ML042740439
| ML042740439 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 09/21/2004 |
| From: | Jensen J Indiana Michigan Power Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| AEP:NRC:4034-16, TAC MC1202, TAC MC1203 | |
| Download: ML042740439 (19) | |
Text
Indiana Michigan Power Company 500 Circle Drive Buchanan, MI 49107 1395 INDIANA MICHIGAN POWER September 21, 2004 AEP:NRC:4034-16 10 CFR 54 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop O-PI-17 Washington, DC 20555-0001
SUBJECT:
Donald C. Cook Nuclear Plant, Units 1 and 2 Docket Nos. 50-315 and 50-316 License Renewal Application -
Supplemental Responses to Requests for Additional Information (TAC Nos. MC 1202 and MC 1203)
Dear Sir or Madam:
By letter dated October 31, 2003, Indiana Michigan Power Company (I&M) submitted an application to renew the operating licenses for Donald C. Cook Nuclear Plant, Units 1 and 2 (Reference 1).
During the conduct of its review, the Nuclear Regulatory Commission (NRC)
Staff has identified areas where additional information is needed to complete its review of the license renewal application (LRA), and issued requests for additional information (RAIs) to obtain the needed information. In some cases, the NRC Staff determined that the information provided in I&M's response to the RAIs did not entirely satisfy the NRC Staff's information needs, and additional clarification was requested via telephone conference calls or meetings.
This letter provides I&M's response to one new request (RAI 3.1.2-4) and supplements I&M's original responses to RAIs provided in the following letters:
The enclosure to this letter provides an affirmation pertaining to the statements made in this letter. Attachment I to this letter provides I&M's responses to the NRC Staff's RAI and supplemental clarifications to RAIs.
provides a list of regulatory commitments made in this submittal.
U S. Nuclear Regulatory Commission Page 2 AEP:NRC:4034-1 6 Should you have any questions, please contact Mr. Richard J. Grumbir, Project Manager, License Renewal, at (269) 697-5141.
Sincerely, Site Vice President NH/rdw
Enclosure:
Affirmation Attachments: 1. Response to Requests for Additional Information for the Donald C. Cook Nuclear Plant License Renewal Application
- 2. List of Regulatory Commitments
U S. Nuclear Regulatory Commission AEP:NRC:4034-16 Page 3
References:
- 1. Letter from M. K. Nazar, I&M, to NRC Document Control Desk, "Donald C.
Cook Nuclear Plant Units 1 and 2, Application for Renewed Operating Licenses,"
AEP:NRC:3034, dated October 31, 2003
[Accession No. ML033070177].
- 2. Letter from M. K. Nazar, I&M, to NRC Document Control Desk, "Donald C.
Cook Nuclear Plant, Units I and 2, Docket Nos. 50-315 and 50-316, License Renewal Application - Response to Requests for Additional Information on Scoping and Screening Results," AEP:NRC:4034-01, dated May7,2004
[Accession No. ML041390360].
- 3. Letter from M. K. Nazar, I&M, to NRC Document Control Desk, "Donald C.
Cook Nuclear Plant, Units 1 and 2, License Renewal Application - Response to Requests for Additional Information on Engineered Safety Features, Auxiliary Systems, and Steam and Power Conversion Systems (TAC Nos.
MC1202 and MC1203)," AEP:NRC:4034-09, dated June 30, 2004 [Accession No. ML041890378].
- 4. Letter from M. K. Nazar, I&M, to NRC Document Control Desk, "Donald C.
Cook Nuclear Plant, Units 1 and 2, Docket Nos. 50-315 and 50-316, License Renewal Application - Response to Requests for Additional Information on Structures and Component Supports (TAC Nos. MC1202 and MC1203),"
AEP:NRC:4034-07, dated June 8, 2004 [ML041670523].
- 5. Letter from M. K. Nazar, I&M, to NRC Document Control Desk, "Donald C.
Cook Nuclear Plant, Units 1 and 2, Docket Nos. 50-315 and 50-316, License Renewal Application - Response to Requests for Additional Information on Time-Limited Aging Analyses (TAC Nos. MC 1202 and MC 1203),"
AEP:NRC:4034-08, dated June 16, 2004 [ML041750561].
c:
J. L. Caldwell, NRC Region III K. D. Curry, AEP Ft. Wayne, Nv/o attachments J. T. King, MPSC, wv/o attachments C. F. Lyon, NRC Washington DC MDEQ - WHMD/HWRPS, w/o attachments NRC Resident Inspector J. G. Rowley, NRC Washington DC
Enclosure to AEP:NRC:4034-16 AFFIRMATION I, Joseph N. Jensen, being duly sworn, state that I am Site Vice President of Indiana Michigan Power Company (I&M), that I am authorized to sign and file this request with the Nuclear Regulatory Commission on behalf of I&M, and that the statements made and the matters set forth herein pertaining to I&M are true and correct to the best of my knowledge, information, and belief.
Indiana Michigan Power Company Joseph N. Jensen Site Vice President SWORN TO AND SUBSCRIBED BEFORE ME THIS DAY OF b-f, 2004
"~My Public My Commission Expires Ob/,o/20o0 7
Z
Attachment I to AEP:NRC:4034-16 Page I Response to Requests for Additional Information for the Donald C. Cook Nuclear Plant License Renewal Application By letter dated October 31, 2003, Indiana Michigan Power Company (I&M) submitted an application to renew the operating licenses for Donald C. Cook Nuclear Plant (CNP), Units 1 and 2 (Reference 1). This attachment provides I&M's response to one new Nuclear Regulatory Commission (NRC) Staff request for additional information (RAI), RAI 3.1.2-4, and supplements I&M's original responses to RAIs provided in the following letters:
- License Renewal Application (LRA) Section 2.3.3.7 RAI - Letter dated May 7, 2004 (Reference 2)
References
- 1. Letter from M. K. Nazar, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant Units I and 2, Application for Renewed Operating Licenses," AEP:NRC:3034, dated October 31, 2003 [Accession No. ML033070177].
- 2. Letter from M. K. Nazar, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant, Units 1 and 2, Docket Nos. 50-315 and 50-316, License Renewal Application -
Response to Requests for Additional Information on Scoping and Screening Results,"
AEP:NRC:4034-01, dated May 7, 2004 [Accession No. ML041390360].
- 3. Letter from M. K. Nazar, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant, Units I and 2, License Renewal Application - Response to Requests for Additional Information on Engineered Safety Features, Auxiliary Systems, and Steam and Power Conversion Systems (TAC Nos. MC1202 and MC1203)," AEP:NRC:4034-09, dated June 30,2004 [Accession No. ML041890378].
- 4. Letter from M. K. Nazar, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant, Units I and 2, Docket Nos. 50-315 and 50-316, License Renewal Application -
Response to Requests for Additional Information on Structures and Component Supports (TAC Nos.
AEP:NRC:4034-07, dated June 8,
2004
[ML041670523].
- 5. Letter from M. K. Nazar, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant, Units I and 2, Docket Nos. 50-315 and 50-316, License Renewal Application -
Response to Requests for Additional Information on Time-Limited Aging Analyses (TAC Nos.
AEP:NRC:4034-08, dated June 16, 2004
[ML041750561].
Attachment I to AEP:NRC:4034-16 Page 2 New Request for Additional Information RAI 3.1.2-4:
In CNP LRA Section 3.1.2.2.6 and Table 3.1.1-11, the applicant credits CNP AMP [aging management program] B. 1.27, "Reactor Vessel Internals Plates, Forgings, Welds, and Bolting Program, " to manage changes in dimension due to void svelling of reactor vessel internals.
In CNP LRA Table 3.1.2-2, Page 3.1-51, the applicant credits the Reactor Vessel Internals (CASS) Program to manage distortion of the lowver support plate and lower core plate support column cap in a treated water environment. Additionally, in LRA Appendix B, Sections B.1.27, "Reactor Vessel Internals Plates, Forgings, Welds, and Bolting," and B.1.28, "Reactor Vessel Internals Cast A ustenitic Stainless Steel, " the applicant describes the program description for each as managing the aging effect of distortion due to void swelling.
The staff requests the applicant to update LRA Section 3.1.2.2.6 and Table 3.1.1 item number 11 to include the Reactor Vessel Internals (CASS) Program with managing this aging effect or justify why it should not be credited with managing this aging effect.
I&M Response to RAI 3.1.2-4:
As described in LRA Section B.1.28, the Reactor Vessel Internals Cast Austenitic Stainless Steel Program will manage change in dimension by void swelling of reactor vessel internals cast austenitic stainless steel (CASS) components. This program should have been included in LRA Section 3.1.2.2.6 and Table 3.1.1, Item Number 3.1.1-11. LRA Table 3.1.2-2 correctly identifies the CASS components that are susceptible to change in dimension by void swelling (distortion) and credits the Reactor Vessel Internals Cast Austenitic Stainless Steel Program with managing this aging effect.
LRA Table 3.1.1, Item Number 3.1.1-37, also credits this program for management of void swelling.
Supplemental Reponses to Requests for Additional Information RAI 2.3.3.7-1:
The license reneuial boundary drawings referenced in Section 2.3.3.7 did not identify the followving fire protection (FP) systems and components as being within the scope of license renewal and subject to an aging management review (AMR).
The staff believes that the FP systems and components described below are passive, long-lived, and perform a function that demonstrates compliance with 10 CFR 50.48forfire protection. Provide basisfor excluding the following FP systems and components from the scope of license renewal and subject to an AMR:
Attachment I to AEP NRC:4034 16 Page 3 (1) LRA Drauing LRA-12-5152D Fire Protection Water - Aulx. and Containment Buildinzs A note at location D-6 states details on a deluge valve are found on DWTVG. 5152M, which weas not included in the LRA and should be subject to an AMR. Clarify whether the deluge valve should be in scope orjustilf its exclusion.
(Note - the original RAI included 12 sub-parts. Only sub-part (1), which was the subject of the requested clarification, is listed.)
Clarification Requested for RAI 2.3.3.7-1:
The staff indicated that the applicant states that deluge valves serve no license renewvalfiunction.
However, the applicant has not discussed how the passive components of the valve are not subject to an AMR, given the related fire protection system protects a safety related area.
Therefore, the staff is not able to determine that, in the event of afire, the integrity of the reactor coolant pump suppression system will be ensured.
The staffs concern described in RAI 2.3.3.7-1 (1) remains an open issue.
I&M's Supplemental Response to RAI 2.3.3.7-1:
As stated in the original response to RAI 2.3.3.7-1(1) provided in I&M's RAI response letter dated May 7, 2004 (Reference 1), the reactor coolant pump (RCP) suppression system deluge valves are not required by 10 CFR 50, Appendix R, or 10 CFR 50.48, and consequently do not serve a license renewal function. The nonessential service water system, rather than the fire protection system, supplies water to these deluge valves. Because the water supply for these deluge valves does not originate from the in-scope fire protection system piping, which is depicted on license renewal drawing LRA-12-51 52D, a passive failure of these valves would not prevent the fire protection system from supplying fire water to those portions of the system that are required by 10 CFR 50.48.
Section 111.O of Appendix R to 10 CFR 50 requires only a lubricating oil collection system for the protection of the RCP area.
10 CFR 50, Appendix R does not require a fire protection sprinkler system for the protection of this area. As approved in the Safety Evaluation Report for the CNP exemption to the 10 CFR 50, Appendix R, requirements applicable to the design of the CNP RCP motor lube oil collection system (Reference 2), the existing RCP lube oil collection system provides a level of safety equivalent to the technical requirements of 10 CFR 50, Appendix R, Section III.O.
Therefore, the passive components of the deluge valves do not serve a license renewal intended function, and consequently, are not subject to aging management review.
Attachment I to AEP NRC:4034-16 Page 4 References for Supplemental Response to RAI 2.3.3.7-1
- 1. Letter from M. K. Nazar, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant, Units I and 2, Docket Nos. 50-315 and 50-316, License Renewal Application -
Response to Requests for Additional Information on Scoping and Screening Results,"
AEP:NRC:4034-01, dated May 7,2004 [Accession No. ML041390360].
- 2. Letter from D. G. Eisenhut, NRC, to J. Dolan, I&M, "Donald C. Cook Nuclear Power Plant, Unit Nos. I and 2, Fire Protection - Request for Exemption from Requirements of Appendix R to 10 CFR 50, Sections III.G and III.O,"
dated December 23, 1983
[Accession No. ML021020051].
RAI 3.2-1:
LRA Section 3.2.2.2.1 identifies the applicant's aging management for cumulative fatigue damage for components in the ESF [engineered safety featuires] systems. In the discussion the LRA refers to Section 4.3 which states that based on a screening criteria, the applicant determined that components in the ECCS [emergency core cooling systems] exceeded the screening criteria. The piping components that exceeded the screening criteria were evaluated by the applicant for their potential to exceed 7000 thermal cycles in sixty years of plant operation.
The applicant determined that none of the piping components in the EFS [sic] system exceeded 7000 cycles during the period of extended operation. The applicant is requested to provide the highest estimated number of thermal cycles and the basis for derivation for each component type identified in Tables 3.2.2-1, -2, -3, and -4 of the LRA for which TLAA [time-limited aging analysis] - Metal Fatigue has been designated as the aging management program. For those components whose material or aging effect is not specified in NUREG-1801 (designated as 'F' and 'I respectively in the notes), clarify whether or not the applicant performs the thermal cycle evaluation in accordance with NUREG-1801, Section 4.3-1.12. If so, is the applicants TLAA program consistent with NUREG-1801. If not explain any differences. Also the applicant is requested to address how unanticipated transients and thermal stratification are accountedfor in the estimation Clarification Requested by the Staff:
In its response to the RAI the applicant states that unanticipated transients are not accountedfor in its methodology. If they occur, they are identified and evaluatedfor impact on design thermal and loading cycles through the corrective actions program.
The applicant is requested to clarify holw unanticipated transients wtill be identified For example, clarify whether or not monitoring of transients is done to identify the occurrence of to AEP NRC:4034-16 Page 5 unanticipated transients. The applicant should also clarify if any ESF class 1 piping is included in ESF LRA tables or ifESF class I piping is considered within scope of the RCS.
I&M's Supplemental Response to RAI 3.2-1:
The Fatigue Monitoring Program monitors, records, and reviews the plant transients identified in Updated Final Safety Analysis Report (UFSAR) Section 4.1.4 to ensure that these transients are maintained within the limits provided in UFSAR Table 4.1-13. CNP personnel continuously monitor plant operation.
Plant procedures require recording and reporting of transient information upon notification/discovery of unanticipated transients, and evaluation of transient information. The need for corrective action is based on the significance of an unanticipated transient. Evaluation under the corrective action program will determine the potential impact of the specific plant transient on the fatigue life of affected components.
The LRA Section 3 ESF aging management review results tables do not include any Class 1 ESF piping.
The reactor coolant system (RCS) Class 1 piping evaluation boundary extends into portions of ancillary systems, including ESF systems, attached to the RCS. Therefore, Class I ESE piping is part of the RCS pressure boundary evaluated in LRA Sections 2.3.1.4 and 3.1.2.1.3 and associated aging management review results tables. Those portions of ESF systems included in the RCS pressure boundary are depicted as part of the RCS on the specific ESF license renewal drawings.
RAI 3.24:
LRA Table 3.2.2-2 credits the Boric Acid Corrosion Prevention Program for managing loss of mechanical closure integrity for carbon steel bolts in an external air environment. This aging management program relies on implementation of recommendations in NRC Generic Letter (GL) 88-05 "Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in PIWR Plants." Since this program addresses components inside the containment, the applicant is requested to discuss the managementfor the loss of mechanical closure integrity of carbon steel bolts outside the containment.
Clarification Requested by the Staff:
In its response the applicant states that the Boric Acid Corrosion Program manages loss of mechanical closure integrity of carbon steel bolts both inside and outside containment.
The applicant is requested to update the AMP so that the scope goes beyond GALL [Generic Aging Lessons Learned] and covers other leaks besides those from the reactor coolant pressure boundary.
Attachment I to AEP NRC:4034-16 Page 6 I&M's Supplemental Response to RAI 3.24:
I&M's RAI response letter dated August 19, 2004 (Reference 1) included a response to RAI B. 1.4-1, which provides the requested clarification. The response states that the Boric Acid Corrosion Prevention Program applies to portions of systems and structures, both inside and outside containment, that are subject to aging management review and contain borated water or are subject to exposure to leaking borated water.
This information is based on the current program scope, which applies to all carbon steel and low alloy steel components and material in all systems that contain or have the potential to come in contact with borated water; thus no changes to this aging management program are required. RAI responses, including this response to RAI 3.2-4 and the referenced response to RAI B.1.4-1, supplement the information provided in the LRA; therefore, no changes to the LRA description of the Boric Acid Corrosion Prevention Program are required.
Reference for Response to RAI 3.2-4
- 1. Letter from J. N. Jensen, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant, Units 1 and 2, License Renewal Application - Response to Requests for Additional Information on Aging Management Programs," AEP:NRC:4034-13, dated August 19, 2004
[Accession No. ML042390469].
RAI 3.2-6:
LRA Table 3.2.2-1 identifies a plant specific In-service Inspection Program for managing the aging effect due to cracking and loss of material of stainless steel thermowells and valves in a sodium hydroxide environment. This combination of environment, material and component is not evaluated in the GALL report. The applicant is requested to discuss the plant specific inspection methods including frequency of inspections and acceptance criteria. Also identify the differences with the appropriate ASME [American Society of Mechanical Engineers] Section XI requirements if any, andprovidejuistiflicationfor the differences.
Clarification Requested by the Staff:
The applicant 's response states that the frequency of the inspections will be 10 years. The.
applicant is requested to clarify the basis for this frequency.
If the frequency is based on operating experience the applicant should provide relevant operating experience to justify this inspection interval.
I&M's Supplemental Response to RAI 3.2-6:
I&M's response to RAI 3.2-6, as provided in the referenced RAI response letter, dated June 30, 2004, states that the 10-year inspection frequency is based on providing consistency
Attachment I to AEP NRC:4034-16 Page 7 with ASME Section XI, Subsection IWC, requirements for comparable Class 2 components.
The stainless steel thermowells and valves in the containment spray system are not ASME components; thus it is conservative to apply ASME requirements to these components.
Operating experience for these components has not indicated degraded conditions that would warrant an increased inspection frequency.
Reference for Supplemental Response to RAI 3.2-6 Letter from M. K. Nazar, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant, Units 1 and 2, License Renewal Application - Response to Requests for Additional Information on Engineered Safety Features, Auxiliary Systems, and Steam and Power Conversion Systems (TAC Nos. MC1202 and MC1203)," AEP:NRC:4034-09, dated June 30,2004 [Accession No. ML041890378].
RAI 3.2-7:
LRA Table 3.2.2-1 identifies a plant specific In-service Inspection Program for managing the aging effect due to cracking and loss of material in stainless steel tanks in an internal sodium hydroxide environment. Neither this component nor the material and environment are evaluated in the GALL report. The applicant is requested to discuss its plant specific inspection methods including frequency of inspections and acceptance criteria. Also identify the difference with the appropriate ASME XI requirements ifany, andprovide justiJfication for the same.
Clarification Requested by the Staff:
The applicant's response states that the frequency of the inspections will be 10 years. The applicant is requested to clarify the basis for this frequency.
If the frequency is based on operating experience the applicant should provide relevant operating experience to justify this inspection interval.
I&M's Supplemental Response to RAI 3.2-7:
As stated in I&M's supplemental response to RAI 3.2-6, the basis for the 10-year inspection frequency is that it is consistent with ASME Section Xl, Subsection IWC, requirements for comparable Class 2 components.
These stainless steel spray additive tanks are not ASME components; thus it is conservative to apply ASME requirements to these components.
Operating experience for these components has not indicated degraded conditions that would warrant an increased frequency.
Attachment I to AEP NRC:4034 16 PPage 8 RAI 3.2-10:
The GALL report recommends a plant-specific aging management program for loss of material due to general, pitting, and crevice corrosion and microbiologically induced corrosion (MIC) in carbon steel components exposed to lubricating oil that may be contaminated with water.
Similar aging effects (except general corrosion) are possible for copper alloy. The NRC staff considers a periodic inspection program appropriate to manage this aging effect. For the oil cooler shell in the emergency core cooling system (LRA Table 3.2.2-3) exposed to an oil environment, the applicant is requested to provide a periodic inspection program in addition to an oil analysis program for aging management for loss of material diue to general (carbon steel),
pitting, and crevice corrosion and MIC, or provide justification for not managing this aging effect.
Clarification Rcqucsted by the Staff:
The response states that loss of material is not an aging effect requiring management for surfaces exposed to lubricating oil unless moisture or contaminants are present. The staff position identified in GALL XI.M32 is that, where aging effects are not expected to occur but.
there is insufficient data to completely rule it out, a one-time inspection is acceptable to confirm that aging effects are not occurring. An alternate acceptable program may include routine maintenance or a review of repair records. The applicant is requested to clarify how the effectiveness of the oil analysis program is assured I&M's Supplemental Response to RAI 3.2-10:
NUREG-1801,Section XI.M32, identifies that a one-time inspection is needed "to address concerns for the potential long incubation period for certain aging effects on structures or components". This inspection is required when "an aging effect is not expected to occur but there is insufficient data to completely rule it out." As described in LRA Section B.1.23, the Oil Analysis Program periodically samples the lubricating oil.
Compliance with recommended levels of water and other contaminants in the lubricating oil provides assurance that these contaminants do not exceed acceptable levels. If water or other contaminants are not present, the aging effect of loss of material is prevented. There are no aging mechanisms with a potential long incubation period that could result in a loss of material if water or contaminants are not present. Oil analysis sampling provides sufficient data to assure the effectiveness of the program in preventing aging effects for components in an oil environment.
Based on CNP operating experience, there is reasonable assurance that the Oil Analysis Program will continue to manage the aging effects of components exposed to lubricating oil. This position is consistent with previously approved NRC Staff positions documented in NUREG 1743, Safety Evaluation Report Related to the License Renewal of Arkansas Nuclear One, Unit I [ML011640177] and NUREG 1769, Safety Evaluation Report Related to the License Renewal of Peach Bottom Atomic Powter Station, Units 2 and 3 [ML031010053].
Attachment I to AEP NRC:4034-16 Page 9 RAI 3.2-11:
LRA Table 3.2.2-3 states that the copper alloy oil cooler tubes for the pump in a cooling wvater environment will be managed for loss of material using the Water Chemistry Control Program.
For this material type and environment, the staff considers selective leaching to be an aging effect requiring management.
The applicant is requested whether selective leaching is considered to be an aging mechanism for the tubes. Ifso, describe the types of inspections used by the applicant to detect selective leaching in the tubes.
Clarification Requested by the Staff:
The applicant is requested to clarify whether or not hardness measurement will be done to manage selective leaching, as required by GALL XI.M33. If not, provide justification as to why hardness measurement is not required.
I&M's Supplemental Response to RAI 3.2-11:
The Heat Exchanger Monitoring Program will include activities to manage loss of material due to selective leaching.
Brinell Hardness testing will be performed on selected heat exchanger tubes that are susceptible to selective leaching, when feasible.
However, Brinell Hardness testing may not be feasible for some components due to form and configuration (e.g., heat exchanger tubes). In such cases, examinations other than Brinell Hardness testing may be used to identify the presence of selective leaching of material. Other mechanical means, such as scraping or chipping), xvil provide an acceptable alternative method of identifying the presence of selective leaching.
RAI 3.4-4:
The AMP 1.2 Bolting and Torquing Activities, an existing plant specific program is creditedfor managing loss of mechanical closure integrity. The program covers bolting in high temperature systems and in applications subject to significant vibration. The staff notes that NUREG-1801 credits AMP XI.M 18 Bolting Integrity for monitoring loss of material, cracking, and loss of preload. In addition, accepted bolting integrity programs (such as EPRI [Electric Powser Research Institute] 104213) recommend monitoringfor loss ofpreload as one of the parameters monitored/inspected. Monitoring for cracking of high strength bolts (actual yield strength equal or greater than IS0 ksi) is also recommended.
Attachment I to AEP NRC:4034-16 Page 1I0 As such, the applicant is requested to provide the following information:
- a. Identify the areas of the Bolting Integrity Program at D.C. Cook which are consistent with the AMP XI.M. 18 in the GALL report, and also those aspects in which it is different.
- b. Discuss how the loss of preload aging effect would be managed by the Bolting and TorquingActivities AMP at D.C. Cook
- c. Discuss the inspections associated with the Bolting and Torquing Activities AMP at D. C. Cook ivhich may be beyond the requirements ofASME Section XI.
- d. Are there any high strength bolts included within the boundary of these systems (Engineered Safety Features and Steam & Power Conversion Systems)?
- e. The occurrence of SCC [stress corrosion cracking] in stainless steel bolts can depend on a combination of factors such as stainless steel grade, method of hardening (for example, strain, precipitation or age hardening) environment and stress levels. Discuss how these factors Here taken into account to determine whether or not SCC is an applicable aging effect.
Clarification Requested by the Staff:
The applicant's response covers the Engineered Safety Features and the Steam and Powver Conversion Systems. The applicant is requested to indicate whether or not the present response would change if the Auxiliary System wvas also included in the original RAI. Also discuss those changes if any. In addition the applicant is requested to indicate why loss of preload is not considered to be an aging mechanism.
I&M's Supplemental Response to RAI 3.4-4:
The Bolting and Torquing Activities Program manages loss of mechanical closure integrity for both stainless steel and carbon steel bolting subjected to high temperature or significant vibration in the Auxiliary Systems. Although the original I&M response to RAI 3.4-4 provided in the referenced June 30, 2004, letter does not include carbon steel bolting, the attributes discussed in the response would not change if the Auxiliary Systems were also included in the original RAI.
Loss of preload due to long-term exposure to elevated temperatures or significant vibration, such as that due to a diesel engine, is an aging mechanism included in the loss of mechanical closure integrity aging effect for carbon steel and stainless steel bolting identified in LRA Tables 3.3-7, 3.3.2-8, 3.3.2-9, and 3.3.2-1.
Attachment I to AEP NRC:4034-16 Page I11 Reference for Supplemental Response to RAI 3.4-4 Letter from M. K. Nazar, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant, Units I and 2, License Renewal Application - Response to Requests for Additional Information on Engineered Safety Features, Auxiliary Systems, and Steam and Power Conversion Systems," AEP:NRC:4034-09, dated June 30, 2004 [Accession No. ML041890378].
RAI 3.5-1:
In line item 3.5.1-3 of Table 3.5.1 of the LRA, the applicant indicates that the aging effects related to loss of material dule to corrosion of bellows, and dissimilar metal welds are managed consistent with NUREG-1801.
NUREG-1801 recommends the examination of penetration bellows and the associated dissimilar welds based on the operating experience with the stress corrosion cracking of bellows as documented in NRC Information Notice 92-20. The applicant is requested to provide the following information related to the examination and/or testing of containment penetration bellowis:
Howv many penetration bellows are in the Cook containments? Please summarize the operating experience related to the examination of these bellows. If provisions are made to assess their leaktightness (as they are not accessible for visual examination), please provide a summary of these provisions (including frequency of tests), and indicate if such leaktightness assessment of the bellows is part of the LRRA AMP [aging management program] B. 1.5 or B 1.8.
Clarification Requested by the Staff:
The I&M response to RAI 3.5-1 indicates that the Cook containments do not have any containment pressure-boundary bellows. Please confirm.
I&M's Supplemental Response to RAI 3.5-1:
Containment penetration bellows do not provide a pressure-boundary function and are not part of the CNP containment isolation barrier.
The UFSAR excerpts provided in I&M's original response to RAI 3.5-1 in l&M's letter dated June 8, 2004, (Reference 1) provide the current licensing basis that reflects this statement. This basis has also been previously reported to the NRC Staff in l&M's license amendment request dated April 11, 2002 (Reference 2), as documented in the NRC safety evaluation that approved that license amendment, dated February 25, 2003 (Reference 3).
For clarification, the component type "Fuel transfer tube penetration" listed in LRA Table3.5.2-1 includes only the mechanical components (tube and closure flange) that are depicted on license renewal drawings LRA-1-5140 and LRA-2-5140 at locations L8 and L4, respectively. The bellows expansion joints that were provided on the outer pipe to compensate
Attachment I to AEP NRC:4034-16 Page 12 for any differential movement between the inner and outer pipes and also between the containment and auxiliary building structures do not serve as part of the containment pressure boundary.
Because penetration bellows do not provide a containment pressure-boundary function, they are not included in the scope of license renewal and are not subject to aging management.
References for Supplemental Response to RAI 3.5-1
- 1. Letter from M. K. Nazar, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant, Units I and 2, Docket Nos. 50-315 and 50-316, License Renewal Application -
Response to Requests for Additional Information on Structures and Component Supports (TAC Nos.
AEP:NRC:4034-07, dated June 8,
2004
[ML041670523].
- 2. Letter from J. E. Pollock, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant, Units I and 2, Docket Nos. 50-315 and 50-316, License Amendment Request for One-Time Extension of Containment Integrated Leakage Rate Test Interval,"
AEP:NRC:2612, dated April 11,2002 [ML021070279].
- 3. Letter from J. F. Stang, NRC, to A. C. Bakken, I&M, "Donald C. Cook Nuclear Plant, Units 1 and 2, Issuance of Amendments (TAC Nos. MB4837 and MB4838)," dated February 25, 2003 [ML030160330].
RAI 3.5-2:
For seals and gaskets related to containment penetrations, in Item Number 3.5.1-6 of the LRA, containment ISI [inservice inspection] and containment leak rate testing have been stated as the aging management programs. For equipment hatches and air-locks at Cook, the staff agrees with the applicant's assertion that the leak rate testing program will monitor aging degradation of their seals and gaskets, as they are leak rate tested after each closing. For other penetrations with seals and gaskets, the applicant is requested to provide information regarding the adequacy of Type B leak rate testing frequency to monitor aging degradation of seals and gaskets at Cook.
Clarification Requested by the Staff:
Item II A3.3-a (Page # II A3-5) of NUREG-1801 addresses the aging management of seals, gaskets, etc. which are part of the containment pressure boundary, i.e. seals and gaskets of equipment hatches, and mechanical & electrical penetrations. If l&Mplans to monitor only Air lock seals by means of containment leakage rate testing, then the applicant is requested to provide additional information regarding the aging management of seals and gaskets of equipment hatches, and mechanical and electrical penetration associated with the containment pressure boundary. If I&M credits containment leak rate testing program (Type B test of
Attachment I to AEP NRC:4034-16 Page 13 Appendix J) for aging management of these seals and gaskets, the applicant is requested to provide additional information as requested in RAI 3.5-2.
I&M's Supplemental Response to RAI 3.5-2:
As indicated in LRA Table 3.5.1, the Containment Leakage Rate Testing Program is credited for aging management of seals and gaskets that are part of the containment pressure boundary (i.e., seals and gaskets associated with equipment hatches and with mechanical and electrical penetrations).
In LRA Table 3.5.1, equipment and personnel hatches are included in Item Number 3.5.1-4, and seals and gaskets are included in Item Number 3.5.1-6.
Gaskets associated with containment mechanical penetrations are consumables that are replaced each time the bolted joint is disassembled or on an established frequency. In addition, such penetrations are tested under the Containment Leakage Rate Testing Program as required by 10 CFR 50, Appendix J. As indicated in LRA Table 3.5.2-1, containment electrical penetrations (which include cable feed-through assemblies) are included in the Containment Leakage Rate Testing Program. The effects of aging on seals and gaskets associated with mechanical and electrical penetrations are also managed by the Containment Leakage Rate Testing Program.
LRA Table 3.5.1, Item Number 3.5.1-6, also includes seals and gaskets associated with mechanical and electrical penetrations.
CNP is committed to Option B of 10 CFR 50, Appendix J, for performing containment leakage rate testing. In accordance with Option B, Type B test intervals are limited to 120 months; however, normally testing is performed more frequently than every 120 months. Type B testing of CNP mechanical and electrical penetrations is performed at least once every 120 months.
Component-specific testing frequency is based on the safety significance and historical performance of the penetrations in accordance with Option B of 10 CFR 50, Appendix J.
RAI 4.3.3-1:
Section 4.3.3 of the LRA discusses I&M's evaluation of the impact of the reactor wvater environment on the fatigue life of components. The discussion references the fatigue sensitive component locations for an early vintage Westinghouse plant identified in NUREG/CR-6260, "Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components." The LRA indicates that the design usage factors provided in Table 5-98 of NUREG/CR-6260 wvere usedfor the evaluation of the charging nozzle, safety injection nozzle and RHR [residual heat removal] tee. The design usage factors wvere based on an evaluation of the Turkey Point facility, including a plant specific evaluation of the RHR piping and detailed finite element analyses of the charging and safety injection nozzles. Discuss the applicability of these analyses to the Cookfacility. The discussion should include a comparison of piping sizes and thicknesses, including the design of the thermal sleeves between Cook and Turkey Point.
to AEP NRC:4034-16 Page 14 The discussion should also include a comparison of the number and type of design transients cycles between Cook and Turkey Point.
Clarification Requested by the Staff:
The response does not contain sufficient information for the Staff to conclude that CNP charging and safety injection nozzles are bounded by charging and safety injection analyses in NUREG-6260 for older vintage Westinghouse plants. The level-of-detail in NUREG-6260 made it very difficult to assure the analyses bound the CNP piping configuration. Additional information or alternatively a commitment similar to what wlas made for the RHR nozzles is needed.
I&M's Supplemental Response to RAI 4.3.3-1:
As a supplement to the Fatigue Monitoring Program enhancement committed to in LRA Section B.2.2, I&M will perform one or more of the following activities prior to the period of extended operation for the Class I charging and safety injection nozzles:
(1) Perform a plant-specific fatigue analysis of the Class 1 charging and safety injection nozzles, which includes environmental effects, to ensure that cumulative usage factors are below 1.0; (2) Manage the effects of fatigue of the Class 1 charging and safety injection nozzles by an NRC-approved inspection program (e.g., periodic non-destructive examination of the affected locations at inspection intervals to be determined by a method accepted by the NRC). The inspections are expected to be able to detect cracking due to thermal fatigue prior to loss of function. Replacement or repair will then be implemented such that the intended.
function will be maintained for the period of extended operation; (3) Repair portions of the Class I charging and safety injection nozzles at the affected locations, as necessary to ensure the intended function will be maintained for the period of extended operation; (4) Replace portions of the Class 1 charging and safety injection nozzles at the affected locations, as necessary to ensure the intended function will be maintained for the period of extended operation; (5) Monitor ASME Code activities to use the environmental fatigue methodology approved by the ASME Code Committee and NRC.
to AEP:NRC:4034-16 Page I LIST OF REGULATORY COMMITMENTS The following table summarizes the actions committed to by Indiana Michigan Power Company (I&M) in this document. Any other actions discussed in this submittal represent intended or planned actions by I&M. They are described to the Nuclear Regulatory Commission (NRC) for information and are not regulatory commitments.
Commitment Date
+
I&M's Supnlemental Resnonse to RAI 4.3.3-1 As a supplement to the Fatigue Monitoring Program enhancement committed to in LRA Section B.2.2, I&M will perform one or more of the following activities prior to the period of extended operation for the Class 1 charging and safety injection nozzles:
(1) Perform a plant-specific fatigue analysis of the Class 1 charging and safety injection nozzles, which includes environmental effects, to ensure that cumulative usage factors are below 1.0; (2) Manage the effects of fatigue of the Class I charging and safety injection nozzles by an NRC-approved inspection program (e.g.,
periodic non-destructive examination of the affected locations at inspection intervals to be determined by a method accepted by the NRC). The inspections are expected to be able to detect cracking due to thermal fatigue prior to loss of function. Replacement or repair will then be implemented such that the intended function will be maintained for the period of extended operation; (3) Repair portions of the Class I charging and safety injection nozzles at the affected locations, as necessary to ensure the intended function will be maintained for the period of extended operation; (4) Replace portions of the Class 1 charging and safety injection nozzles at the affected locations, as necessary to ensure the intended function will be maintained for the period of extended operation; (5) Monitor American Society of Mechanical Engineers (ASME) Code activities to use the environmental fatigue methodology approved by the ASME Code Committee and NRC.
Unit 1:
October 25, 2014 Unit 2:
December 23, 2017