ML041730571
| ML041730571 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 06/21/2004 |
| From: | Gregory Suber NRC/NRR/DRIP/RLEP |
| To: | |
| Suber G, NRR/DRIP/RLEP 301-415-1124 | |
| References | |
| TAC MB8402 | |
| Download: ML041730571 (19) | |
Text
June 21, 2004 LICENSEE:
Entergy Operations Inc.
FACILITY:
Arkansas Nuclear Station, Unit 2
SUBJECT:
SUMMARY
OF TELEPHONE CALL HELD ON MAY 20, 2004, BETWEEN THE U.S. NUCLEAR REGULATORY COMMISSION (NRC) STAFF AND ENTERGY OPERATIONS INC., REPRESENTATIVES CONCERNING DRAFT REQUEST FOR ADDITIONAL INFORMATION PERTAINING TO THE ARKANSAS NUCLEAR ONE, UNIT 2 LICENSE RENEWAL APPLICATION (TAC NO. MB8402)
On May 20, 2004, the NRC staff and representatives of the Entergy Operations Inc., held a telephone conference to discuss draft request for additional information (RAI) pertaining to the technical review for the Arkansas Nuclear One, Unit 2 license renewal application (LRA).
The conference call was used to clarify the intent of the staffs questions. On the basis of the discussion, the applicant acknowledged a better understanding of each question. No staff decisions were made during the telephone conference. In some cases, the applicant agreed to provide information for clarification, and the staff revised the format and content of some RAIs. provides a list of the telephone conference call participants. Enclosure 2 contains a listing of the draft RAIs and a brief description of the status of each item. A copy of this summary was provided to the applicant for comment.
/RA/
Gregory F. Suber, Project Manager License Renewal Section A License Renewal and Environmental Impacts Program Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Docket No.: 50-368
Enclosures:
As stated
June 21, 2004 LICENSEE:
Entergy Operations Inc.
FACILITY:
Arkansas Nuclear Station, Unit 2
SUBJECT:
SUMMARY
OF TELEPHONE CALL HELD ON MAY 20, 2004, BETWEEN THE U.S. NUCLEAR REGULATORY COMMISSION (NRC) STAFF AND ENTERGY OPERATIONS INC., REPRESENTATIVES CONCERNING DRAFT REQUEST FOR ADDITIONAL INFORMATION PERTAINING TO THE ARKANSAS NUCLEAR ONE, UNIT 2 LICENSE RENEWAL APPLICATION (TAC NO. MB8402)
On May 20, 2004, the NRC staff and representatives of the Entergy Operations Inc., held a telephone conference to discuss draft request for additional information (RAI) pertaining to the technical review for the Arkansas Nuclear One, Unit 2 license renewal application (LRA).
The conference call was used to clarify the intent of the staffs questions. On the basis of the discussion, the applicant acknowledged a better understanding of each question. No staff decisions were made during the telephone conference. In some cases, the applicant agreed to provide information for clarification, and the staff revised the format and content of some RAIs. provides a list of the telephone conference call participants. Enclosure 2 contains a listing of the draft RAIs and a brief description of the status of each item. A copy of this summary was provided to the applicant for comment.
/RA/
Gregory F. Suber, Project Manager License Renewal Section A License Renewal and Environmental Impacts Program Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Docket No.: 50-368
Enclosures:
As stated C:\\ORPCheckout\\FileNET\\ML041730571.wpd Accession No: ML041730571 OFFICE:
PM:RLEP LA:RLEP SC:RLEP NAME:
G. Suber MJenkins S. Lee DATE:
6/21/04 6/21/04 6/21/04 OFFICIAL RECORD COPY LIST OF PARTICIPANTS TELEPHONE CALLS WITH ENTERGY OPERATIONS INC.
ARKANSAS NUCLEAR ONE, UNIT 2 LICENSE RENEWAL APPLICATION May 20, 2004 Participants Affiliation Natalie Mosher Entergy Ted Ivy Entergy Andy Taylor Entergy Michael Stroud Entergy Garry Young Entergy Mark Rickle Entergy (Framatome Ariba)
Andrea Lee U.S. Nuclear Regulatory Commission (NRC)
George Georgiev NRC Gregory Suber NRC ANO-2 LICENSE RENEWAL APPLICATION REQUEST FOR ADDITIONAL INFORMATION Table 3.1.2-1 Reactor Vessel and CEDM Pressure Boundary RAI 3.1.2-1.1 The staff requests additional information on the applicants aging management reviews (AMRs) for managing cracking in low alloy steel components that are exposed to an external air environment. Aging management strategies for license renewal are somewhat dependent on the specific types of aging mechanisms that can induce age-related degradation, and not specifically on the general classification of the aging effect. For the low alloy steel components in the reactor coolant system, confirm that cracking is an applicable aging effect requiring aging management. Specifically, define which aging mechanism or mechanisms are known to induce cracking in low alloy steel components that are exposed to an external air environment.
Status: The applicant asked that the staff identify the specific items of concern. The staff provided the clarification and the RAI was revised to reflect the staffs concern with bolting.
This RAI was submitted formally.
RAI 3.1.2-1.2 The staff requests additional information on the applicants AMRs for managing loss of material in nickel-based alloy components that are exposed to an internal environment of treated borated water. For the nickel-based alloy components in the reactor coolant system, confirm that loss of material is an applicable aging effect requiring aging management. Specifically, define which aging mechanism or mechanisms are known to induce loss of material in nickel based alloy components that are exposed to an internal environment of treated borated water.
Status: The applicant acknowledged an understanding of the question. Further staff review of the format of the question resulted in the question being revised. The revised form of the question was submitted formally.
RAI 3.1.2-1.3 The staff requests additional information on the applicants AMRs for managing loss of material in stainless steel components that are exposed to an internal environment of treated borated water. For the stainless steel components in the reactor coolant system, confirm that loss of material is an applicable aging effect requiring aging management. Specifically, define which aging mechanism or mechanisms are known to induce loss of material in stainless steel components that are exposed to an internal environment of treated borated water.
Status: The applicant acknowledged an understanding of the question. Further staff review of the format of the question resulted in the question being revised. The revised form of the question was submitted formally.
RAI 3.1.2-1.4 Clarify where the boric acid corrosion aging mechanism is considered in Section 3.1 of the License Renewal Application (LRA), or in Table 3.1.2-1. Specify which component types, materials, environments, aging effects requiring management, and aging management programs are associated with this aging mechanism.
Status: The applicant acknowledged an understanding of the question. This RAI was submitted formally.
Table 3.1.2-2 Reactor Vessel Internals RAI 3.1.2-2.1 The staff requests additional information on the applicants AMRs for managing loss of material and cracking in cast austenitic stainless steel (CASS) components that are exposed to an internal environment of treated borated water. For the CASS components in the reactor coolant system, confirm that loss of material and cracking are applicable aging effects requiring aging management. Specifically, define which aging mechanism or mechanisms are known to induce loss of material and cracking in CASS components that are exposed to an internal environment of treated borated water.
Status: The applicant acknowledged an understanding of the question. Further staff review of the format of the question resulted in the question being revised. The revised form of the question was submitted formally.
Table 3.1.2-3 Class 1 Piping, Valves, and Reactor Coolant Pumps RAI 3.1.2-3.1 In Table 3.1.2-3, on page 3,1-79, the applicant identifies treated water as the external environment for the reactor coolant pump thermal barrier heat exchanger inner coil. In addition, on page 3,1-80, the applicant identifies treated water as the internal environment for the reactor coolant pump thermal barrier heat exchanger outer coil and bored hole heat exchanger. Loss of material, cracking, and fatigue are defined as the aging effects requiring management.
The aging management programs (AMPs) identified to manage these aging effects are Inservice Inspection and Time-Limited Aging Analysis (TLAA)-Metal Fatigue. The applicant's Auxiliary Systems Water Chemistry Control AMP, described in Section B.1.30.1, identifies its purpose as managing loss of material, cracking, and fouling of components exposed to treated water systems. The applicant has identified similar components of the same material which are exposed to the same environment as being managed by a Water Chemistry AMP and referenced concurrence with NUREG-1801, VII.C2.2-a. Provide justification for excluding an AMP to manage the water chemistry of the treated water environment as applicable to these components.
Status: The applicant acknowledged an understanding of the question. This RAI was submitted formally.
RAI 3.1.2-3.2 Clarify where the boric acid corrosion aging mechanism is considered in Section 3.1 of the LRA, or in Table 3.1.2-3. Specify which component types, materials, environments, aging effects requiring management, and aging management programs are associated with this aging mechanism.
Status: This RAI was similar in format and content to RAI 3.1.2-1.4. After discussing the issue, it was mutually agreed that the two questions be combined. The information in this RAI was included in RAI 3.1.2-1.4 and this RAI number was deleted from the formal submission.
Table 3.1.2-4 Pressurizer RAI 3.1.2-4.1 The staff requests additional information on the applicants AMRs for managing cracking in carbon steel components that are exposed to an external air environment. For the carbon steel components in the reactor coolant system, confirm that cracking is an applicable aging effect requiring aging management. Specifically, define which aging mechanism or mechanisms are known to induce cracking in carbon steel components that are exposed to an external air environment.
Status: The applicant understood the question in principle but needed clarification on the specific component type. The staff stated that it was concerned with the support skirt. The RAI was revised to include this information and submitted formally.
RAI 3.1.2-4.2 The staff requests additional information on the applicants AMRs for managing cracking in stainless steel components that are exposed to an external air environment. For the stainless steel components in the reactor coolant system, confirm that cracking is an applicable aging effect requiring aging management. Specifically, define which aging mechanism or mechanisms are known to induce cracking in stainless steel components that are exposed to an external air environment.
Status: The applicant understood the question in principle but needed clarification on the specific component types. The staff stated that it was concerned with the mechanical nozzle seal assembly clamp bolting (studs, nuts, and washers). The RAI was revised to include this information and submitted formally.
RAI 3.1.2-4.3 The staff requests additional information on the applicants AMRs for managing loss of material and cracking in low alloy steel clad with stainless steel and nickel based alloy components that are exposed to an internal environment of treated borated water. For the low alloy steel clad with stainless steel and nickel based alloy components in the reactor coolant system, confirm that loss of material and cracking are applicable aging effects requiring aging management.
Specifically, define which aging mechanism or mechanisms are known to induce loss of material and cracking in low alloy steel clad with stainless steel and nickel based alloy components that are exposed to an internal environment of treated borated water.
Status: The applicant acknowledged an understanding of the question. Further staff review of the format of the question resulted in the question being revised. The revised form of the question was submitted formally.
RAI 3.1.2-4.4 Table 3.1.2-4, Page 3,1-84 identifies the pressurizer lower head, lower shell, upper shell, and upper head as component types. The applicant identified the aging effect of loss of material, and specified that it is applicable to the unclad low alloy steel of the lower head only. Provide justification for limiting the aging effect to only the lower head since many components of the pressurizer are susceptible to boric acid corrosion in a treated borated water environment, and would require that the aging effect of loss of material is managed.
Status: The applicant acknowledged an understanding of the question. This RAI was submitted formally.
RAI 3.1.2-4.5 Recent operational experience at both domestic and foreign facilities (Palo Verde Units 2 and 3, Millstone Unit 2, Waterford Unit 3, and Tsuruga Unit 2 in Japan) has shown that leakage of pressurizer penetrations due to primary water stress corrosion cracking (PWSCC) is an aging effect that requires management. Since AMP B.1.19 Pressurizer Examinations is limited only to managing cracking of the stainless steel and nickel-based alloy cladding and attachment welds by examination of the adjacent base metal, discuss how the aging effect of PWSCC will be managed for the pressurizer penetrations for the period of extended operation at ANO-2.
Include scope, frequency, technique, acceptance criteria, and the technical basis for future examinations.
Status: The applicant acknowledged an understanding of the question. This RAI was submitted formally.
Section 4.2 Reactor Vessel Neutron Embrittlement RAI-4.2-1 The applicant assumes a capacity factor of 80% for the time limiting aging analyses associated with reactor vessel neutron embrittlement that are described in Section 4.2 of the LRA. These evaluations are based on end-of-license (EOL) fluences corresponding to 48 effective power years (EFPY). Staff reviews of current and future trends for plant operations in the nuclear power industry indicate capacity factors of 90% or greater for many plants. Provide justification for the estimated 48 EFPY fluence for ANO-2. If the estimated 48 EFPY fluence cannot be justified, provide results of revised evaluations for reactor vessel neutron embrittlement at higher levels of fluence.
Status: The applicant acknowledged an understanding of the question. This RAI was submitted formally.
RAI 4.2-2 Pursuant to 10 CFR Part 54.21(d), the FSAR Supplement for a facility license renewal application (LRA) must contain a summary description for each aging management program and time-limited aging analysis (TLAA) proposed for management of the effects of aging. The staff has determined that Appendix A of the LRA (FSAR Supplement) did not include a corresponding FSAR Supplement summary description for Table 4.2-2 in TLAA 4.2, Reactor Vessel Neutron Embrittlement of the LRA. Table 4.2-2 contains an evaluation of reactor vessel extended life for pressurized thermal shock. The staff notes that the corresponding table for the upper-shelf energy extended life evaluation (Table 4.2-1) was included in the FSAR Supplement. Pursuant to 10 CFR 54.21(d), the staff requests that a corresponding FSAR Supplement summary description for LRA Table 4.2-2 be included in the FSAR Supplement.
Status: The applicant acknowledged an understanding of the question. This RAI was submitted formally.
Section 4.7.1 Reactor Coolant System Piping Leak-Before-Break RAI-4.7.1-1 A fatigue crack growth analysis supports the TLAA for Reactor Coolant System (RCS) piping leak-before-break (LBB) as described in Section 4.7.1. The fatigue cycles for the time period of extended plant operation are based on thermal transients from a fatigue monitoring program.
What commitments are in place for ANO-2 to continue the fatigue monitoring program through the period of extended plant operation?
Status: The applicant is required to carry its CLB into the period of extended operation. No additional commitment is required. This question was withdrawn.
RAI-4.7.1-2 In Section 4.7.1 of the LRA, the applicant addresses the RCS piping LBB TLAA and concludes that the LBB evaluation for fatigue crack growth remains valid for the period of extended plant operation. How much additional crack growth was predicted by the updated calculations for the end of 60 years compared to that originally predicted for 40 years? What was the criteria or basis for concluding that this amount of additional crack growth was insufficient to exclude fatigue as a damage mechanism that would limit the application of LBB to ANO-2 RCS piping in accordance with the NRC guidance for LBB?
Status: The applicant acknowledged an understanding of the question. This RAI was submitted formally.
Section 4.7.2 Reactor Coolant Pump Code Case N-481 RAI 4.7.2-1 In Section 4.7.2 of the LRA, the applicant addresses the reactor coolant pump (RCP) Code Case N-481 time limiting aging analysis. The applicant used fully aged (saturated) properties in the analysis, and concluded that effects of thermal aging on material properties of cast austenitic stainless steel are addressed for the period of extended operation. Discuss whether these properties are the same bounding properties that were used for embrittled cast stainless materials assumed in the Combustion Engineering report, CEN-367-A which is an analysis for LBB. If other material properties were used, provide justification for the properties that were used for the Code Case N-481 analysis.
Status: The applicant acknowledged an understanding of the question. This RAI was submitted formally.
RAI-4.7.2-2 A fatigue crack growth analysis supports the TLAA for application of RCP Code Case N-481 as described in Section 4.7.2. The fatigue cycles for the period of extended operation are based on thermal transients from a fatigue monitoring program. Describe any commitments that are in place for ANO-2 to continue the fatigue monitoring program through the period of extended operation.
Status: The applicant has committed to maintaining the fatigue monitoring program as stated in LRA Section A.2.1.9. No additional commitment is required. This question was withdrawn.
RAI-4.7.2-3 In Section 4.7.2 of the LRA, the applicant addresses the RCP Code Case N-481 time limiting aging analysis and concludes that the evaluation for fatigue crack growth remains valid for the period of extended plant operation. Discuss the additional crack growth that was predicted by the updated calculations at the end of 60 years, and compare the crack growth to that originally predicted for 40 years. Provide the criteria or basis for concluding that this amount of additional crack growth is sufficiently small to justify continued application of Code Case N-481.
Status: The applicant acknowledged an understanding of the question. This RAI was submitted formally.
Section 4.7.3 Reactor Coolant Pump Flywheel RAI 4.7.3-1 In 4.7.3 (Page 4.7-2) of the LRA, the applicant concluded that the RCP flywheel is not a TLAA.
The basis for this conclusion is a 1997 safety evaluation of a fatigue crack growth analysis that was presented in a Combustion Engineering Owners Group topical report. This safety evaluation allowed the licensee to lengthen the RCP flywheel inspection period for ANO Units 1 and 2 and five other units. The fatigue crack growth analysis for ANO Units 1 and 2 is based on 4,000 RCP startup and shutdown cycles. The RCP flywheel was identified as a TLAA in the LRA for ANO Unit 1, and two other units that are identified in the topical report and that have been granted renewed licenses.
Justify why the RCP flywheel is not a TLAA or provide the TLAA for the RCP flywheel for ANO Unit 2, and include the justification for why 4,000 RCP startup and shutdown cycles remain bounding through the end of the extended period of operation for ANO-2. In addition, the applicant must include an FSAR Supplement summary description, in Appendix A, of the LRA for the TLAA on fatigue-induced crack growth of the ANO-2 RCP flywheel. The summary description should include a discussion on the safety margin for the acceptable flaw size, and the justification for why 4,000 RCP startup and shutdown cycles remain bounding through the end of the extended period of operation for ANO.
Status: The applicant acknowledged an understanding of the question but stated that the form of the question required them to answer the portion of the question concerning the 4,000 cycles even if the RCP flywheel was not a TLAA. The staff agreed to revise the question. The revised RAI was submitted formally.
Section 4.7.5 Alloy 600 Nozzle Repairs RAI-4.7.5-1 A fatigue crack growth analysis supports the TLAA for the evaluation of Alloy 600 Nozzle Repairs as described in Section 4.7.5. The fatigue cycles for the time period of extended plant operation are based on thermal transients from a fatigue monitoring program.
What are the commitments for ANO-2 to continue the fatigue monitoring program through the period of extended plant operation?
Status: The applicant has committed to maintaining the fatigue monitoring program as stated in LRA Section A.2.1.9. No additional commitment is required. This question was withdrawn.
RAI 4.7.5-2 Demonstrate that the designs of repaired nozzles will have sufficient structural integrity against loss of material by corrosion and will meet their minimum wall thickness requirements through the expiration of the extended period of operation.
Status: The applicant acknowledged an understanding of the question. This RAI was submitted formally.
RAI 4.7.5-3 Justify and validate the CEOGs conclusion that growth of the existing flaw in the original Alloy 600 J-groove weld material by stress corrosion cracking is not a plausible effect during the period of extended operation.
Status: The applicant acknowledged an understanding of the question. This RAI was submitted formally.
B.1.1 Alloy 600 Aging Management RAI B.1.1-1 Confirm that all of the components listed in the Alloy 600 Aging Management Program are covered under the inservice inspection requirements of Section XI of the ASME Code, and for any components not covered by Section XI inservice inspection requirements, provide a complete description of the proposed inspections including a technical justification for the inspection method and frequency.
Status: The applicant acknowledged an understanding of the question. This RAI was submitted formally.
RAI B.1.1-2 The applicant states that components listed in the Alloy 600 Aging Management Program will be inspected in accordance with the requirements of Section XI of the ASME code. Confirm that currently docketed requests for relief for ANO-2 do not affect compliance with the full scope of the Section XI inservice inspection requirements proposed for components listed in the scope of the Alloy 600 Aging Management Program. If any requests for relief from the ASME Code requirements do affect the components listed in the scope of the Alloy 600 Aging Management Program, provide a complete description of the proposed inspections including technical justifications for any proposed alternative examinations.
Status: Since current request for relief will expire before entering the period of extended operation, this information is not relevant. This RAI was withdrawn.
RAI B.1.1-3 The applicant stated that no preventative actions will be taken as part of the Alloy 600 Aging Management Program to prevent aging effects or mitigate aging degradation. The NRC staff notes that several preventive actions and common industry practices have been used to mitigate PWSCC associated with Alloy 600. Examples of these include: nickel plating of the surfaces of Alloy 600 components that are exposed to treated water, replacement of leaking Alloy 600 instrument nozzles with Alloy 690 material, preventive replacement of selected pressurizer and RCS piping instrument nozzles with Alloy 690 material, monitoring the electrochemical potential, and water chemistry control. Provide a description of any preventive actions and/or water chemistry monitoring programs ANO-2 is currently implementing that may be used to address the Alloy 600 cracking issue.
Status: The applicant acknowledged an understanding of the question. This RAI was submitted formally as RAI B.1.1-2.
RAI B.1.1-4 In the Alloy 600 Aging Management Program under the program attribute, Detection of Aging Effects, the applicant states that the measurement, vent, upper level, and temperature nozzles, and heater sheath, heater sleeve, and end plug received visual examination (VT-2) from the exterior of the vessel in accordance with ASME Section XI, Examination Category B-P. For many of these components, the Alloy 600 pressure boundary welds are covered by insulation.
Service experience has shown that, early indications of through-wall leakage (e.g., boric acid on the component surface) are very difficult to detect when VT-2 examinations are performed with the insulation in place. Provide justification for not removing insulation when performing VT-2 examinations on the components mentioned above. In addition, provide the frequency of inspection, and the results of any volumetric non-destructive examination that has been performed.
Status: The applicant acknowledged an understanding of the question. This RAI was submitted formally as RAI B.1.1-3 RAI B.1.1-5
- a. In the Alloy 600 Aging Management Program under the program attribute, Detection of Aging Effects, the applicant states that guidance from the Materials Reliability Project (MRP) in conjunction with the PWR owners groups will be used to identify critical locations for inspection and augmentation of existing ISI inspections at ANO-2 where appropriate. The staff believes that the strategic plan developed by the industry will be comprehensive and recommendations may be applicable to all 10 elements of the Alloy 600 Aging Management Program. Identify the date that ANO-2 commits to submit, for review and approval, an augmented aging management program that includes all recommendations from the industrys strategic plan, and meets the 10 elements in accordance with the guidance in NUREG-1800, Appendix A.1, Aging Management Review - Generic, Table A.1-1, Elements of an Aging Management Program for License Renewal. The date must be prior to the period of extended operation.
- b. The Updated Final Safety Analysis Report (UFSAR) for ANO-2 does not contain a commitment to use guidance developed by the Electric Power Research Institute (EPRI)
MRP program, and to submit the inspection plan for review and approval. Confirm that the UFSAR will be revised to reflect the above mentioned commitment for management of Alloy 600 components.
Status: The applicant acknowledged an understanding of the question. This RAI was submitted formally as RAI B.1.1-4 RAI B.1.1-6 In the Alloy 600 Aging Management Program under the program attribute, Operating Experience, the applicant states that the Alloy 600 aging management program is a new program for which there is no specific operating experience for ANO-2. The staff is aware of several reported failures related to Alloy 600 welded components in other PWRs including several failures in other Combustion Engineering (CE) NSSS design. Specifically, PWSCC has been reported in Alloy 82/182 J-groove welds that are used to join Alloy 600 small bore nozzles to CE-designed pressurizers, steam generators, and/or hot legs. The staff believes it important for the applicant to review relevant industry service experience and incorporate lessons learned into the Alloy 600 program. Therefore, the applicant should discuss what industry initiatives it plans to follow in order to incorporate experience related to Alloy 600 into the ANO-2 Alloy 600 AMP.
Status: The applicant acknowledged an understanding of the question. This RAI was submitted formally as RAI B.1.1-5 B.1.3 Boric Acid Corrosion Prevention RAI B.1.3-1 Provide the basis for the proposed acceptance criteria that will be developed as part of the following enhancement to the Boric Acid Corrosion Prevention AMP:
The program acceptance criteria will be revised to address electrical components in addition to ferritic steel.
The applicant retains the program description of the Boric Acid Corrosion Prevention Program, as well as the descriptions of the programs 10 elements, on record at the Arkansas Nuclear One Unit 2 facility.
Status: The applicant acknowledged an understanding of the question. The staff revised the question to remove unnecessary background information. The revised RAI was submitted formally.
RAI B.1.3-2 In the Operating Experience Section of B.1.3, Boric Acid Corrosion Prevention, the applicant states that recent industry events regarding reactor vessel head degradation required assessments at each site to ensure that boric acid corrosion prevention programs are adequate and functioning effectively. The applicant also states that a self assessment was performed in February 2003, and no significant findings were identified during this assessment. Discuss how program revisions have incorporated lessons learned from the Davis-Besse vessel head degradation and the control rod drive mechanism penetration cracking discussed in NRC Bulletins 2001-01, 2002-01, 2002-02, and NRC Order EA-03-009. Also, provide a discussion on implementation of corrective actions in the program to prevent reoccurrence of degradation caused by boric acid leakage, as required by Generic Letter 88-05.
Status: The applicant acknowledged an understanding of the question. This RAI was submitted formally.
B.1.20 Reactor Vessel Head Penetrations RAI B.1.20-1
- a. The applicant states that the Corrective Action Program was used to incorporate industry operating experience into the Reactor Vessel Head Penetration program, and to develop inspection requirements that are specific to ANO-2. The applicant also states that recent reactor vessel head penetration nozzle inspections were performed in accordance with the commitments in the ANO-2 response to NRC Bulletin 2001-01.
The NRC staff notes that in February 2003, NRC Order EA-03-009 was issued. This order supercedes NRC Bulletins 2001-01 and 2002-01, and requires that licensees assess the susceptibility of the reactor vessel head to PWSCC-related degradation.
The Order also requires the licensee to commit to an augmented inspection program for the reactor pressure vessel head based upon the susceptibility to PWSCC. Provide a commitment for implementation of the ANO-2 augmented inspection plan for the period of extended operation.
- b. The Updated Final Safety Analysis Report (UFSAR) for ANO-2 does not contain a commitment to incorporate the requirements of NRC Order EA-03-009, and to submit the inspection plan for review and approval. Confirm that the UFSAR will be revised to reflect the above mentioned commitment for inspection of the reactor pressure vessel head.
Status: The applicant acknowledged an understanding of the question. This RAI was revised by the staff to be consistent with previous commitments. The request for the applicant to re-submit the inspection for review and approval was eliminated from the RAI. The RAI was revised and submitted formally.
B.1.21 Reactor Vessel Integrity RAI B.1.21-1 The description of this AMP includes the following enhancement, and also includes a statement that the enhancement will be initiated prior to the period of extended operation:
The ANO-2 specimen capsule withdrawal schedule will be revised to withdraw and test a standby capsule to cover the peak fluence expected through the end of the period of extended operation.
The Updated Final Safety Analysis Report Supplement (FSAR) Table 5.2-12 on page A-8 of the application proposes a change to the withdrawal schedule where the three standby capsules are designated with note (a) which states:
(a) If required, Capsules 4, 5, or 6 will be repositioned to ensure that peak fluence is obtained prior to 60 years.
The proposed FSAR change does not accurately reflect the enhancement described in AMP B.1.21. Please provide a commitment to revise the surveillance capsule withdrawal schedule and the FSAR Supplement to ensure that a standby capsule will be withdrawn and tested in order to cover the peak RPV fluence expected through the end of the period of extended operation. As stated in the FSAR Supplement, the revised withdrawal schedule must be submitted to the NRC for review and approval, prior to implementation.
In addition, the following license condition will be placed on the ANO-2 renewed license:
All capsules in the reactor vessel that are removed and tested must meet the test procedures and reporting requirements of the American Society for Testing and Materials (ASTM) E 185-82 to the extent practicable for the configuration of the specimens in the capsule. Any changes to the capsule withdrawal schedule, including spare capsules, must be approved by the NRC prior to implementation. All capsules placed in storage must be maintained for future insertion. Any changes to storage requirements must be approved by the NRC.
Status: The applicant acknowledged an understanding of the question. However, the staff revised the question to request that the applicant submit the specific reactor vessel specimen capsule withdrawal schedule through the end of the period of extended operation. The RAI was revised and submitted formally.
B.1.22 Reactor Vessel Internals Cast Austenitic Stainless Steel Components RAI B.1.22-1 The Reactor Vessel Internals Cast Austenitic Stainless Steel Components (CASS) AMP is currently not in place and the applicant states in LRA Section B.1.22, that it will initiate the program before the period of extended operation. The staff requests that: (1) the applicant formally make a commitment to submit the program for NRC review and approval no later than three years before the period of extended operation, and (2) the applicant include this commitment in the ANO-2 LRA commitment tracking system.
Status: The applicant acknowledged an understanding of the question. This RAI was submitted formally.
RAI B.1.22-2 The staff found some differences between Table 3.1-2 in NUREG-1800 (Standard Review Plan for License Renewal) and the ANO-2 LRA UFSAR Section A.2.1.23, Reactor Vessel Internals CASS Program. Table 3.1-2 in the Standard Review Plan provides a description of what should be included in the UFSAR Supplement for aging management of RV internals, and the reactor coolant system for license renewal reviews. The proposed UFSAR change does not state that the inservice inspection program will be augmented to include enhanced examinations of non-bolted components, and other demonstrated acceptable methods for bolted components for certain susceptible or limiting components or locations. Also, please clarify why the enhanced examination and/or component-specific flaw evaluation for the CASS component, which are specified in NUREG-1800, are not included in ANO-2 LRA UFSAR Section A.2.1.23.
Status: The applicant acknowledged an understanding of the question. This RAI was submitted formally.
RAI B.1.22-3 The staff requests that the applicant include in the ANO-2 LRA UFSAR Section A.2.1.23 the commitment to submit the Reactor Vessel Internals Cast Austenitic Stainless Steel Components (CASS) AMP for NRC review and approval no later than three years before the period of extended operation.
Status: The applicant stated that the Reactor Vessel Internals CASS AMP would be consistent with GALL. The staff determined that there was no need to have the program submitted for review and approval. This RAI was withdrawn.
B.1.23 Reactor Vessel Internals Stainless Steel Plates, Forgings, Welds, and Bolting RAI B.1.23-1 This Reactor Vessel Internals Stainless Steel Plates, Forgings, Welds, and Bolting AMP is currently not in place and the applicant states in LRA Section B.1.23, that it will initiate the program before the period of extended operation. The staff requests that: (1) the applicant formally make a commitment to submit the program for NRC review and approval no later than three years before the period of extended operation, and (2) the applicant include this commitment in the ANO-2 LRA commitment tracking system.
Status: The applicant acknowledged an understanding of the question. The staff revised the question to better address the issue of void swelling, which was the staffs primary concern.
The staff requested that the applicant submit the inspection program to manage the aging effects associated with void swelling to the NRC for review and approval. The RAI was revised and submitted formally.
RAI B.1.23-2 The staff noted that the water chemistry system and the enhanced examination of non-bolted components are not discussed in LRA UFSAR Supplement, Section A.2.1.24 as is stated in NUREG-1800, Table 3.1-2, Page 3,1-27. The applicant should revise LRA Section A.2.1.24 to be consistent with Table 3.1-2 of NUREG-1800 (Page 3,1-27).
Status: The applicant acknowledged an understanding of the question. This RAI was submitted formally.
RAI B.1.23-3 The staff requests that the applicant include in the ANO-2 LRA UFSAR Section A.2.1.24 the commitment to submit the Reactor Vessel Internals Stainless Steel Plates, Forgings, Welds, and Bolting AMP for NRC review and approval no later than three years before the period of extended operation.
Status: The staff determine that there was no need to have the program submitted for review and approval. This RAI was withdrawn.
DISTRIBUTION: Dated: June 21, 2004 Accession No.: ML041730571 HARD COPY RLEP RF G. Suber (PM)
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Arkansas Nuclear One, Unit 2 cc:
Executive Vice President
& Chief Operating Officer Entergy Operations, Inc.
P. O. Box 31995 Jackson, MS 39286-1995 Director, Division of Radiation Control and Emergency Management Arkansas Department of Health 4815 West Markham Street, Slot 30 Little Rock, AR 72205-3867 Winston & Strawn 1400 L Street, N.W.
Washington, DC 20005-3502 Mr. Mike Schoppman Framatome ANP, Richland, Inc.
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P. O. Box 31995 Jackson, MS 39286-1995 Wise, Carter, Child & Caraway P. O. Box 651 Jackson, MS 39205 Garry Young 1448 SR 333 Russellville, AR 72802 Mr. Fred Emerson Nuclear Energy Institute 1776 I St., N.W., Suite 400 Washington, DC 20006-3708