ML11167A171
| ML11167A171 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 06/20/2011 |
| From: | Cuadradodejesus S License Renewal Projects Branch 1 |
| To: | Allen B FirstEnergy Nuclear Operating Co |
| CuadradoDeJesus S | |
| References | |
| TAC ME4640 | |
| Download: ML11167A171 (14) | |
Text
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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001
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"'I} ~o Barry S. Allen Vice President, Davis-Besse Nuclear Power Station FirstEnergy Nuclear Operating Company 5501 North State Route 2 Oak Harbor, OH 43449
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE DAVIS-BESSE NUCLEAR POWER STATION (TAC NO.: ME4640)
Dear Mr. Allen:
By letter dated August 27,2010, FirstEnergy Nuclear Operating Company, submitted an application pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Part 54 for renewal of Operating License NPF-3 for the Davis-Besse Nuclear Power Station. The staff of the U.S.
Nuclear Regulatory Commission (NRC or the staff) is reviewing this application in accordance with the guidance in NUREG-1800, "Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants." During its review, the staff has identified areas where additional information is needed to complete the review. The staff's requests for additional information are included in the enclosure. Further requests for additional information may be issued in the future.
Items in the enclosure were discussed with Cliff Custer, of your staff, and a mutually agreeable date for the response is within 30 days from the date of this letter. If you have any questions, please contact me by telephone at 301-415-2946 or bye-mail at Samuel. CuadradoDeJesus@nrc.gov.
Sincerely,
.mu~iD~s,iler
~ts Branch 1 Ivision of License Renewal Office of Nuclear Reactor Regulation Docket No. 50-346
Enclosure:
As stated cc w/encl: Listserv
DAVIS-BESSE NUCLEAR POWER STATION LICENSE RENEWAL APPLICATION REQUEST FOR ADDITIONAL INFORMATION RAI 8.2.34-1
Background:
The preventive actions program element of Generic Aging Lessons Learned (GALL), Rev. 2, aging management program (AMP) XI.M3, "Reactor Head Closure Stud Bolting," references the guidance outlined in Regulatory Guide (RG) 1.65, Materials and Inspections for Reactor Vessel Closure Studs," and NUREG-1339, "Resolution of Generic Safety Issue 29: Bolting Degradation or Failure in Nuclear Power Plants." AMP XI.M3 states that one of the preventive measures that can reduce the potential for stress-corrosion cracking includes using bolting material for closure studs that has an actual measured yield strength less than 150 ksi. During its audit, the U.S. Nuclear Regulatory Commission (NRC or the staff) noted that the FirstEnergy Nuclear Operating Company's (FENOC or the applicant) program basis document for its Reactor Head Closure Studs Program states that the reactor head closure studs and nuts are manufactured from SA-540, Grade 23 material.
Issue:
License renewal application (LRA) Section 8.2.34 and the applicant's program basis document do not include the preventive action of using stud materials with an actual measured yield strength level less than 150 ksi. The staff needs to confirm the actual measured yield strength of the applicant's reactor head closure stud material to determine whether the applicant's program is adequate to manage stress-corrosion cracking.
Reguest:
The staff requests the following information:
- 1) Clarify whether the actual measured yield strength of the reactor head closure stud material is less than 150 ksi. If the reactor head closure stud material has a measured yield strength level greater than or equal to 150 ksi, justify the adequacy of the AMP to manage stress-corrosion cracking in the high-strength material.
- 2) Clarify if preventive actions will be added to the Reactor Head Closure Studs Program that would preclude the future use of replacement closure stud bolting fabricated from material with actual measured yield strength greater than or equal to 150 ksi. If not, and in view of the greater susceptibility of the studs for stress-corrosion cracking, describe any preventative actions to avoid exposure of the studs to environments conducive to stress-corrosion cracking. Otherwise, justify why preventative measures to mitigate stress-corrosion cracking of high strength studs wi" not be required.
ENCLOSURE
-2 RAI B.2.9-3
Background:
"Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants" (SRP-LR), Section A 1.2.3.4, "detection of aging effects," states that the parameters to be monitored include aspects such as frequency and sample size. This program element states that the basis for the inspection population and sample size should consider the environment, locations most susceptible to the aging effect, and include provisions for expanding the sample size when degradation is detected in the initial sample. The SRP-LR also states that the applicant should provide a justification, including codes and standards referenced, that the technique and frequency are adequate to detect the aging effects before a loss of component's intended function.
The GALL Report recommends the use of AMP XI,M20 "Open-Cycle Cooling Water System" for materials included within the scope of the plant-specific Collection, Drainage and Treatment Components Inspection Program that are exposed to raw water. In LRA Section B.2.9, Collection, Drainage and Treatment Components Inspection Program, under the "detection of aging effects" program element, the applicant stated that if opportunistic inspections have not occurred prior to the period of extended operation then a focused inspection, inclusive of each material in the scope of the program will be performed. The application further states that any evidence of degradation that could lead to loss of a component intended function will be evaluated through the corrective action program to determine the need for subsequent inspections, expansion, and for monitoring and trending the results.
Issue:
Although the systems referenced in the Collection, Drainage and Treatment Components Inspection Program are in some cases different than those listed in the GALL Report for which AMP XI,M20 "Open-Cycle Cooling Water System" is recommended, it is clear that for the material and environment combination stated in the LRA, a periodic inspection program is recommended.
The staff lacks sufficient information to determine if the inspections conducted for the Collection, Drainage and Treatment Components Inspection Program will be periodic or one-time, or the basis for the inspection frequency, if in the absence of evidence of degradation during a planned inspection, no further inspections are conducted. The staff also lacks sufficient information to find the basis for the inspection size and locations acceptable.
Request:
State the basis for why a one-time inspection would be sufficient for managing the effects of aging for collection, drainage, and treatment components or revise the Collection, Drainage and Treatment Components Inspection Program to ensure that periodic inspections are performed.
State the basis for the sample size, the selection factors for the "most susceptible" materials and locations, the frequency to be used during the period of extended operation, and the percentage increase in sample size should degradation be detected.
- 3 RAI8.2.9-4
Background:
SRP-LR Section A.1.2.3.10, "operating experience," states that "Additionally, an applicant should commit to a review of future plant-specific and industry operating experience for new programs to confirm their effectiveness." In LRA Appendix A.3, Table A-1, License Renewal Commitment List the new Collection, Drainage and Treatment Components Inspection Program does not include a commitment to perform a review of future operating experience to confirm the effectiveness of this program.
Issue:
This program's LRA commitment list is not consistent with the current staff position as stated within the SRP-LR, Revision 2 concerning reviews of future operating experience for new AMPs.
Request:
Revise LRA Appendix A.3, Table A-1, License Renewal Commitment List, item no. 4 for the Collection, Drainage and Treatment Components Inspection Program to include a commitment to perform a future review of operation experience to confirm the effectiveness of this program or justify why such a review is not necessary.
RAI8.2.9-5
Background:
SRP-LR Section A.1.2.3.3, "parameters monitored or inspected," states that an applicant should provide a link between the parameter(s) that will be monitored and how the monitoring of these parameters will ensure adequate aging management. In LRA Section B.2.9, "Collection, Drainage and Treatment Components Inspection Program," under the "parameters monitored or inspected" program element, the applicant stated that parameters monitored or inspected are directly related to degradation of the components under review.
Issue:
This program does not provide the details for what parameters, such as wall thickness and surface degradation will be monitored and used to ensure adequate aging management will be completed.
Request:
State what parameters will be linked to detecting the following: (a) loss of material; (b) cracking; and (c) a reduction in heat transfer during the visual inspection. State the basis for detecting loss of material on inaccessible surfaces (e.g., tank bottoms sitting on concrete) using a visual
-4 inspection, or revise the program to include volumetric inspections that are capable of adequately managing this aging effect.
RAI 8.2.30-1
Background:
The GALL Report Rev. 2, AMP XI.M32, "One-Time Inspection," states in the "detection of aging effects program element" that for components managed by the AMP XI.M2, "Water Chemistry;"
AMP XI.M30, "Fuel Oil Chemistry;" and AMP XI.M39, "Lubricating Oil Analysis;" a representative sample size is 20 percent of the population (defined as components having the same material, environment. and aging effect combination) or a maximum of 25 components. LRA AMP B.2.30 states that the sample population will be determined by engineering evaluation, and where practical, will be focused on the (bounding or lead) components considered most susceptible to aging degradation due to time in service, the severity of the operating cOr:lditions, and the lowest design margin.
Issue:
Given that the GALL Report. Rev. 2, represents the current staff position on the sample size for the "One-Time Inspection" Program, LRA Section B.2.30 does not provide enough information for the staff to determine if the sample size for this program is consistent with the GALL Report AMP XI.M32.
Request:
State the planned sample size for the One-Time Inspections of the Pressurized-Water Reactor (PWR) Water Chemistry, Fuel Oil Chemistry, and Lubricating Oil Analysis Programs. If the sample size is less than 20 percent of the population (defined as components having the same material, environment, and aging effect combination) or a maximum of 25 components, then state the basis for why the sample size will be representative of aging effects in the systems, and will be sufficient to verify the system-wide effectiveness of the chemistry programs.
RAI 8.2.30-2
Background:
SRP-LR Rev. 2, Table 3.0-1, "FSAR [Final Safety Analysis Report] Supplement for Aging Management of Applicable Systems," states that GALL Report AMP XI.M32. "One-Time Inspection" Program, cannot be used for structures or components with known age-related degradation mechanisms or when the environment in the period of extended operation is not expected to be equivalent to that in the prior 40 years, and that periodic inspections should be proposed in these cases. SRP-LR Section 3.0.1 states that the FSAR Supplement should also contain a commitment to implement the LRA AMP enhancement prior to the period of extended operation. Title 10 of the Code of Federal Regulations (10 CFR) 54.21(d) states that the FSAR supplement must contain a summary description of the program and the activities for managing the effects of aging. In addition, SRP-LR 3.3.2.4 states that the summary description of the
- 5 programs and activities for managing the effects of aging for the period of extended operation in the FSAR Supplement should be sufficiently comprehensive such that later changes can be controlled by 10 CFR 50.59, and the description should contain information associated with the bases for determining that aging effects will be managed during the period of extended operation.
In its response to RAls 3.3.2.2.5-1, 3.3.2.71-2, B.2.8-1, and B.2.18-1, dated May 24,2011, LRA Section A.1.30, "One-Time Inspection," was revised; however, the change did not include the above wording from SRP-LR Table 3.0-1.
Issue:
The updated (UFSAR) supplement does not reflect change that occurred in Revision 2 to the SRP-LR Table 3.0-1, as stated above. The staff believes that this information is associated with the bases for determining that the aging effects for buried in-scope components will be effectively managed during the period of extended operation. The staff also believes that this information should be explicitly stated in the FSAR supplement to ensure that the licensing basis for the period of extended operation is clear.
Request:
Revise LRA Section A.1.30 to be consistent with and provide the equivalent information as stated within SRP-LR, Rev. 2, Table 3.0-1 GALL Report AMP XI.M32, "One-Time Inspection" Program.
RAI B.2.30-3
Background:
SRP-LR Revision 2, A.1.2.3.10.3 states that, "Additionally, an applicant should commit to a review of future plant-specific and industry operating experience for new programs to confirm their effectiveness."
In LRA Appendix A.3, Table A-1, "License Renewal Commitment List," the new One-Time Inspection Program does not include a commitment to perform a future review of operating experience to confirm the effectiveness of this program.
Issue:
The new One-Time Inspection Program's LRA commitments are not consistent with the current staff position as stated within the SRP-LR, Rev. 2, concerning reviews of future operating experience for new programs.
- 6 Request:
Revise LRA Appendix A3, Table A-1, "License Renewal Commitment List," for the One-Time Inspection Program to include a commitment to perform a future review of operating experience to confirm the effectiveness of this program or state why such a review is not necessary.
Follow-up RAI B.2.1-2
Background:
The applicant responded to the RAI B.2.1-1 by proposing to revise Subsection 2.1.2 of the Davis Besse Nuclear Power Station (DBNPS) Surveillance Test Procedure DB-PF-03009, Revision 06, "Containment Vessel and Shielding Building Visual Inspection." Revised Subsection 2.1.2 shall state "Personnel who performed general visual examinations of the exterior surface of the Containment vessel and the interior and exterior surfaces of the Shielding Building shall meet the requirements for a general visual examiner in accordance with Nuclear Operating Procedure NOP-CC-5708, Written Practice for the Qualification and Certification of Nondestructive Examination Personnel."
Issue:
Element 5 "Detection of Aging Effects" in GALL AMP XI.S4 recommends the implementation of periodic in-service examinations for the containment structures by applying the requirements of subsections in ASME Section XI. The associated Subsection IWE-3510.1 of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI (1995), requires that "The general Visual Examination shall be performed by, or under the direction of, a Registered Professional Engineer or other individual, knowledgeable in the requirements for design, in-service inspections, and testing of Class MC and metallic liners of Class CC components."
Request:
To comply with the ASME Code,Section XI requirement, the associated Subsection IWE-3510.1 of ASME Code,Section XI (1995) code requirement must be referenced in the new revision of the DBNPS's Nuclear Operating Procedure and/or Surveillance Test Procedure.
RAI B.2.21-6
Background:
SRP-LR Section A 1.2.3.1 0 states, in part, that for new AMP that have yet to be implemented at an applicant's facility, the programs have not yet generated any operating experience (OE).
However, there may be other relevant plant-specific OE at the plant or generic OE in the industry that is relevant to the AMP's program elements even though the OE was not identified as a result of the implementation of the new program. Thus, for new programs, the applicant may need to consider the impact of relevant OE that results from past implementation of its
-7 existing AMPs that are existing programs and the impact of relevant generic OE on developing program elements.
As part of RAI B.2.21-1, the staff requested the applicant provide a summary of their evaluation of recently identified industry operating experience and any plant-specific operating experience concerning inaccessible low voltage power cable failures within the scope of license renewal.
The staff also requested the applicant provide an evaluation showing how the Non-EO Inaccessible Medium-Voltage Program test and inspection frequencies, including event driven inspections, incorporate recent industry and plant specific operating experience for both inaccessible low and medium voltage power cable.
Issue:
In its RAI response dated May 5, 2011, the applicant referenced their response to Generic Letter (GL) 2007-01, "Inaccessible or Underground Power Cable Failures that Disable Accident Mitigation Systems or Cause Plant TranSients," dated May 8,2007. The applicant did not provide additional operating experience for inaccessible low and medium voltage power cable subsequent to the applicant's GL response.
Request:
Provide a summary of inaccessible low and medium voltage cable operating experience (both testing and operating) subsequent to your May 8, 2007 response to GL 2007-01.
RAI 8.2.21-7
Background:
SRP-LR Section 3.0.1 states, in part, that each LRA will provide an FSAR Supplement which defines the changes to the FSAR that will be made as a condition of a renewed license. The FSAR Supplement defines the AMPs the applicant is crediting to satisfy 10 CFR 54.21(a)(3).
SRP-LR Table 3.0.1 states (along with an inspection performed at least annually and event driven inspections) that the inspection frequency for water collection is established and performed based on plant-specific operating experience with cable wetting or submergence.
As part of RAI 8.2.21-1, the staff requested the applicant to explain how DBNPS will manage the effects of aging on inaccessible low voltage power cables within the scope of license renewal with consideration of recent industry operating experience and applicable plant-specific operating experience including an assessment of the program elements and the USAR summary description for the Inaccessible Power Cables Not Subject to 10 CFR 50.49 EO Requirements Program. The applicant's RAI response indicates that the USAR will be revised to include the change. The applicant's RAI response did revise Commitment No. 11 to include the above change.
-8 Issue As part of the applicant's response to RAI B.2.21-1 the applicant revised the LRA USAR summary description for the Inaccessible Power Cables Not Subject to 10 CFR 50.49 EQ Requirements Program but did not state that the inspection frequency for water collection is established and performed based on plant-specific operating experience with cable wetting or submergence consistent with SRP-LR Table 3.0.1 and GALL AMP XI.E3.
Request:
Explain why the USAR summary description provided in the response to RAI B.2.21-1 does not include the provision that manhole inspection frequencies will be based on plant-specific operating experience consistent with SRP-LR Table 3.0.1 and GALL AMP XI.E3.
RAI 3.1.2.2-2
Background:
LRA Table 3.1.2-2 indicates the reactor vessel internals components made of cast austenitic stainless steel (CASS) subject to reduction in fracture toughness and managed by the PWR Vessel Internals Program. These CASS components are the following: (1) Incore guide tube assembly spider in the core support assembly (CSA); (2) Plenum control rod guide tube (CRGT). spacer casting; (3) CSA vent valve assembly valve body; and (4) Plenum cylinder reinforcing plate.
LRA Section 8.2.32 states that the PWR Vessel Internals Program is based on the examination requirements provided in Electric Power Research Institute (EPRI) Topical Report 1016596.
"Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227, Rev. 0)," along with the implementation guidance described in NEI 03-08.
The staff noted that MRP-227, Rev. 0, which is referenced in the GALL Report Rev. 2, categorizes the reactor vessel internals components based on the following functional groups:
primary. expansion, existing programs, and no additional measures. MRP-227 also specifies relevant examination methods and coverage for the expansion group components based on the examination findings of the primary group components.
In addition, GALL Report, Rev. 2, AMP XI.M16A and MRP-227, Rev. 0, Tables 3-1,4-1 and 4-4 indicate that, in B&W plants. the following CASS vessel internals are the primary group components to be managed for loss of fracture toughness: (1) core support shield (CSS) cast outlet nozzles, (2) CSS vent valve discs, and (3) incore monitoring instrumentation (lMI) guide tube assembly spiders (accessible top surfaces). MRP-227, Rev. 0, also indicates that these primary group components have link relationships with CRGT spacer castings (accessible surfaces at four screw locations), which are the associated expansion group components.
- 9 Issue The staff noted that, in contrast with MRP-227, Rev. 0, LRA Table 3.1.2-2 does not clearly identify the functional groups and link relationships for the following components: (1) CSS outlet nozzles, (2) CSS vent valve discs, (3) Incore guide tube assembly spiders, and (4) CRGT spacer castings. In addition, LRA Table 3.1.2-2 does not clearly indicate the functional groups and link relationships for the following two components: (1) CSA vent valve body and (2) plenum cylinder reinforcing plate.
Request:
- 2. Describe the functional groups for the following components: (1) CSS outlet nozzles, (2) CSS vent valve discs and (3) IMI guide tube assembly spiders (accessible top surfaces),
and (4) CRGT spacer castings (accessible surfaces at four screw locations). In addition, describe the link relationships for these components (such as primary/expansion link). If the aSSigned functional groups or links are not consistent with MRP-227, Rev. 0, justify why the inconsistency is acceptable to manage the reduction in fracture toughness of these components.
- 3. Describe the functional groups for the following two components addressed in LRA Table 3.1.2-2: (1) CSA vent valve body, and (2) plenum cylinder reinforcing plate. If existent, describe their link relationships (such as primary/expansion link) with other components. In addition, describe the assigned inspection method including frequency of the components.
Also, provide the technical basis for the assigned component groups, link relationships and inspection method/freq uency.
- 4. Revise LRA Table 3.1.2-2 and other related information in the LRA consistent with the response to this RAI.
RAI 8.2.36-4
Background:
The "acceptance criteria" program element of GALL Report (Revision 2) AMP XI.M33, "Selective Leaching," recommends that the acceptance criteria include no visible evidence of selective leaching or no more than a 20 percent decrease in hardness. GALL Report AMP XI.M33 also recommends that for copper alloys with greater than 15 percent zinc, the acceptance criteria is no noticeable change in color from the normal yellow color to the reddish copper color. LRA Section B.2.36 states that the selective leaching inspection will utilize approved inspection techniques to identify selective leaching, and inspection results that identify selective leaching will be entered into the Corrective Action Program.
- 10 Issue:
It is not clear to the staff how the GALL Report, Rev. 2, AMP XI.M33 recommendations in the "acceptance criteria" program element are addressed in the applicant's Selective Leaching Inspection Program.
Request:
Describe how the GALL Report, Rev. 2, AMP XI.M33 recommendations in the "acceptance criteria" program element are addressed in the Selective Leaching Inspection Program. If the recommended acceptance criteria are not included, state the basis for not including these acceptance criteria in the Selective Leaching Inspection Program and propose an alternate acceptance criteria that is capable of identifying the aging effects before a loss of intended function.
RAI A.1.2-1
Background:
SRP-LR 3.3.2.4 states that the summary description of the programs and activities for managing the effects of aging for the period of extended operation in the FSAR Supplement should be sufficiently comprehensive such that later changes can be controlled by 10 CFR 50.59, and the description should contain information associated with the bases for determining that aging effects will be managed during the period of extended operation. In addition, 10 CFR 54.21 (d) states that the FSAR supplement must contain a summary description of the program and the activities for managing the effects of aging.
SRP-LR Tables 3.3-2 and 3.4-2 recommend that the FSAR Supplement for the Aboveground Steel Tanks Program should state that the program includes preventive measures to mitigate corrosion by protecting the external surface of steel components per standard industry practice and with sealant or caulking at the interface of concrete and component, and verification of the effectiveness of the program by measuring the thickness of the tank bottoms to ensure that Significant degradation is not occurring.
Issue:
The USAR supplement, LRA Section A.1.2, does not reflect that the Aboveground Steel Tanks Inspection Program includes the above information.
Request:
Amend the USAR supplement to include statements that the Aboveground Steel Tanks Program includes preventive measures to mitigate corrosion by protecting the external surface of steel components per standard industry practice and with sealant or caulking at the interface of concrete and component, if applicable (see RAI B.2.2-3), and verification of the effectiveness of the program by measuring the thickness of the tank bottoms to ensure that significant degradation is not occurring.
- 11 RAI A.1.6-1
Background:
10 CFR 54.21(d) states that the FSAR supplement must contain a summary description of the program and the activities for managing the effects of aging. SRP-LR Rev. 2, Section 3.1.2.5 states that the summary description of the programs and activities for managing the effects of aging for the period of extended operation in the FSAR Supplement should be sufficiently comprehensive such that later changes can be controlled by 10 CFR 50.59, and the description should contain information associated with the bases for determining that aging effects will be managed during the period of extended operation.
The SRP-LR Rev. 2, Table 3.0-1, provides an example FSAR Supplement description of GALL AMP XI,M10 "Boric Acid Corrosion," which includes: (a) visual inspection of external surfaces that are potentially exposed to borated water leakage; (b) timely discovery of leak path and removal of the boric acid residues; (c) assessment of the damage; and (d) follow-up inspection for adequacy. In the USAR Supplement in LRA Section A.1.6, the applicant stated that the Boric Acid Corrosion Program consists of visual inspections.
Issue:
The USAR Supplement does not describe several details of the Boric Acid Corrosion Program that ensure that boric acid corrosion will not lead to degradation on the reactor coolant pressure boundary.
Request:
Revise the USAR supplement to state that the program includes activities associated with discovered evidence of boric acid leakage, including, but not limited to, determination of the principal location of leakage, removal of boric acid residues, and engineering evaluations to establish the impact on the reactor coolant pressure boundary.
ML11167A171 *concurrence via email i
OFFICE LA:DLR/RPB1* PM:DLR/RPB1 BC:DLR/RPB1
. NAME IKing SCuadrado BPham DATE 06/17/2011 06/20/2011 06/20/2011 Letter to Barry S. Allen from Samuel Cuadrado-De Jesus dated June 20, 2011
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE DAVIS-BESSE NUCLEAR POWER STATION (TAC NO.: ME4640)
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