated water systems with operating temperatures equal to or greater than 212 degree F (degF) o Treated water, temperature < 140 degF, lowflow o Treated water or steam, temperature > 482 degF o Treated water or steam, temperature > 482 degF, lowflow However, some of the components such as valves and piping are also exposed to air and hydratlicfluid. Please provide information as jo why air and hydraulicfluid are not listed in the environment.
Response
The NMP 1 Reactor Recirculation System contains components subject to an air environment but not a hydraulic fluid environment. The external surfaces of the system piping and fittings, pumps and valves are exposed to air and should have been identified in LRA Section 3.1.2.A.4 Page 1 of 21
and Table 3.1.2.A-4. The applicable aging effect for the carbon or low alloy steel valves is Loss of Material, for which the credited aging management program is the Systems Walkdown Program. For the remaining components, which are all stainless steel, there are no aging effects requiring management. The associated changes to this LRA section and table are provided below.
Hydraulic fluid is not an applicable environment for the NMP 1 Reactor Recirculation System since none of the components are hydraulically controlled, unlike NMP2. NMP2 contains hydraulic flow control valves to control flow through the reactor recirculation system. However, NMP1 is a BWRI2 which utilizes variable speed pumps to control flow through the system.
Valves provide loop isolation only. Additional information regarding the designs of the NMP1 and NMP2 Reactor Recirculation Systems are contained in the applicable scoping and screening sections of the LRA (2.3.1.A.4 and 2.3.1.B.4).
Similar to the above change for the NMP1 Reactor Recirculation System, the NMP2 system components are also exposed to an external environment of air. LRA Table 3.1.2.B-4 requires a revision to identify that this environment applies to the piping and fittings, pumps, and valves, and that Loss of Material and the Systems Walkdown Program are the associated aging effect and aging management program. Changes to Table 3.1.2.B-4 are provided below.
LRA Revisions In LRA Section 3.1.2.A.4 (page 3.1-8), under the "Environments" heading, "Air" is added as an applicable environment for the NMP 1 Reactor Recirculation System.
LRA Tables 3.1.2.A-4 and 3.1.2.B-4 are revised to add rows for piping and fittings, pumps, and valves exposed to an air environment, as shown on the following pages.
Page 2 of 21
Table 3.1.2.A-4 Reactor Vessel, Internals, and Reactor Coolant System I Reactor Recirculation Svstem - Summarv of Aaina Manaaement Evaluation NMPI Aging ffectNUREG-Component Intended AginEfe Aging Management 1801 Table I Notes TpFucinManagement Program Volume 2 Item
_ _ _ _Ite m_
Piping and PB Wrought Air None None None Fittings Austeniti_
Stainless Steel Pumps PB Cast Austenitic Air None 1 None 1
1 l None l Stainless Steel l
yalves PB Carbon or Low Air Loss of Systems Walkdown H
Alloy Steel Material Proqrarmi (Yield Strength
-<1 00 Ksi)
Cast Austenitid Air None None None Stainless Steel Wrought Air None None None Austenitic_
Stainless__teel Page 3 of 21
Table 3.1.2.B-4 Reactor Vessel, Internals, and Reactor Coolant System NMP2 Reactor Recirculation System - Summary of Aging Management Evaluation Aging ffectNUREG-Component Intended Material Environmet RenguiErinegct Aging Management 1801 Table 1 Notes Type Function Matrilnnvrometmequrng Program Volume 2 Item ManagmentItem Piping and PB Nickel Based Air None None None Fittings Alloys Wrought Air None None None Austenitic
_Stainless Steel Pumps PB Cast Austenitic Air None None None Stainless Steel Wrought Air None None None AustenitIic Stainless 'Steel Valves PB Cast Austenitic Air None None None Stainless Steel Wrought Air None None None Austenitic Stainless St.ee Page 4 of 21
RAI 3.1.2.A.4-2 The applicant identifies cracking as awl applicable aging effectfor the recirculation system closure bolting, piping andfittings, recirculation pumps, and valves; however, the applicant has not considered culnulativefatigue damagefor NMPJ piping as an aging effect. Please provide information relative to exclusion of cumulative fatigue damage for NMPI reactor recirculation piping.
Response
Subsequent to submittal of the original LRA, dated May 26, 2004, NMPNS submitted a revised LRA Section 3.1, "Aging Management of Reactor Vessel, Internals, and Reactor Coolant Systems," in letter NMPIL 1892, dated December 6, 2004. Included in the revisions were changes that identified cumulative fatigue damage as an aging effect requiring management, where applicable. For the NMP1 Reactor Recirculation System piping and fittings, "Cumulative Fatigue Damage" was added as an aging effect in addition to "Cracking." Cumulative Fatigue Damage was also added as an aging effect requiring management for the recirculation system closure bolting, flow elements, pumps, pump seal flanges and valves. The Fatigue Monitoring Program is the credited aging management program for this aging effect.
Based upon the above described revision to the NMP1 Reactor Recirculation System aging management review, it is not necessary to provide information relative to the exclusion of cumulative fatigue damage for the system piping.
RAI 3.1.2.A.4-3 In LRA Table 3.1.2.A-4, the applicant identifies cracking as an applicable aging effect for the recirculation system austenitic stainless steel components (piping and fittings, tubing, valve bodies, flow elements, thermnowells, restricting orifices) andfor the high-strength low-alloy steel primary pressure closure bolting exposed to reactor coolant water. The applicant also identifies this aging effectfor cast stainless steel components exposed to reactor coolant water. The applicant identifies crack initiation and growth due to thermal and mechanical loading as an applicable aging effect for small-bore stainless steel piping and fittings and lolv-alloy steel pressure boundary closure bolting in the reactor recirculation system. Please provide information to indicate that, for NMPJ, the applicant has noflaws evaluated in accordance with IJVB-3600 "Analytical Evaluation of Flaws" under the inservice inspection program ofASME Code,Section XI since this would require a time-limited aging analysis (TLAA,) under IO CFR 54.21(c).
Response
The NMP1 Reactor Recirculation System undergoes examination and flaw evaluation in accordance with the American Society of Mechanical Engineers (ASME) Code Section XI.
Plant-specific experience has shown that five welds contained indications that were evaluated in accordance with paragraph IWvB-3600, "Analytical Evaluation of Flaws." For each of these welds, subsequent reexamination and evaluation determined that the indications were not caused Page 5 of 21
by intergranular stress corrosion cracking (IGSCC) but were fabrication related. However, increased frequency of inspection (i.e., every inspection period) is still applicable.
For the five recirculation system welds identified above, the following provides summary information regarding the discovery, evaluation and resolution of the indications.
During the NMP1 1997 refueling outage, an indication was identified in a piping weld (32-WD-050) in the suction riser of the #12 recirculation loop. The flaw is circumferentially oriented on the pipe inside diameter (ID) and is located in the piping base metal immediately adjacent to a circumferential weld. Flaws of this type may result from an original weld defect or from IGSCC attack. Since the indication did not meet the acceptance criteria of ASME Section XI Table IWB-3410-1, a flaw evaluation in accordance with IWB-3600 was performed to accept the weld for continued plant operation. ByNMP letterNMP1L 1201, dated April 7, 1997, the flaw evaluation was submitted to the NRC for review and approval in accordance with plant Technical Specifications and Generic Letter (GL) 88-01. This letter also stated that this weld would be reclassified as a Category F weld and examined during each refueling outage. By letter dated April 30, 1997, the NRC transmitted their safety evaluation that concluded that NMP 1 could operate safely for the following operating cycle with the weld indication.
During the 1999 refueling outage, weld 32-WD-050 was re-inspected. The results indicated that the size did not change and, therefore, it was concluded that the indication was an original construction weld defect and not a flaw associated with IGSCC. These results were submitted to the NRC in letter NMP1L 1439, dated June 1, 1999. It was indicated in this letter that the weld would be returned to the original GL 88-01 Category A classification, and that the weld would be re-inspected during each of the next three inspection periods, in accordance with ASME Code Section XI, Subsection IWB, paragraph IWB-2420(b). By letter dated July 23, 1999, the NRC concurred with the NMP conclusion regarding weld 32-WD-050 except that the weld should not be returned to Category A until the re-inspection program is completed. As such, weld 32-WD-050 remains a Category F weld and is re-inspected each inspection period (until the end of the current inspection interval) to provide additional assurance that the indication is not caused by IGSCC.
With respect to the remaining four welds identified above, indications were discovered during the NMP1 1999 refueling outage. Two indications were identified in one safe-end to elbow weld and one indication in each of three additional safe-end to elbow welds in reactor recirculation loops 12, 13, 14, and 15. These indications exceeded the specifications of ASME Section XI Table IWB-3514-2 and flaw evaluations were performed in accordance with IWB-3600 since IGSCC could not be conclusively excluded as a cause. The welds were also reclassified as Category F such that re-inspection every inspection period would be required. These findings and results were submitted to the NRC for review and approval in letter NMP1L 1467, dated September 14, 1999. The NRC accepted the NMP evaluation and conclusions in their safety evaluation dated January 14, 2000.
During the subsequent refueling outage in 2002, the re-inspection of these four welds indicated that there was no growth in the indications. Based upon these results and additional evaluations, NMP concluded that the indications were fabrication related and not attributable to IGSCC.
Page 6 of 21
NMPNS submitted these findings and results to the NRC in letter NMP1L 1673, dated September 13, 2002, and indicated that the re-inspection frequency would be in accordance with the ASME Code,Section XI, Subsection IWB-2420(b) to confirm the absence of IGSCC, rather than follow the schedule for Category F welds due, in part, to the significant dose exposure associated with performing the exams. By letter dated March 27, 2003, the NRC transmitted their safety evaluation and accepted the NMP conclusions and re-inspection schedule.
Based upon the above, the NMP1 Reactor Recirculation System contains five welds that have had flaw evaluations performed in accordance with IWB-3600. However, since each indication was determined not to be caused by IGSCC, there are no time-limited aging analyses (TLAAs) required relative to the subject flaw evaluations.
(2) The following RAls are related to both NMP1 and NMP2:
RAI 3.1.2.C.4-1 In LRA Tables 3.1.2.A-4 and 3.1.2.B-4, there are non-safety related (NSR) piping, fittings, and equipinent of any material whose failure could affect safety-related equipment within the scope of license renewal. The applicant identified cracking and loss of material as aging effects requiring management. The aging management program to manage the aging effects is the JMater C6hemistry Control Program. Please provide additional infonnation to indicate if these components in both NMP1 and NMP2 are to be covered in the One-Time Inspection Program for condition evaluation prior to the extended period of operation and to perform periodic inspection of the components under the preventative maintenance (PM) Program.
Response
For the NMPl Reactor Recirculation System non-safety-related (NSR) piping, fittings, and equipment (LRA Table 3.1.2.A-4), the associated components are small-bore components associated with non-safety-related instrumentation. For the NMP2 Reactor Recirculation System non-safety-related piping, fittings, and equipment (LRA Table 3.1.2.B-4), the associated components are also small-bore components associated with non-safety-related instrumentation.
However, the NMP2 system also includes small-bore piping, fittings, and equipment (such as pumps, valves, filters, and heat exchangers) associated with the hydraulic power units for the recirculation loop flow control valves. The system pump bearing coolers, motor winding coolers, and pump seal water lines also contain small-bore components. Currently, only the Water Chemistry Control Program is credited with managing the aging effects of Cracking and Loss of Material. However, to validate that the aging effects are not expected to occur for these components, the One-Time Inspection Program will be added for those components exposed to a water environment. Those components associated with the hydraulic power units for the NMP2 recirculation flow control valves contain hydraulic fluid and will be conservatively included in the One-Time Inspection Program for the aging effect of Loss of Material only. The Water Chemistry Control Program would not apply to these components.
Page 7 of 21
The changes to Tables 3.1.2.A-4 and 3.1.2.B-4 needed to reflect the above discussion are shown below.
LRA Revisions LRA Tables 3.1.2.A-4 (page 3.1-57) and 3.1.2.B-4 (page 3.1-95) are revised to add the One-Time Inspection Program for NSR piping, fittings, and equipment exposed to a water environment. Table 3.1.2.B-4 is also revised to add a row for NSR piping, fittings, and equipment exposed to a hydraulic fluid environment. These revisions are shown on the following pages.
Page 8 of 21
Table 3.1.2.A-4 Reactor Vessel, Internals, and Reactor Coolant System I Rpactor Recirculation Svstam - Summarv of Aaina Manaaement Evaluation NMP1 Copnn neddAging Effect AgnNMngeet R81EaleGNoe Typoe Functionde Material Environment Requiring AgnMage nt 10Tbl Nos Type Function Management Program Volume 2 Item
_ _ _Ite m
NSR piping, PFASRE Any Treated Water, Cracking; Loss Water Chemistry J
fittings, and temperature of Material Control Program equipment
< 140'F, Low Flow One-Time nsoectioni Program Treated Water or Steam, temperature
Ž4820F Treated Water or Steam, temperature
Ž4820F, Low
_Flow Page 9 of 21
Table 3.1.2.64 Reactor Vessel, Internals, and Reactor Coolant System NMP2 Reactor Recir ulation System - Summary of Ag ing Management Evaluation Copnn neddAging Effect NUREG-Noe Comoonent Function Material Environment Requiring Aging Management 1801 Table I Notes TpFucinManagement Program Volume 2 Item NSR piping, PFASRE Any Treated Water, Cracking; Loss Water Chemistry J
fittings, and temperature of Material Control Program equipment
< 1400 F, Low Flow One-Time Inspection Program Treated Water, temperature 21400F, but
< 212 0F, Low Flow Treated Water or Steam, temperature 24820F Treated Water or Steam, temperature 24820F, Low Flow Hydraulic Fluid Loss of One-Time Inspection Material Proqramr Page 10 of 21
RAI 3.1.2.C.4-2 In LRA Table 3.1.2.A-1 for NMPI applicable to vessel drain line, the applicant identifies ASME Section XIlnservice Inspection as an aging managenmentprogramn to detect loss of material.
Because of the size of the drain line, volumetric examination is not required by the Code.
Please provide additional information on the aging management programn on how it is applicable to reactor vessel drain line [due to its size and being mostly inaccessiblel in the detection of loss of material under the inservice inspection progranifor NMPJ as well as NMP2.
Response
For NMP I and NMP2, the subject drain lines are covered by the respective plant's ASME Section XI Inservice Pressure Testing Program, Water Chemistry Control Program, and Fatigue Monitoring Program, which will continue into the period of extended operation. The ASME Section XI pressure test is performed every refueling outage. As a function of the pressure test, concurrent VT-2 examination is performed to IWB-3522 acceptance criteria. Relevant conditions, per IWA-9000, that are detected during this examination must be corrected to meet the requirements of IWA-5250 prior to return of the system to continued service. These relevant conditions include discovered leakage, which could be as a result of through-wall pitting or crevice corrosion, the applicable loss of material mechanisms for stainless steel piping and components. Also performed under the inservice inspection (ISI) program, to IWB-3517 acceptance standards, is a VT-I examination of all reactor vessel drain line bolting, studs, and nuts at every inspection interval.
As a function of the site's Risk Informed ISI program, two other examination activities are also performed. A sampling of system piping welds are ultrasonically examined in the same sequence during successive inspection intervals to IWB-3514 acceptance standards. Also performed during each refueling outage are VT-2 examinations of system socket welds to IWB-3522 acceptance standards.
The combination of all of these examination activities, in conjunction with the mitigative functions of the Water Chemistry Control and Fatigue Monitoring Programs, is deemed to be sufficient for effective aging management of the reactor vessel drain line during the period of extended operation.
RAI 3.1.2.C.4-3 In LRA Table 3.1.2.A-4, the applicant identifies loss of material as an aging effect for the reactor recirculation system high-strength low-alloy steel bolting exposed to air with operating temperatures in excess of 212 degF. However, under discussion for itemn numbers 3.J1. LA-26 and
- 3. 1.1.B-26, the applicant states that the aging mechanism of loss ofpreload due to stress relaxation is not an aging effect/mechanism at NMPI (and NMP2) for this environment. Please provide additional information on the effects of oxygenated water on the bolting material at temperatures >482 degF and a justification for excluding periodic inspection of the closure boltingfor indication of loss ofpreload.
Page 11 of21
At.
Response
Subsequent to the submittal of the original LRA, dated May 26, 2004, NMPNS submitted a revision to LRA Section 3.1 in letter NMPIL 1892, dated December 6, 2004. This revision deleted the statements regarding loss of preload as not being applicable at NMP in Items 3.1.1.A-26 and 3.1.1.B-26, and added loss of preload as an aging effect requiring management (AERM) for the closure bolting for the NMP1 and NMP2 Reactor Recirculation Systems in LRA Tables 3.1.2.A-4 and 3.1.2.B-4, respectively. These revisions are in line with the guidance provided in NUREG-1 801 for reactor coolant pressure boundary closure bolting. These revisions were incorporated in response to the results of the NRC audits of the NMP aging management reviews conducted in the fourth quarter of 2004. Based upon these revisions, additional information on the effects of oxygenated water on the bolting material at temperatures >4820F and a justification for excluding periodic inspection of the closure bolting for indication of loss of preload is not necessary.
Note that the revised LRA Tables 3.1.2.A-4 and 3.1.2.B-4 currently indicate that the loss of preload AERM is managed by the ASME Section XI Inservice Inspection Program. NMPNS is in the process of developing a Bolting Integrity Program which includes use of ASME Section XI inservice inspections. When this program development is complete, a LRA supplemental letter will be submitted to replace the ASME Section XI Inservice Inspection Program with the Bolting Integrity Program. The LRA Section 3.1 table line items for the NMPI and NMP2 Reactor Recirculation System closure bolting will be among the LRA locations where the aging management program will be changed to the Bolting Integrity Program. This additional revision to LRA Section 3.1 will be submitted by February 28, 2005.
RAI 3.1.2.C.4-4 and RAI 3.1.2.C.4-5 In LRA Tables 3.1.J.A and 3.1.1.B, under item numbers 3.J1. LA-23 and 3.1.1.B-23for cast austenitic stainless steel (CASS) pump casing, the applicant identifies loss offracture toughness due to thermal aging embrittlement as an applicable aging effect and inservice inspection as the aging management program. In orderfor the staff to properly evaluate these items, please (1) provide information regarding NMPJ 's and NMP2 s plant-speci~fc experience in ultrasonic examination of CASS components since ultrasonic examination of cast austenitic stainless steel material with coarse grain structure is impractical, and (2) provide justification for not considering cumulativefatigue damage as an aging effect for the reactor recirculation pump casing welds.
Response
(1) NMPNS agrees that ultrasonic testing of cast austenitic stainless steel (CASS) material with course grain structure is impractical, and thus such testing is not performed at NMP1 or NMP2. LRA Table Item Numbers 3.1.1.A-23 and 3.1.1.B-23 are applicable only for the NMP1 and NMP2 Reactor Recirculation Pumps. These pumps are the only ASME Section XI Class 1 pumps at either station and are inspected in accordance with the ASME Section XI Code requirements. The Code requirements are a visual examination (VT-3) of the interior surfaces of the pump when disassembled for maintenance. A visual examination Page 12 of 21
(VT-2) is also required each refueling outage during the reactor coolant system boundary leak test.
(2) There are no welds in the NMPI or NMP2 reactor recirculation pump casings. They are of bolted construction. Therefore, consideration of cumulative fatigue damage as an aging effect is not applicable.
RAI 3.1.2.C.4-6 In LRA Table 3.1. LA, item number 3.1. 1.A-09, the applicant identifies stress corrosion cracking and cyclic loading as the aging effects for the isolation condenser and credits the Preventive Maintenance (PM) Program as the aging management program to manage the aging effects.
The isolation condensers are part of the reactor coolant pressure boundary and, therefore, should be inspected in accordance with the ASME Code, Section N. The PMprogram does not require volumetric examination to assure structural integrity ofpressure boundary material or welds. Please provide additional information on how the aging effects of cracking in stainless steel tubes and in shell welds are managed through PMprogram that relies on visual inspection to prevent the loss of its intendedfiunction.
Response
Subsequent to the submittal of the original LRA, dated May 26, 2004, NMPNS revised the credited aging management programs for the NMPI Isolation Condensers. In NMPNS letter NMPIL 1892, dated December 6, 2004, LRA Table 3.1.l.A was revised to indicate that, for Item Number 3.1.1.A-09, the ASME Section XI Inservice Inspection (Subsections IWB, IWC, IWD) and Water Chemistry Control Programs are credited in addition to the Preventive Maintenance Program for managing the aging effect of cracking for the stainless steel tubes and shell welds.
Continuous radiation monitoring of the Isolation Condenser shell is also credited. These changes utilize detection methods in addition to visual inspection to ensure aging degradation is identified and corrected prior to a loss of intended function. The above revision to LRA Table 3.1.1.A brings the credited programs in line with the guidance provided in NUREG-1801, Item IV.Cl.4-a, for the NMP1 Isolation Condensers.
A response to a related information request, RAI 3.2-12, was provided in NMPNS letter NMP1L 1902, dated December 21, 2004. As noted in that letter, the corresponding changes to LRA Table 3.2.2.A-3 wvill be submitted by February 28, 2005.
RAI 3.1.2.C.4-7 The applicant credits LRA Appendix B2.1.6, BWR Stress Corrosion Cracking Program, for mitigating intergranular stress corrosion cracking (IGSCC) in austenitic stainless steel reactor coolant pressure boundary components, including piping four inches and greater nominal pipe size. The applicant also states that the BWR Stress Corrosion Cracking Program is based on industry guidelines approved by the NRC. Please provide information about its plant-specific experience related to IGSCC of the reactor coolant pressure boundary piping, mitigative actions taken, and the revised inspection schedules following the B VRVIP-75 guidelines. Please also Page 13 of 21
I provide information on implementation of hydrogen water chemistry and noble metal chemical application (NMMCA) at both NMPJ and NMP2, and how this implementation has affected monitoring of ivater chemistry parameters.
Response
In the late 1970's and early 1980's, NMPI identified IGSCC-related indications in reactor recirculation system piping welds. As a mitigative measure, the original piping was replaced with low carbon austenitic stainless steel material that is considered to be IGSCC-resistant material. Since the piping replacement predated the issuance of Generic Letter (GL) 88-01, the replacement material did not meet all the criteria of GL 88-01. The criterion that was not met was with regard to the ferrite content of the weld filler metal. However, most of the welds in the replacement reactor recirculation system piping were categorized as Category A welds per GL 88-01, as accepted by the NRC in a Safety Evaluation Report dated May 15, 1990. NMPI currently has 143 Category A welds, 142 Category D welds, one (1) Category E weld, five (5)
Category F welds, and 19 Category G welds. The Category A welds are included in the Alternate Risk-Informed Inservice Inspection Program.
With respect to more recent plant-specific experience at NMP1, indications were identified in four reactor recirculation system welds during thel999 refueling outage. Two indications were identified in one safe-end to elbow weld and one indication each in three additional safe-end to elbow welds. The indications were conservatively assumed to be caused by IGSCC and flaw evaluations were performed using the methodology of NUREG-0313, Revision 2, to estimate the flaw growth due to IGSCC. Since the welds were not repaired, the re-inspection frequency was increased to every refueling outage, in accordance with NUREG-0313, Revision 2, requirements.
NMPNS submitted the flaw evaluations in letterNMPIL 1467, dated September 14, 1999. The NRC reviewed these evaluations and concluded, in a safety evaluation dated January 14, 2000, that the welds were acceptable for continued operation without repair during the subsequent operating cycle. Re-inspection of these four welds during the 2001 refueling outage revealed essentially no growth in the indications. Based upon the re-inspection results and additional reviews and evaluations, it was determined that the indications were fabrication-related and not due to IGSCC. These results were submitted to the NRC in letter NMP1L 1673, dated September 13, 2002, and the NRC concurred with the results in their safety evaluation dated March 27, 2003.
For NMP 1, the inspection scope and schedule for IGSCC is in accordance with GL 88-01, as modified by BWRVIP-75. NMP1 has implemented hydrogen water chemistry (HWC) and noble metal chemical application (NMCA). The current inspection schedule, except for Category A welds, is in accordance with the revised inspection frequency allowed by BWRVIP-75 for normal water chemistry. NMP1 intends to adjust the inspection schedules in the future to take credit for the reduced inspection frequency allowed by BWRVIP-75 for locations where HWC/NMCA is effective.
NMP2 was constructed during the 1970's and 1980's and IGSCC-resistant materials were generally used to fabricate the reactor coolant pressure boundary. NMP2 has 113 Category A Page 14 of 21
f welds and 47 Category D welds. The Category A welds are inspected under the Alternate Risk-Informed Inservice Inspection Program.
For NMP2, there have been two plant-specific occurrences related to potential IGSCC indications. The first involves a High Pressure Core Spray (HPCS) nozzle safe-end to safe-end extension weld (KC-32) indication identified during the first refueling outage. After evaluating the indication, NMPNS applied the Mechanical Stress Improvement Process (MSIP) to improve the residual stress distribution in the region of the flaw to eliminate the potential for flaw growth.
This weld was also reclassified to Category F in accordance with GL 88-01. As such, the frequency of weld re-inspection was increased to once each refueling outage. By letter NMP2L 1572, dated September 22, 1995, NMPNS requested NRC approval to reclassify the weld as a Category E weld since the required number of inspections had been performed which showed no adverse change in the indication. By NRC letter dated February 12, 1996, the staff concurred with the NMPNS request. The HPCS nozzle safe-end to safe-end extension weld (KC-32) is now classified as a GL 88-01 Category E weld and is inspected once every six years. The results from the last two inspections also showed no detectable growth from the previous examinations.
The second occurrence at NMP2 identified an indication on the safe-end side of the Feedwater nozzle-to-safe end weld (KB-20) during the 1998 refueling outage (RFO6). The weld consists of Alloy 182 buttering on the nozzle and Alloy 82 filler. Both of these are nickel-based alloy weld fillers. While the indication was believed to be fabrication related, NMPNS performed a flaw evaluation and submitted it to the NRC in letters dated June 17, 1998 and June 23, 1998. The NRC reviewed this evaluation and concluded, in their June 25, 1998 safety evaluation, that continued plant operation without repairs to the weld was acceptable for the following operating cycle. Since the weld was not repaired, the re-inspection frequency was increased to every refueling outage. During the examination of this weld during the 2000 refueling outage (RFO7), growth of the indication was detected, suggesting that IGSCC was active in the nozzle. Ultrasonic examination data from RFO6 for the feedwater nozzle was reanalyzed using improved equipment and compared to the RFO7 ultrasonic examination data.
An additional indication was determined to have been present in 1998 that was not detected at the time. Also, comparison of the reanalyzed RFO6 data with the RFO7 data shows that the indication is increasing in through-wall dimension. The growth was judged to be due to IGSCC.
The feedwater nozzle containing these indications was repaired with a weld overlay during the 2000 refueling outage. This is the only mitigative action that has been required for IGSCC of components within scope of the BWR SCC program.
For NMP2, the inspection scope and schedule for IGSCC is also in accordance with GL 88-01, as modified by BWRVIP-75. The current inspection schedule, except for Category A welds, is in accordance with the revised inspection frequency allowed by BWRVIP-75 for normal water chemistry. NMP2 intends to adjust the inspection schedules in the future to take credit for the reduced inspection frequency allowed by BWRVIP-75 for locations where HWC/NMCA is effective.
With respect to implementation of HWC and NMCA, NMP I began injecting hydrogen into reactor water in June 2000 and treated the reactor vessel internals with noble metal chemicals in Page 15 of 21
9' May 2000. For NMP2, hydrogen injection commenced in January 2001 and the addition of noble metals was initiated in September 2000.
With respect to impacts on water chemistry parameters, the only significant change for NMP 1 to reactor water chemistry when operating under HWC versus normal water chemistry is that the electrochemical potential (ECP) is monitored, with a goal of < - 0.23V SHE (standard hydrogen reference electrode) to verify the effectiveness of HWC. NMP 1 reactor water chloride, sulfate, and conductivity action levels are Technical Specification requirements that are based on Electric Power Research Institute (EPRI) TR-103515-Rl, "BWR Water Chemistry Guidelines - 1996 Revision," December 1996. Other parameters are consistent with EPRI TR-103515-R2, "BWR Water Chemistry Guidelines - 2000 Revision," February 2000.
For NMP2, the significant change in water chemistry control when HWC is in operation is the addition of hydrogen-to-oxygen molar ratio monitoring, as an indirect means of determining the electrochemical potential. NMP2 follows the guidelines of EPRI TR-103515-R2, "BWR Water Chemistry Guidelines - 2000 Revision," February 2000, for the monitoring and evaluation of reactor water chemistry parameters.
More details regarding the NMP1 and NMP2 Water Chemistry Control Program are available in LRA Section B2.1.2 and NMPNS supplemental letter NMP1 L 1880, dated October 29, 2004.
RAI 3.1.2.C.4-8 In Table 3.1.1-A, under the discussion of item number 3.1.1.A-01 forfeedwater nozzles, the applicant credits the B WR Feedwvater Nozzle Program (Appendix B2.1.5) in addition to the Fatigue Monitoring Program because an enhanced inservice inspection program of NUREG-0619 was implemented at NMPI. Th7e B WI? Feedwater Nozzle Program is based upon an enhanced inservice inspection in accordance with the ASME Code, Section A7, 1989 Edition.
Howvever, in reviewing the program in Appendix B2.1.5, the staff did not find any enhancement to the inservice inspection. Please provide additional information that would resolve this apparent inconsistency.
Response
The NMP BWR Feedwater Nozzle Program, as described in LRA Section B2.1.5, does not require enhancement to be consistent with the guidance of NUREG-1801. This program implements enhanced inservice inspections in accordance with the requirements of ASME Section XI, Subsection IWB, and the recommendations of General Electric (GE) NE-523-A71-0594, Alternate BWR Feedwater Nozzle Inspection Requirements. The statement referenced in LRA Table 3.1.1.A, Item 3.1.1.A-01, as an "enhanced inservice inspection," was meant to indicate that the inservice inspection technique is "enhanced." It was not meant to imply that the BWR Feedwater Nozzle Program at NMP required an enhancement. Therefore, the program description in LRA Section B2.1.5 is correct as submitted and there is no inconsistency between the program description and LRA Table 3.1.1LA, Item 3.1.l.A-01.
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I FV Note that LRA Table 3.1.1.A, Item 3.1.1.A-01, was revised byNMPNS letter NMPIL 1892, dated December 6, 2004, to delete the write-up in the Discussion column that contained the statement referenced in the RAI. The aging management programs credited for the NMP1 Feedwater nozzles have not changed. However, the write-up could be interpreted to imply that the BWR Feedwater Nozzle Program was also credited for managing the aging effect of cumulative fatigue damage, which it does not. Therefore, it was deleted.
RAI 3.1.2.C.4-9 In LRA Tables 3.1.2.A-2 and 3.1.2.B-2 for orificedfuel support, the applicant states that there is no aging effect for the cast stainless steel fuel support in the treated water or steam environment at temperature exceeding 482 degrees F and, therefore, has no aging management. This component is identified in the GALL tinder IVBJ.5.I to be potentially susceptible to loss of fracture toughness due to thermal aging and neutron irradiation embrittlemient. Furthermore, ctumulative fatigue damage is also likely, thus requiring a TLAA for the period of extended operation. Please provide additional information that justify the rationale for no aging management of this component.
Response
Subsequent to submittal of the original LRA, dated May 26, 2004, NMPNS submitted a revised LRA Section 3.1, "Aging Management of Reactor Vessel, Internals, and Reactor Coolant Systems," in letter NMP1L 1892, dated December 6, 2004. Included in this new Section 3.1 were revisions to the aging effects and aging management programs applicable to the NMP 1 and NMP2 Orificed Fuel Supports (OFS). In the original LRA, NMP stated that there were no applicable aging effects or aging management programs related to these components (Ref. LRA Table 3.1.1.A, Item 3.1.1.A-33; Table 3.1.1.B, Item 3.1.1.B-33; Table 3.1.2.A-2; and Table 3.1.2.B-2). In the revised LRA Section 3.1, NMP states that, for the NMP1 and NMP2 OFS components, the applicable aging effect requiring management is Loss of Fracture Toughness and the aging management program is the BWR Vessel Internals Program. These changes were also determined to be consistent with the guidance of NUREG-1801, Item IV.B1.5-a, except that a different aging management program is credited. The specific changes can be seen in the revised LRA Table 3.l1.LA, Item 3.l1.I.A-33 (page 3.1-29); Table 3.1.1.B, Item 3.1.1.B-33 (page 3.1-40); Table 3.1.2.A-2 (page 3.1-53); and Table 3.1.2.B-2 (page 3.1-86).
The applicable BWRVIP guidance for the OFS is BWRVIP-47. BWRVIP-47 currently recommends no inspections for the OFS. The BWRVIP position can be summarized as follows:
While the BWRVIP concedes that CASS materials may experience reduced fracture toughness in reactor internals due to thermal embrittlement, neutron embrittlement, or the combined effect of both mechanisms, loss of fracture toughness is not considered significant without a crack.
Cracking is judged to be an aging effect that does not require management due to the high resistance of CASS materials to IGSCC. The only mechanism for producing a crack is impact, which can only occur during refueling or maintenance activities. Any impact causing cracking during such activities would prompt corrective action.
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I t1 BWRVIP-47 does not identify cumulative fatigue damage as being an aging effect requiring management of any lower plenum internals, including the OFS, because the lower plenum internals are not subject to significant thermal cycles or vibrati6n. Since aging effects for NMP internals are consistent with the aging effects determined in the BWRVIP evaluations, fatigue has not been identified as an aging effect. The fatigue TLAA for the reactor vessel internals is discussed in LRA Section 4.3.5. In the evaluation of the fatigue TLAA for the reactor vessel internals, the OFS was not identified as a high fatigue usage location that would require monitoring in the Fatigue Monitoring Program.
The NRC has accepted and endorsed the provisions of BWRVIP-47 for use as the aging management program for the OFS during the period of extended operation in the NRC letter to BWRVIP Chairman, "Acceptance For Referencing of 'BWR Vessel and Internals Project, BWR Lower Plenum Inspection and Flaw Evaluation Guidelines (BWRVIP-47),' for Compliance with the License Renewal Rule (10 CFR Part 54)," dated December 7, 2000.
Additional information regarding the NMPNS commitments to the BWRVIP program, specifically the applicant action items included in the NRC license renewal safety evaluation reports for license renewal, is contained in NMPNS letter NMP1L 1888 dated November 19, 2004.
RAT 3.1.2.C.4-10 In LRA Tables 3.1.2.A-1 and 3.1.2.B-1, the applicant states that thermal sleeves serve a pressure boundaryfiunction. However, the staff is not clear regarding the design of the thermal sleeve as a pressure boundary. Please provide additional information regarding the design of thermal sleeves and its pressure boundaryffunction at both NMPJ and NMP2.
Response
Based upon further review of the design drawings for the nozzles of the relevant NMP1 and NMP2 systems, it has been concluded that the thermal sleeves do not perform a "Pressure Boundary" intended function. Their only intended function is "Thermal Shielding." The relevant NMP1 systems addressed by LRA Table 3.1.2.A-1 (page 3.1-45) are Core Spray, Control Rod Drive, and Feedwater. The relevant NMP2 systems addressed by LRA Table 3.1.2.B-1 (page 3.1-71) are Control Rod Drive, Core Spray, Reactor Recirculation, Residual Heat Removal, and Feedwater. The LRA tables will be revised accordingly.
LRA Revisions LRA Tables 2.3.1.A.1-1 (page 2.3-3) and 2.3.1.B.l-1 (page 2.3-17) are revised to change the Intended Function for the component type "Thermal Sleeves" and "Nozzle Thermal Sleeves,"
respectively, from "Pressure Boundary" to "Thermal Shielding," as follows:
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irr Table 2.3.1.A.1-1 NMP1 Reactor Pressure Vessel Component Type Intended Functions Thermal Sleeves l
-rFalshid igl Table 2.3.1.B.1-1 NMP2 Reactor Pressure Vessel Component Type Intended Functions Nozzle Thermal Sleeves lteml Shielding LRA Tables 3.1.2.A-1 (page 3.1-45) and 3.1.2.B-1 (page 3.1-71) are revised to change the Intended Function for the component type "Thermal Sleeves" and "Nozzle Thermal Sleeves,"
respectively, from "PB" (Pressure Boundary) to "TS" (Thermal Shielding), as shown on the following pages.
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Table 3.1.2.A-1 Reactor Vessel, Internals, and Reactor Coolant System NMP1 Reactor Pressure Vessel - Summary of Agin Management Evaluation Copnn neddAging Effect NUREG-Compone Functiontl Material Environment Requiring Aging Management 1801 Table I Notes TpFucinManagement Program Volume 2 Item Noe
_ _ _ _ _Ite m
Thermal Sleeves TS Nickel Based Treated Water or Cumulative Fatigue Monitoring IV.B13.4-b 3.1.1.A-01 C 6 Alloys Steam, High Fatigue Program Temperature -
Damage BWR Reactor Pressure Vessel Cracking ASME Section XI IV.B1.1-e 3.1.1.A-32: e, 6 Inservice Inspection (Subsections IWB.
IWC. IWD) Program Water Chemistry Control Proqram Carbon or Low Treated Water or Cracking Water Chemistry 0Q 68 Alloy Steel Steam, High Control Program (Yield Strength Temperature-Cumulative Fatigue Monitoring IV.A1.3-d 3.1.1.A-01 C, 68
< 100 Ksi)
BWR Reactor Fatigue Program (Clad with Pressure Vessel Damage Stainless Steel)
Wrought Treated Water or Cumulative Fatigue Monitoring IV.1BI.3-b 3.1.1.A-01 A
Austenitic Steam, High Fatigue Program Stainless Steel Temperature -
Damage BWR Reactor Cracking BWR Vessel IV.B13.3-a 3.1.1.A-31 B
Pressure Vessel Internals Program Water Chemistry Control Program Page 20 of 21
4a Table 3.1.2.B-1 Reactor Vessel, Internals, and Reactor Coolant System NMP2 Reactor Pressure Vessel - Summary of AginC Management Evaluation Cmoet ItneMaeal Evrnet Aging Effect NUREG-Tympol Intended l
Material l
Environment lRequiring Aging Management 1801 Table I Notes Type Function Management Program Volume 2 Item
_ _ _ _Ite m
Nozzle Thermal TS Nickel Based Treated Water or Cracking BWR Vessel IV.A1.4-a 3.1.1.B-31 E, 31 Sleeves Alloys Steam, High Internals Program Temperature -
BWR Reactor Water Chemistry Pressure Vessel Control Proaram Water Chemistry IV.A1.4-a 3.1.1.B-31 E, Q, Control Program 58 Cumulative Fatigue Monitoring IV.A1.4-b 3.1.1.B-01 C, 31, Fatigue Program 58 Damage Wrought Treated Water or Cracking BWR Vessel IV.A1.4-a 3.1.1.B-31 D, 59 Austenitic Steam, High Internals Program Stainless Steel Temperature -
BWR Reactor Water Chemistry Pressure Vessel Control Program Water Chemistry IV.A1.4-a 3.1.11.B-31 E, Q, Control Program 60 Cumulative Fatigue Monitoring IV.A1.4-b 3.1.1.B-01 C, 59, Fatigue Program 60 Damage Page 21 of 21
ATTACHMENT 2 List of Regulatorv Commitments The following table identifies those actions committed to by Nine Mile Point Nuclear Station, LLC (NMPNS) in this submittal. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.
REGULATORY COMMITMENT DUE DATE Submit a LRA supplemental letter, revising LRA Section 3.1 February 28, 2005 tables to replace the ASME Section XI Inservice Inspection Program with the Bolting Integrity Program for the NMP 1 and NMP2 Reactor Recirculation System closure bolting.
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