ML022210514
| ML022210514 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 07/30/2002 |
| From: | Mecredy R Rochester Gas & Electric Corp |
| To: | Jack Cushing Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| -nr | |
| Download: ML022210514 (116) | |
Text
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information 3.5 Aging Management of Steam and Power Conversion Systems The results of the aging management review of the Steam and Power Conversion Systems components are provided in this section and summarized in Tables 3.5-1 and 3.5-2. Table 3.5-1 shows the aging management of system components evaluated in NUREG-1 801 that are relied on for license renewal of the Steam and Power Conversion Systems components at Ginna. Included in the table is a discussion column. The discussion column will provide a conclusion indicating if the aging management evaluation results are consistent with NUREG-1 801 along with any clarifications or explanations required to support the stated conclusion if that conclusion is different than those of the NUREG. For a determination to be made that a table line item is "Consistent with NUREG-1 801" several criteria must be met. First the plant specific component is reviewed against the GALL to ensure that the component, materials of construction and internal or external service environment are comparable to those described in a particular GALL item. Second, for those that are comparable, the results of the plant aging management review- aging effect evaluation are compared to the aging effects/mechanisms in the GALL. Finally, the programs credited in the GALL for managing those aging effects are compared to the programs invoked in the plant evaluation. If, using good engineering judgment, it could be reasonably concluded that the plant evaluation is in agreement with the GALL evaluation a line item was considered consistent with NUREG-1 801. There are cases where components and component material/environment combinations and aging effects are common between a NUREG-1 801 line item and the plant evaluation but the aging management program selections differ. In those cases the discussion column will indicate the plant aging management program selection but no conclusion will be made that the line item is consistent with the GALL. Table 3.5-2 contains the Steam and Power Conversion Systems components aging management review results that are not addressed in NUREG-1 801. A plant component is considered not addressed by the NUREG if the component type is not evaluated in the GALL or has a different material of construction or operating environment than evaluated in the GALL. This table includes the component types, materials, environments, aging effects requiring management, the programs and activities for managing aging, and a discussion column. To avoid confusion, no attempt was made to interrelate material/environment/aging effects from one NUREG-1801 chapter to another. Note that these tables only include those components, materials and environments that are applicable to a PWR.
Page 3-194
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Materials The materials of construction of a component have a major influence on the evaluation of aging effects applicable to the component. Sources of information used to identify materials of construction include original equipment specifications, vendor technical manuals and drawings, fabrication drawings, piping line specifications, modification design records and field walkdowns/verifications. The tables below account for the materials of construction for the components requiring an aging management review.
Since similar materials are susceptible to the same aging effects/mechanisms, the tables itemize the component types (i.e., groupings) while factoring in the materials of construction.
Environment As previously described, the environment(s) to which components are exposed are critical in the determination of potential aging mechanisms and effects. A review of plant design documentation was performed to quantify the environmental conditions to which Ginna Station equipment is exposed. This review identified that some equipment is exposed to a variety of environments. This can include normal operating conditions and post accident conditions. Since aging mechanisms and effects will be primarily driven by the environmental conditions to which equipment is exposed on a daily basis, under normal operating conditions, these conditions will differ from the design parameters which are established based upon the worst case scenario (e.g., LOCA conditions). Ginna Station equipment environments may be categorized into basic external and internal environments detailed in Section 3.1.2.
Aging Effects Requiring Management After the components requiring aging management review were identified and grouped by materials of construction and environment, a review of industry and plant-specific operating experience was performed. The purpose of this review was to assure that all applicable aging effects were identified, and to evaluate the effectiveness of existing aging management programs.
This experience review was performed utilizing various industry and plant-specific programs and databases. Industry operating experience sources included NRC Generic Publications (including Information Notices, Circulars, Bulletins, and Generic Letters),
INPO Significant Operating Event Reports (SOER), EPRI Technical Reports, and other information sources, such as the B&W Owners Group Non-Class 1 Mechanical Tools Implementation document, Westinghouse Generic Technical Reports (GTRs), and the Generic Aging Lessons Learned (GALL) report.
Page 3-195
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Plant specific operating experience sources included Semi-annual and Annual Reports to AEC/NRC, Abnormal Occurrence and Licensee Event Reports (LERs),
Non-Conformance Reports (NCRs), Corrective Action Reports (CARs), Refueling, Inspection and Overhaul Reports (RIOs), Inservice Inspection (ISI) Reports, Identified Deficiency Reports (IDRs), and ACTION Reports (ARs) from 1969 to the present.
Information from these sources was compiled in various databases. Based upon the material of construction, the applicable environments, and operating experience the potential aging effects requiring management for each of the components was identified as documented in the tables below.
Time-Limited Aging Analysis In addition to those identified in NUREG-1 801, any additional time-limited aging analyses (TLAA) identified as appropriate to the system are identified in Section 4.0.
Conclusion The programs and activities selected to manage the aging effects of the Steam and Power Conversion Systems are identified in Table 3.5-1 and Table 3.5-2. A description of these aging management activities is provided in Appendix B, along with the demonstration that the identified aging effects will be managed for the period of extended operation. Therefore, based on the demonstrations provided in Appendix B, the effects of aging associated with the system components will be adequately managed so that there is reasonable assurance that the intended function(s) will be maintained consistent with the current licensing basis during the pedod of extended operation.
Page 3-196
Table 3.5-1 Steam and Power Conversion Systems - Aging Management Programs Evaluated in NUREG-1801 that are Relied on for License Renewal Aging Further Aging Management Evaluation Component Effect/Mechanism Programs Recommended Discussion (1) Piping and fittings in Cumulative fatigue TLAA, evaluated in Yes, TLAA Consistent with NUREG-1801. Cumulative Fatigue Damage is main feedwater line, steam damage accordance with 10 addressed as a TLAA in Section 4.3.
line and AFW piping (PWR CFR 54.21(c) only)
(2) Piping and fittings, Loss of material Water chemistry Yes, detection of The Periodic Surveillance and Preventive Maintenance valve bodies and bonnets, due to general and one-time aging effects is to Program will be used to verify the effectiveness of the Water pump casings, tanks, tubes, (carbon steel only), inspection be further Chemistry Control Program.
pitting, and crevice evaluated tubesheets, channel head corrosion and shell (except main steam system)
(3) Auxiliary feedwater Loss of material Plant specific Yes, plant specific The combination of components, materials and environments (AFW) piping due to general, identified in Item VIII.G.1 -d are evaluated in the Service Water pitting, and crevice System. The Service Water System components are reviewed corrosion, MIC, and under NUREG-1801 Chapter VII (Auxiliary Systems), Section biofouling C1.
(4) Oil coolers in AFW Loss of material Plant specific Yes, plant specific Consistent with NUREG-1 801. The Periodic Surveillance and system (lubricating oil side due to general Preventive Maintenance Program is credited with managing all possibly contaminated with (carbon steel only), applicable aginq effects. Other component types such as water) pitting, and crevice accumulators, filter housings, orifices, piping, speed corrosion and MIC increasers, tanks, and valve bodies have been included in this line item at Ginna Station. Although these specific component types were not included in the NUREG section, the material, environment, aging effect/mechanism, and aging management program are consistent.
(5) External surface of Loss of material Plant specific Yes, plant specific Consistent with NUREG-1 801. The Systems Monitoring carbon steel components due to general Program is credited with managing the aging effect "loss of corrosion material due to general corrosion" on the external surfaces of carbon steel components.
Page 3-197 Application for Renewed Operating License
Table 3.5-1 Steam and Power Conversion Systems - Aging Management Programs Evaluated in NUREG-1801 that are Relied on for License Renewal Aging Further Aging Management Evaluation Component Effect/Mechanism Programs Recommended Discussion (6) Carbon steel piping and Wall thinning due to Flow-accelerated No Consistent with NU REG-1 801. The Flow-Accelerated valve bodies flow-accelerated corrosion Corrosion Program implements the guidelines of EPRI corrosion NSAC-202L and is credited with managing the aging effect "wall thinning due to flow-accelerated corrosion."
(7) Carbon steel piping and Loss of material Water chemistry No Consistent with NUREG-1801. The Water Chemistry Control valve bodies in main steam due to pitting and Program is credited with managing the aging effects "loss of system crevice corrosion material due to pitting and crevice corrosion" for components within the Main Steam System.
(8) Closure bolting in Loss of material Bolting integrity No Consistent with NUREG-1801. The Bolting Integrity Program is high-pressure or due to general credited for managing the aging effects "loss of material due to high-temperature systems corrosion; crack general corrosion; crack initiation and growth due to cyclic initiation and loading and/or SCC." There are no bolts with a specified growth due to cyclic minimum yield strength > 150 ksi in the Steam and Power loading and/or SCC Conversion Systems. Therefore, SCC is not an applicable aging effect/mechanism.
(9) Heat exchangers and Loss of material Open-cycle cooling No The Periodic Surveillance and Preventive Maintenance coolers/ condensers due to general water system Program will be credited with managing the applicable aging serviced by open-cycle (carbon steel only), effects in lieu of the Open-Cycle Cooling (Service) Water pitting, and crevice System Program.
cooling water corrosion, MIC, and biofouling; buildup of deposit due to biofouling (10) Heat exchangers and Loss of material Closed-cycle No There are no heat exchangers in the Steam and Power coolers/ condensers due to general cooling water Conversion Systems that are serviced by closed-cycle cooling serviced by closed-cycle (carbon steel only), system water at Ginna Station.
cooling water pitting, and crevice corrosion Page 3-198 Application for Renewed Operating License
Table 3.5-1 Steam and Power Conversion Systems - Aging Management Programs Evaluated in NUREG-1801 that are Relied on for License Renewal Aging Further Aging Management Evaluation Component Effect/Mechanism Programs Recommended Discussion (11) External surface of Loss of material Above ground No The above ground condensate storage tank at Ginna Station is above ground condensate due to general carbon steel tanks not in scope to License Renewal and therefore not subject to storage tank (carbon steel only), an aging management review.
pitting, and crevice corrosion (12) External surface of Loss of material Buried piping and No There are no buried tanks or piping in the Steam and Power buried condensate storage due to general, tanks surveillance Conversion Systems at Ginna Station.
tank and AFW piping pitting, and crevice corrosion and MIC or Buried piping and Yes, detection of tanks inspection aging effects and operating experience are to be further evaluated (13) External surface of Loss of material Boric acid corrosion No Consistent with NURGE 1801. The Boric Acid Corrosion carbon steel components due to boric acid Program is credited with managing the aging effect "loss of corrosion material due to boric acid corrosion."
Page 3-199 Application for Renewed Operating License
Table 3.5-2 Steam and Power Conversion Systems - Component Types Subject to Aging Management not Evaluated in NUREG-1801 Component Types Material Environment AERMs Program/Activity Discussion (1) CONTROLLER 1 Stainless Steel Indoor (No Air No Aging Effects No Aging Material and environment grouping are not Conditioning) Management included in NUREG-1801.
Program Required (2) CONTROLLER 1 Stainless Steel Treated Water Loss of Material One-Time Material and environment grouping are not Secondary Inspection included in NUREG-1801. The aging management (Stagnant) <120 Program program(s) referenced are appropriate for the aging effects identified and provides assurance that the aging effects are effectively managed through the period of extended operation.
One-time inspections are used to verify the effectiveness of the Water Chemistry Control Program.
(3) CONTROLLER 1 Stainless Steel Treated Water Loss of Material Water Chemistry Material and environment grouping are not Secondary Control Program included in NUREG-1801. The aging management (Stagnant) <120 program(s) referenced are appropriate for the aging effects identified and provides assurance that the aging effects are effectively managed through the period of extended operation.
(4) CONVERTER 1 Stainless Steel Air and Gas No Aging Effects No Aging Material and environment grouping are not Management included in NUREG-1801.
Program Required (5) CONVERTER 1 Stainless Steel Indoor (No Air No Aging Effects No Aging Material and environment grouping are not Conditioning) Management included in NUREG-1801.
Program Required (6) COOLER Stainless Steel Indoor (No Air No Aging Effects No Aging Material and environment grouping are not Conditioning) Management included in NUREG-1801.
Program Required (7) FASTENERS Carbon/Low Alloy Indoor (No Air Cracking due to Bolting Integrity There are no bolts with a specified minimum yield (BOLTING) Steel Conditioning) SCC Program strength > 150 ksi in this system. Therefore, SCC is not an applicable aging effect/mechanism.
Page 3-200 Application for Renewed Operating License
Table 3.5-2 Steam and Power Conversion Systems - Component Types Subject to Aging Management not Evaluated in NUREG-1801 Component Types Material Environment AERMs Program/Activity Discussion (8) FASTENERS Carbon/Low Alloy Indoor (No Air Loss of Preload Bolting Integrity Material and environment grouping are included in (BOLTING) Steel Conditioning) due to Stress Program NUREG-1801. Aging effect of loss of preload due Relaxation to stress relaxation is applicable, but is not included in Chapter V - Section E, Chapter VII Section I, or Chapter VIII - Section H of the NUREG.
(9) FASTENERS Stainless Steel Borated Water No Aging Effects No Aging Material and environment grouping are not (BOLTING) Leaks Management included in NUREG-1801.
Program Required (10) FILTER Stainless Steel Indoor (No Air No Aging Effects No Aging Material and environment grouping are not HOUSING Conditioning) Management included in NUREG-1801.
Program Required (11) FLOW Stainless Steel Containment No Aging Effects No Aging Material and environment grouping are not ELEMENT Management included in NUREG-1801.
Program Required (12) FLOW Stainless Steel Indoor (No Air No Aging Effects No Aging Material and environment grouping are not ELEMENT Conditioning) Management included in NUREG-1801.
Program Required (13) FLOW Stainless Steel Treated Water Loss of Material One-Time Material and environment grouping are not ELEMENT Secondary Inspection included in NUREG-1801.The aging management (Stagnant) <120 Program program(s) referenced are appropriate for the aging effects identified and provides assurance that the aging effects are effectively managed through the period of extended operation.
One-time inspections are used to verify the effectiveness of the Water Chemistry Control Program.
(14) FLOW Stainless Steel Treated Water Loss of Material Water Chemistry Material and environment grouping are not ELEMENT Secondary Control Program included in NUREG-1801. The aging management (Stagnant) <120 program(s) referenced are appropriate for the aging effects identified and provides assurance that the aging effects are effectively managed through the period of extended operation.
Page 3-201 Application for Renewed Operating License
Table 3.5-2 Steam and Power Conversion Systems - Component Types Subject to Aging Management not Evaluated in NUREG-1801 Component Types Material Environment AERMs Program/Activity Discussion (15) FLOW Stainless Steel Treated Water Cracking due to Water Chemistry Material and environment grouping are not ELEMENT Secondary >120 SCC Control Program included in NUREG-1801. The aging management program(s) referenced are appropriate for the aging effects identified and provides assurance that the aging effects are effectively managed through the period of extended operation.
(16) FLOW Stainless Steel Treated Water Loss of Material Water Chemistry Material and environment grouping are not ELEMENT Secondary >120 Control Program included in NUREG-1801. The aging management program(s) referenced are appropriate for the aging effects identified and provides assurance that the aging effects are effectively managed through the period of extended operation.
(17) HEAT Cast Iron Indoor (No Air Loss of Material Systems Material and environment grouping are not EXCHANGER Conditioning) Monitoring included in NUREG-1801. The aging management Program program(s) referenced are appropriate for the aging effects identified and provides assurance that the aging effects are effectively managed through the period of extended operation.
(18) HEAT HX-Cast Iron 2 Oil and Fuel Oil Loss of Heat Periodic Material and environment grouping are not EXCHANGER Transfer Surveillance and included in NUREG-1801. The aging management Preventive program(s) referenced are appropriate for the Maintenance aging effects identified and provides assurance Program that the aging effects are effectively managed through the period of extended operation.
(19) HEAT Cast Iron Oil and Fuel Oil Loss of Material Periodic Material and environment grouping are not EXCHANGER Surveillance and included in NUREG-1801. The aging management Preventive program(s) referenced are appropriate for the Maintenance aging effects identified and provides assurance Program that the aging effects are effectively managed through the period of extended operation.
Page 3-202 Application for Renewed Operating License
Table 3.5-2 Steam and Power Conversion Systems - Component Types Subject to Aging Management not Evaluated in NUREG-1801 Component Types Material Environment AERMs Program/Activity Discussion (20) HEAT HX-Cast Iron 2 Raw Water Loss of Heat One-Time Material and environment grouping are not EXCHANGER Transfer Inspection included in NUREG-1801. The aging management Program program(s) referenced are appropriate for the aging effects identified and provides assurance that the aging effects are effectively managed through the period of extended operation.
(21) HEAT Cast Iron Raw Water Loss of Material One-Time Material and environment grouping are not EXCHANGER Inspection included in NUREG-1 801. The aging management Program program(s) referenced are appropriate for the aging effects identified and provides assurance that the aging effects are effectively managed through the period of extended operation.
(22) LEVEL GLASS Aluminum Indoor (No Air No Aging Effects No Aging Material and environment grouping are not Conditioning) Management included in NUREG-1 801.
Program Required (23) LEVEL GLASS Aluminum Oil and Fuel Oil Loss of Material Periodic Material and environment grouping are not Surveillance and included in NUREG-1801. The aging management Preventive program(s) referenced are appropriate for the Maintenance aging effects identified and provides assurance Program that the aging effects are effectively managed through the period of extended operation.
(24) OPERATOR Carbon/Low Alloy Air and Gas No Aging Effects No Aging Material and environment grouping are not Steel Management included in NUREG-1801.
Program Required (25) ORIFICE Stainless Steel Indoor (No Air No Aging Effects No Aging Material and environment grouping are not Conditioning) Management included in NUREG-1801.
Program Required Page 3-203 Application for Renewed Operating License
Table 3.5-2 Steam and Power Conversion Systems - Component Types Subject to Aging Management not Evaluated in NUREG-1801 Component Types Material Environment AERMs Program/Activity Discussion (26) ORIFICE Stainless Steel Treated Water Loss of Material One-Time Material and environment grouping are not Secondary Inspection included in NUREG-1801. The aging management (Stagnant) <120 Program program(s) referenced are appropriate for the aging effects identified and provides assurance that the aging effects are effectively managed through the period of extended operation.
One-time inspections are used to verify the effectiveness of the Water Chemistry Control Program.
(27) ORIFICE Stainless Steel Treated Water Loss of Material Water Chemistry Material and environment grouping are not Secondary Control Program included in NUREG-1801. The aging management (Stagnant) <120 program(s) referenced are appropriate for the aging effects identified and provides assurance that the aging effects are effectively managed through the period of extended operation.
(28) PIPE Carbon/Low Alloy Air and Gas No Aging Effects No Aging Material and environment grouping are not Steel Management included in NUREG-1801.
Program Required (29) PIPE Carbon/Low Alloy Air and Gas Loss of Material One-Time Material and environment grouping are not Steel (Wetted) < 140 Inspection included in NUREG-1801. The aging management Program program(s) referenced are appropriate for the aging effects identified and provides assurance that the aging effects are effectively managed through the period of extended operation.
(30) PIPE Copper Alloy (Zn Indoor (No Air No Aging Effects No Aging Material and environment grouping are not
< 15%) Conditioning) Management included in NUREG-1801.
Program Required Page 3-204 Application for Renewed Operating License
Table 3.5-2 Steam and Power Conversion Systems - Component Types Subject to Aging Management not Evaluated in NUREG-1801 Component Types Material Environment AERMs Program/Activity Discussion (31) PIPE Copper Alloy (Zn Treated Water Loss of Material One-Time Material and environment grouping are not
< 15%) Secondary Inspection included in NUREG-1801. The aging management (Stagnant) <120 Program program(s) referenced are appropriate for the aging effects identified and provides assurance that the aging effects are effectively managed through the period of extended operation.
One-time inspections are used to verify the effectiveness of the Water Chemistry Control Program.
(32) PIPE Copper Alloy (Zn Treated Water Loss of Material Water Chemistry Material and environment grouping are not
< 15%) Secondary Control Program included in NUREG-1801. The aging management (Stagnant) <120 program(s) referenced are appropriate for the aging effects identified and provides assurance that the aging effects are effectively managed through the period of extended operation.
(33) PIPE Stainless Steel Air and Gas No Aging Effects No Aging Material and environment grouping are not Management included in NUREG-1801.
Program Required (34) PIPE Stainless Steel Containment No Aging Effects No Aging Material and environment grouping are not Management included in NUREG-1801.
Program Required (35) PIPE Stainless Steel Indoor (No Air No Aging Effects No Aging Material and environment grouping are not Conditioning) Management included in NUREG-1801.
Program Required (36) PIPE Stainless Steel Treated Water Loss of Material One-Time Material and environment grouping are not Secondary Inspection included in NUREG-1801. The aging management (Stagnant) <120 Program program(s) referenced are appropriate for the aging effects identified and provides assurance that the aging effects are effectively managed through the period of extended operation.
One-time inspections are used to verify the effectiveness of the Water Chemistry Control Program.
Page 3-205 Application for Renewed Operating License
Table 3.5-2 Steam and Power Conversion Systems - Component Types Subject to Aging Management not Evaluated in NUREG-1801 Component Types Material Environment AERMs Program/Activity Discussion (37) PIPE Stainless Steel Treated Water Loss of Material Water Chemistry Material and environment grouping are not Secondary Control Program included in NUREG-1801. The aging management (Stagnant) <120 program(s) referenced are appropriate for the aging effects identified and provides assurance that the aging effects are effectively managed through the period of extended operation.
(38) PIPE Stainless Steel Treated Water Cracking due to One-Time Material and environment grouping are not Secondary >120 SCC Inspection included in NUREG-1801. The aging management Program program(s) referenced are appropriate for the aging effects identified and provides assurance that the aging effects are effectively managed through the period of extended operation.
One-time inspections are used to verify the effectiveness of the Water Chemistry Control Program.
(39) PIPE Stainless Steel Treated Water Cracking due to Water Chemistry Material and environment grouping are not Secondary >120 SCC Control Program included in NUREG-1801. The aging management program(s) referenced are appropriate for the aging effects identified and provides assurance that the aging effects are effectively managed through the period of extended operation.
(40) PIPE Stainless Steel Treated Water Loss of Material One-Time Material and environment grouping are not Secondary >120 Inspection included in NUREG-1801. The aging management Program program(s) referenced are appropriate for the aging effects identified and provides assurance that the aging effects are effectively managed through the period of extended operation.
One-time inspections are used to verify the effectiveness of the Water Chemistry Control Program.
Page 3-206 Application for Renewed Operating License
Table 3.5-2 Steam and Power Conversion Systems - Component Types Subject to Aging Management not Evaluated in NUREG-1801 Component Types Material Environment AERMs Program/Activity Discussion (41) PIPE Stainless Steel Treated Water Loss of Material Water Chemistry Material and environment grouping are not Secondary >120 Control Program included in NUREG-1801. The aging management program(s) referenced are appropriate for the aging effects identified and provides assurance that the aging effects are effectively managed through the period of extended operation.
(42) POSITIONER 1 Stainless Steel Air and Gas No Aging Effects No Aging Material and environment grouping are not Management included in NUREG-1801.
Program Required 1 Indoor (No Air No Aging Effects No Aging Material and environment grouping are not (43) POSITIONER Stainless Steel Conditioning) Management included in NUREG-1801.
Program Required (44) PRESSURE Aluminum Air and Gas No Aging Effects No Aging Material and environment grouping are not RELAY1 Management included in NUREG-1801.
Program Required (45) PRESSURE Aluminum Indoor (No Air No Aging Effects No Aging Material and environment grouping are not RELAY1 Conditioning) Management included in NUREG-1801.
Program Required (46) PUMP CASING Cast Iron Indoor (No Air Loss of Material Systems Material and environment grouping are not Conditioning) Monitoring included in NUREG-1801.
Program (47) PUMP CASING Cast Iron Oil and Fuel Oil Loss of Material Periodic Material and environment grouping are not Surveillance and included in NUREG-1801. The aging management Preventive program(s) referenced are appropriate for the Maintenance aging effects identified and provides assurance Program that the aging effects are effectively managed through the period of extended operation.
(48) SCREEN Stainless Steel Air and Gas No Aging Effects No Aging Material and environment grouping are not (Wetted) <140 Management included in NUREG-1801.
Program Required Page 3-207 Application for Renewed Operating License
Table 3.5-2 Steam and Power Conversion Systems - Component Types Subject to Aging Management not Evaluated in NUREG-1801 Component Types Material Environment AERMs Program/Activity Discussion (49) SCREEN Stainless Steel Indoor (No Air No Aging Effects No Aging Material and environment grouping are not Conditioning) Management included in NUREG-1801.
Program Required (50) TANK Neoprene Indoor (No Air Change in Periodic Material and environment grouping are not Conditioning) Material Surveillance and included in NUREG-1801. The aging management Properties and Preventive program(s) referenced are appropriate for the Cracking Maintenance aging effects identified and provides assurance Program that the aging effects are effectively managed through the period of extended operation.
(51) TEMPERATURE Stainless Steel Indoor (No Air No Aging Effects No Aging Material and environment grouping are not ELEMENT 1 Conditioning) Management included in NUREG-1801.
Program Required (52) TEMPERATURE Stainless Steel Treated Water Cracking due to One-Time Material and environment grouping are not ELEMENT1 Secondary >120 SCC Inspection included in NUREG-1801. The aging management Program program(s) referenced are appropriate for the aging effects identified and provides assurance that the aging effects are effectively managed through the period of extended operation.
One-time inspections are used to verify the effectiveness of the Water Chemistry Control Program.
(53) TEMPERATURE Stainless Steel Treated Water Cracking due to Water Chemistry Material and environment grouping are not ELEMENT1 Secondary >120 SCC Control Program included in NUREG-1801. The aging management program(s) referenced are appropriate for the aging effects identified and provides assurance that the aging effects are effectively managed through the period of extended operation.
Page 3-208 Application for Renewed Operating License
Table 3.5-2 Steam and Power Conversion Systems - Component Types Subject to Aging Management not Evaluated in NUREG-1801 Component Types Material Environment AERMs Program/Activity Discussion (54) TEMPERATURE Stainless Steel Treated Water Loss of Material One-Time Material and environment grouping are not ELEMENT 1 Secondary >120 Inspection included in NUREG-1801. The aging management Program program(s) referenced are appropriate for the aging effects identified and provides assurance that the aging effects are effectively managed through the period of extended operation.
One-time inspections are used to verify the effectiveness of the Water Chemistry Control Program.
(55) TEMPERATURE Stainless Steel Treated Water Loss of Material Water Chemistry Material and environment grouping are not ELEMENT 1 Secondary >120 Control Program included in NUREG-1801. The aging management program(s) referenced are appropriate for the aging effects identified and provides assurance that the aging effects are effectively managed through the period of extended operation.
(56) VALVE BODY Aluminum Air and Gas No Aging Effects No Aging Material and environment grouping are not Management included in NUREG-1801.
Program Required (57) VALVE BODY Aluminum Indoor (No Air No Aging Effects No Aging Material and environment grouping are not Conditioning) Management included in NUREG-1801.
Program Required (58) VALVE BODY Carbon/Low Alloy Air and Gas No Aging Effects No Aging Material and environment grouping are not Steel Management included in NUREG-1801.
Program Required (59) VALVE BODY Cast Austenitic Indoor (No Air No Aging Effects No Aging Material and environment grouping are not Stainless Steel Conditioning) Management included in NUREG-1801.
Program Required Page 3-209 Application for Renewed Operating License
Table 3.5-2 Steam and Power Conversion Systems - Component Types Subject to Aging Management not Evaluated in NUREG-1801 Component Types Material Environment AERMs Program/Activity Discussion (60) VALVE BODY Cast Austenitic Treated Water Loss of Material One-Time Material and environment grouping are not Stainless Steel Secondary Inspection included in NUREG-1801. The aging management (Stagnant) <120 Program program(s) referenced are appropriate for the aging effects identified and provides assurance that the aging effects are effectively managed through the period of extended operation.
One-time inspections are used to verify the effectiveness of the Water Chemistry Control Program.
(61) VALVE BODY Cast Austenitic Treated Water Loss of Material Water Chemistry Material and environment grouping are not Stainless Steel Secondary Control Program included in NUREG-1801. The aging management (Stagnant) <120 program(s) referenced are appropriate for the aging effects identified and provides assurance that the aging effects are effectively managed through the period of extended operation.
(62) VALVE BODY Copper Alloy (Zn Air and Gas No Aging Effects No Aging Material and environment grouping are not
< 15%) Management included in NUREG-1801.
Program Required (63) VALVE BODY Copper Alloy (Zn Indoor (No Air No Aging Effects No Aging Material and environment grouping are not
< 15%) Conditioning) Management included in NUREG-1801.
Program Required (64) VALVE BODY Copper Alloy (Zn Oil and Fuel Oil Loss of Material Periodic Material and environment grouping are not
< 15%) Surveillance and included in NUREG-1801. The aging management Preventive program(s) referenced are appropriate for the Maintenance aging effects identified and provides assurance Program that the aging effects are effectively managed through the period of extended operation.
Page 3-210 Application for Renewed Operating License
Table 3.5-2 Steam and Power Conversion Systems - Component Types Subject to Aging Management not Evaluated in NUREG-1801 Component Types Material Environment AERMs Program/Activity Discussion (65) VALVE BODY Copper Alloy (Zn Treated Water Loss of Material One-Time Material and environment grouping are not
< 15%) Secondary Inspection included in NUREG-1801. The aging management (Stagnant) <120 Program program(s) referenced are appropriate for the aging effects identified and provides assurance that the aging effects are effectively managed through the period of extended operation.
One-time inspections are used to verify the effectiveness of the Water Chemistry Control Program.
(66) VALVE BODY Copper Alloy (Zn Treated Water Loss of Material Water Chemistry Material and environment grouping are not
< 15%) Secondary Control Program included in NUREG-1801. The aging management (Stagnant) <120 program(s) referenced are appropriate for the aging effects identified and provides assurance that the aging effects are effectively managed through the period of extended operation.
(67) VALVE BODY Stainless Steel Air and Gas No Aging Effects No Aging Material and environment grouping are not Management included in NUREG-1801.
Program Required (68) VALVE BODY Stainless Steel Containment No Aging Effects No Aging Material and environment grouping are not Management included in NUREG-1801.
Program Required (69) VALVE BODY Stainless Steel Indoor (No Air No Aging Effects No Aging Material and environment grouping are not Conditioning) Management included in NUREG-1801.
Program Required (70) VALVE BODY Stainless Steel Treated Water Loss of Material One-Time Material and environment grouping are not Secondary Inspection included in NUREG-1801. The aging management (Stagnant) <120 Program program(s) referenced are appropriate for the aging effects identified and provides assurance that the aging effects are effectively managed through the period of extended operation.
One-time inspections are used to verify the effectiveness of the Water Chemistry Control Program.
Page 3-211 Application for Renewed Operating License
Table 3.5-2 Steam and Power Conversion Systems - Component Types Subject to Aging Management not Evaluated in NUREG-1801 Component Types Material Environment AERMs Program/Activity Discussion (71) VALVE BODY Stainless Steel Treated Water Loss of Material Water Chemistry Material and environment grouping are not Secondary Control Program included in NUREG-1801. The aging management (Stagnant) <120 program(s) referenced are appropriate for the aging effects identified and provides assurance that the aging effects are effectively managed through the period of extended operation.
(72) VALVE BODY Stainless Steel Treated Water Cracking due to One-Time Material and environment grouping are not Secondary >120 SCC Inspection included in NUREG-1801. The aging management Program program(s) referenced are appropriate for the aging effects identified and provides assurance that the aging effects are effectively managed through the period of extended operation.
One-time inspections are used to verify the effectiveness of the Water Chemistry Control Program.
(73) VALVE BODY Stainless Steel Treated Water Cracking due to Water Chemistry Material and environment grouping are not Secondary >120 SCC Control Program included in NUREG-1801. The aging management program(s) referenced are appropriate for the aging effects identified and provides assurance that the aging effects are effectively managed through the period of extended operation.
(74) VALVE BODY Stainless Steel Treated Water Loss of Material One-Time Material and environment grouping are not Secondary >120 Inspection included in NUREG-1801. The aging management Program program(s) referenced are appropriate for the aging effects identified and provides assurance that the aging effects are effectively managed through the period of extended operation.
One-time inspections are used to verify the effectiveness of the Water Chemistry Control Program.
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Table 3.5-2 Steam and Power Conversion Systems - Component Types Subject to Aging Management not Evaluated in NUREG-1801 Component Types Material Environment AERMs Program/Activity Discussion (75) VALVE BODY Stainless Steel Treated Water Loss of Material Water Chemistry Material and environment grouping are not Secondary >120 Control Program included in NUREG-1801. The aging management program(s) referenced are appropriate for the aging effects identified and provides assurance that the aging effects are effectively managed through the period of extended operation.
- 1. Selected instruments were conservatively included within the scope of License Renewal. Consideration was given to the consequences of an instrument housing pressure boundary failure. Where an instrument was unisolable from a pressure source and is of sufficient size that a system function would be degraded should the pressure boundary fail, that instrument is included for License Renewal review.
- 2. Material prefixes with HX are used to identify heat exchanger materials which perform a heat transfer intended function in addition to the typical material usage function of pressure boundary.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information 3.6 Aging Management of Structures and Component Supports The results of the aging management review of the Structures and Component Supports are provided in this section and summarized in Table 3.6-1 and Table 3.6-2. Table 3.6-1 shows the aging management of system components evaluated in NUREG-1 801 that are relied on for license renewal of the Structures and Component Supports at Ginna.
Included in the table is a discussion column. The discussion column will provide a conclusion indicating if the aging management evaluation results are consistent with NUREG-1 801 along with any clarifications or explanations required to support the stated conclusion if that conclusion is different than those of the NUREG. For a determination to be made that a table line item is "Consistent with NUREG-1 801" several criteria must be met. First the plant specific component is reviewed against the GALL to ensure that the component, materials of construction and internal or external service environment are comparable to those described in a particular GALL item. Second, for those that are comparable, the results of the plant aging management review- aging effect evaluation are compared to the aging effects/mechanisms in the GALL. Finally, the programs credited in the GALL for managing those aging effects are compared to the programs invoked in the plant evaluation. If, using good engineering judgment, it could be reasonably concluded that the plant evaluation is in agreement with the GALL evaluation a line item was considered consistent with NUREG-1 801. There are cases where components and component material/environment combinations and aging effects are common between a NUREG-1801 line item and the plant evaluation but the aging management program selections differ. In those cases the discussion column will indicate the plant aging management program selection but no conclusion will be made that the line item is consistent with the GALL. Table 3.6-2 contains the Structures and Component Supports aging management review results that are not addressed in NUREG-1 801. A plant component is considered not addressed by the NUREG if the component type is not evaluated in the GALL or has a different material of construction or operating environment than evaluated in the GALL. This table includes the component types, materials, environments, aging effects requiring management, the programs and activities for managing aging, and a discussion column. To avoid confusion, no attempt was made to interrelate material/environment/aging effects from one NUREG-1801 chapter to another. Note that these tables only include those components, materials and environments that are applicable to a PWR.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Materials The materials of construction of a component have a major influence on the evaluation of aging effects applicable to the component. Sources of information used to identify materials of construction include original equipment specifications, vendor technical manuals and drawings, fabrication drawings, piping line specifications, modification design records and field walkdowns/verifications. The tables below account for the materials of construction for the components requiring an aging management review.
Since similar materials are susceptible to the same aging effects/mechanisms, the tables itemize the component types (i.e., groupings) while factoring in the materials of construction.
Environment As previously described, the environment(s) to which components are exposed are critical in the determination of potential aging mechanisms and effects. A review of plant design documentation was performed to quantify the environmental conditions to which Ginna Station equipment is exposed. This review identified that some equipment is exposed to a variety of environments. This can include normal operating conditions and post accident conditions. Since aging mechanisms and effects will be primarily driven by the environmental conditions to which equipment is exposed on a daily basis, under normal operating conditions, these conditions will differ from the design parameters which are established based upon the worst case scenario (e.g., LOCA conditions). Ginna Station equipment environments may be categorized into basic external and internal environments detailed in Section 3.1.2.
Aging Effects Requiring Management After the components requiring aging management review were identified and grouped by materials of construction and environment, a review of industry and plant-specific operating experience was performed. The purpose of this review was to assure that all applicable aging effects were identified, and to evaluate the effectiveness of existing aging management programs.
This experience review was performed utilizing various industry and plant-specific programs and databases. Industry operating experience sources included NRC Generic Publications (including Information Notices, Circulars, Bulletins, and Generic Letters),
INPO Significant Operating Event Reports (SOER), EPRI Technical Reports, and other information sources, such as the B&W Owners Group Non-Class 1 Mechanical Tools Implementation document, Westinghouse Generic Technical Reports (GTRs), and the Generic Aging Lessons Learned (GALL) report.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Plant specific operating experience sources included Semi-annual and Annual Reports to AEC/NRC, Abnormal Occurrence and Licensee Event Reports (LERs),
Non-Conformance Reports (NCRs), Corrective Action Reports (CARs), Refueling, Inspection and Overhaul Reports (RIOs), Inservice Inspection (ISI) Reports, Identified Deficiency Reports (IDRs), and ACTION Reports (ARs) from 1969 to the present.
Information from these sources was compiled in various databases. Based upon the material of construction, the applicable environments, and operating experience the potential aging effects requiring management for each of the components was identified as documented in the tables below.
Time-Limited Aging Analysis In addition to those identified in NUREG-1 801, any additional time-limited aging analyses (TLAA) identified as appropriate to the system are identified in Section 4.0.
Confirmation of Topical Report Applicability Containment Structures The Westinghouse Owners' Group Life Cycle Management & License Renewal Program has prepared topical report, WCAP-14756-A, Aging Management Evaluation for Pressurized Water Reactor Containment Structure (Reference 1), which has been utilized in the aging management review of the Ginna Containment Structures. Therefore, reconciliation of the final SER for WCAP-1 4756-A applicant action items is provided in Table 3.6.0-1. A description of aging management activities is provided in Appendix B, along with the demonstration that the identified aging effects will be managed for the period of extended operation.
Therefore, based on the demonstrations provided in Appendix B, the effects of aging associated with the Containment Structures will be adequately managed so that there is reasonable assurance that the intended function(s) will be maintained consistent with the current licensing basis during the period of extended operation.
Reactor Coolant System Supports The Westinghouse Owners' Group Life Cycle Management & License Renewal Program has prepared topical report, WCAP-1 4422, Rev. 2-A, License Renewal Evaluation: Aging Management for Reactor Coolant System Supports (Reference 2), which has been utilized in the aging management review of the Ginna RC System Supports components.
The scope of the RC System supports components described in the topical report bounds the Ginna RC System Supports components.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information A reconciliation of the final SER for WCAP-14422 Rev.2-A applicant action items is provided in Table 3.6.0-2. A description of aging management activities is provided in Appendix B, along with the demonstration that the identified aging effects will be managed for the period of extended operation.
Therefore, based on the demonstrations provided in Appendix B, the effects of aging associated with the RC System Supports components will be adequately managed so that there is reasonable assurance that the intended function(s) will be maintained consistent with the current licensing basis during the period of extended operation.
Conclusion The programs and activities selected to manage the aging effects of the Structures and Component Supports are identified in Table 3.6-1 and Table 3.6-2. The results of the applicant action item reviews are also contained in these tables, but in the SRP format.
A description of these aging management activities is provided in Appendix B, along with the demonstration that the identified aging effects will be managed for the period of extended operation.
Therefore, based on the demonstrations provided in Appendix B, the effects of aging associated with the Structures and Component Supports components will be adequately managed so that there is reasonable assurance that the intended function(s) will be maintained consistent with the current licensing basis during the period of extended operation.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 3.6.0-1 Containment Structure - WCAP-14756-A, Final Safety Evaluation Report Response to Applicant Action Items Renewal Applicant Action Item Plant-Specific Response (1) The license renewal applicant will (i) verify The design, configuration, materials of that its plant is bounded by the GTR, (ii) commit construction, and normal operating service to implement programs described as necessary environment of the Ginna Station in the GTR to manage the effects of aging during Containment structure are bounded by the the period of extended operation, and (iii) verify GTR.
that the programs committed to are conducted in accordance with appropriate regulatory controls As part of license renewal, Ginna Station (e.g. 10 CFR Part 50, Appendix B). Further, the aging management programs will implement renewal applicant will identify any deviations the programs described in the GTR as from the aging management programs which necessary to manage the effects of aging this GTR describes as necessary to manage the during the period of extended operation.
effects of aging during the period of extended Further, the programs committed to will be operation or to maintain the functionality of the conducted in accordance with appropriate containment structure, and deviations from other regulatory controls (e.g., 10 CFR Part 50, information presented in the GTR (e.g., Appendix B).
materials of construction). The renewal applicant will evaluate any such deviations in accordance The GTR evaluated aging of the pressurized with 10 CFR 54.21 (a)(3) and (c)(1) on a water reactor containment structure to ensure plant-specific basis. that the intended functions will be maintained during the extended period of operation. Four The following functions, which are specific to intended functions performed by the PWR containment structures and are understood to containment structure are identified in the be covered by the various intended functions, GTR.
should be addressed explicitly in the license renewal application: (1) providing structural or "*Ensuring the integrity of the reactor coolant functional support of safety-related systems, pressure boundary (RCPB) structures, and components following a "*Ensuring the capability to contain a shut design basis accident (DBA); (2) serving as down of the reactor and maintain it in a safe an external missile barrier consistent with the shutdown condition design and licensing basis; and (3) providing "*Ensuring the capability to prevent or mitigate passive heat sinks during a DBA or station the consequences of accidents that could blackout in addition to the spray system. result in potential offsite exposure comparable to the 10 CFR 100 guidelines
"*Ensuring compliance with the USNRC regulations for environmental qualification (10 CFR 50.49).
Three additional intended functions are identified in Action Item Number 1. These additional intended functions are indirectly addressed in the GTR aging management I
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 3.6.0-1 Containment Structure - WCAP-14756-A, Final Safety Evaluation Report Response to Applicant Action Items Renewal Applicant Action Item Plant-Specific Response (1) (continued) evaluation and aging management options since they are a subset of the four given in the GTR. Specifically, each of the additional intended functions are subsets of the following:
"*The additional intended function (1),
providing structural or functional support of safety-related systems, structures, and components following a design basis accident (DBA), is a subset of the GTR intended function 3.
"*The additional intended function (2), serving as an external missile barrier consistent with the design and licensing basis, is a subset of GTR intended functions 3 and 4.
- The additional intended function (3),
providing passive heat sinks during a DBA or station blackout in addition to the spray system, is a subset of GTR intended functions 2 and 3.
Therefore, there is no need to further address the added functions in the license renewal application.
i4 (2) A summary description of the programs and A summary description of aging management activities for managing the effects of aging and programs credited for managing the effects of the evaluation of TLAAs is to be provided in the aging and evaluation of the TLAA's for the license renewal FSAR supplement, in Ginna Station Containment Structure is accordance with 10 CFR 54.21(d). provided in Appendix A of the LRA.
+
(3) Individual plant applicants will need to A comprehensive list of the structures and provide a comprehensive list of structures and components subject to aging management components subject to an aging management review was developed in accordance with the review and the methodology used to develop methodology described in Engineering this list as part of their license renewal Procedure EP-3-S-0713, "Scoping and applications. Any components determined by Screening for License Renewal." This list is the applicant to be subject to an aging available for on-site review.
management review for license renewal but not within the scope of the GTR are required to be addressed in the license renewal application.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 3.6.0-1 Containment Structure - WCAP-14756-A, Final Safety Evaluation Report Response to Applicant Action Items Renewal Applicant Action Item Plant-Specific Response (4) Provide cross-section drawings for the Drawings showing the cross-section of the containment structures, and detailed drawings of Ginna Station Containment structure and the sand pocket region and other plant-specific other specific features, including the sand-box features, if applicable, region, are available at the plant site for review.
(5) Provide legible drawings of equipment and Drawings of equipment and penetration penetration details as part of the description of details of the Ginna Station Containment the containment structure components structure are available at the plant site for review.
(6) For prestressed concrete containments, The wall tendons in the Ginna Station indicate whether the tendon access gallery is Containment structure are permanently included as a containment structure component coupled at the base of the cylinder wall to subject to an aging management review. If it is, rock anchors. The rock anchors consist of provide the details of the aging management 90-wire tendons which were inserted into 6-in.
review and the credited aging management dia. holes drilled 43 feet into base rock, then program. If not, provide a technical basis for its grouted and pre-stressed. The rock anchors exclusion, addressing the potential for were grouted up to the bearing plate for degradation of the lower vertical tendon anchors corrosion protection. The tendon anchorage resulting from the environmental conditions in at the rock-anchor coupling as well as the wall the tendon access gallery, tendons are encased in a steel sheath which is filled with protective grease (paraffin-based mineral oil blended with a micro-crystalline wax). Therefore, there is no lower tendon access gallery at Ginna Station.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 3.6.0-1 Containment Structure - WCAP-14756-A, Final Safety Evaluation Report Response to Applicant Action Items Renewal Applicant Action Item Plant-Specific Response (7) Discuss plant-specific operating experience A thorough review of Ginna Station operating relevant to age-related degradation of experience was conducted. Sources included containment structure components and how this Semi-annual and Annual Reports to experience has been considered in the aging AEC/NRC, Abnormal Occurrence and management review. Licensee Event Reports (LERs),
Non-Conformance Reports (NCRs),
Corrective Action Reports (CARs), Refueling, Inspection and Overhaul Reports (RIOs),
Inservice Inspection (ISI) Reports, Identified Deficiency Reports (IDRs), and ACTION Reports (ARs) from 1969 to the present. In addition, plant-specific response to any NRC generic communication was reviewed for applicability. The results of unique inspections of opportunity and NRC required ASME Section XI, Subsections IWE/IWL inspections were also reviewed. No additional or unique aging effects requiring management were identified from this review beyond those identified in Table 3.6-1 of the LRA.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 3.6.0-1 Containment Structure - WCAP-14756-A, Final Safety Evaluation Report Response to Applicant Action Items Renewal Applicant Action Item Plant-Specific Response i
(8) For concrete containments, verify that the Leaching of Calcium Hydroxide original plant design and construction specifications satisfy the criteria which are relied According to NUREG-1801, leaching of upon to exclude leaching of calcium hydroxide calcium hydroxide from reinforced concrete and reaction with aggregates as significant becomes significant only if the concrete is aging mechanisms. If these mechanisms are not exposed to flowing water. Even if reinforced excluded, describe the aging management concrete is exposed to flowing water, such program (AMP) which is credited to manage the leaching is not significant if the concrete is aging effects associated with these aging constructed to ensure that it is dense, well mechanisms. cured, has low permeability, and that cracking is well controlled. Cracking is controlled through proper arrangement and distribution of reinforcing bars. All of the above characteristics are assured if the concrete was constructed with the guidance of ACI 201.2R-77.
The reinforced concrete of the Containment structure (base mat, ring beam and cylinder walls) is not exposed to flowing water. The only reinforced concrete that is exposed to flowing water is at the Screen House and Discharge Canal. In addition, concrete at Ginna Station was specified in accordance with the guidance of ACI 201.2R.
Reaction with Aggregates Construction of concrete structures at Ginna Station was performed under three contracts:
Gilbert specification, GC-3799, 'Technical Specification for Structural Concrete", Gilbert specification GC-3526, "General Construction Specification Service Building", and Rochester Gas and Electric (RGE)
Corporation specification, "Construction of Screen House, Heating Boiler Room, and Service Water Building Foundations, Walls, Floors to Elevation 253'-6".
Specification GC-3799 states that aggregate shall conform to ACI 301-66 and to the State of New York Department of Public Works Specification latest edition. ACI 301-66 states that aggregate shall conform to ASTM C33.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 3.6.0-1 Containment Structure - WCAP-14756-A, Final Safety Evaluation Report Response to Applicant Action Items Renewal Applicant Action Item Plant-Specific Response (8) (continued) At the time of these contracts, ASTM C33 required that aggregates be tested for potential reactivity in accordance with ASTM C227 and ASTM C295. Additionally, the New York State Department of Transportation requires aggregates to be tested in accordance with ASTM C227 and C295.
Specification GC-3526 states that concrete and concrete work shall conform to ACI 318-63. ACI 318-63 requires aggregate to conform to ASTM C-33. ASTM C33 requires aggregate to be tested in accordance with ASTM C227 and C295. The RGE Corporation specification required aggregates to conform to ASTM C33.
Since the aggregates were tested for potential reactivity in accordance with ASTM C227 and ASTM C295, cracking and expansion or concrete due to reaction with aggregates is not an aging effect requiring management at Ginna Station.
These facts notwithstanding, the current mandated inspections performed in accordance with the requirements of 10 CFR 50.55a and ASME Code Section XI, Subsection IWL, Examination Category L-A, and the Structures Monitoring Program manage the concrete aging effects associated with leaching of calcium hydroxide and reaction with aggregates. Therefore, the potential aging effects associated with leaching of calcium hydroxide and reaction with aggregates will be adequately managed during the period of extended operation in the unlikely event that these mechanisms should become active.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 3.6.0-1 Containment Structure - WCAP-14756-A, Final Safety Evaluation Report Response to Applicant Action Items Renewal Applicant Action Item Plant-Specific Response (9) For concrete containments, discuss whether Elevated temperature was evaluated as an local heating of containment concrete at the aging mechanism for Containment structure main steam and/or any other penetrations concrete components. The temperature of the results in sustained concrete temperatures concrete around hot-piping penetrations such exceeding 200 0 F. If this condition exists, provide as main steam and feedwater is maintained at an aging management review and describe the Ginna Station below the limits in ASME credited aging management program. Section III, Div. 2, Subsection CC-3340, i.e.,
200OF for long term periods during normal operation. These penetrations were designed with a forced air cooling system connected to cooling coils integrated with the penetration sleeves. The cooling coils are in the form of an embossing welded directly to the inner surface of the penetration sleeves. The cooling air exit temperature is monitored and can be related to the concrete -to-sleeve interface temperature (UFSAR Section 3.8.1.5.4). The Penetration Cooling System consists of two full capacity fans located in the Auxiliary Bldg., cooling coils, piping to the appropriate penetration cooling coils, associated valves and instrumentation. The Penetration Cooling System is within the scope of license renewal and is included in the aging management review for the Essential Ventilation System.
The primary shield wall concrete is also subject to extended local heatup at Ginna Station. The purpose of the reactor compartment cooling system is to remove the heat generated by gamma rays in the primary shield and the thermal radiation from the reactor vessel and out-of-core detectors electrical load. Removal of this heat maintains the concrete temperature in the primary shield walls below degradation threshold and localized temperature limits of ACI standards (i.e., 150 0 F). The reactor compartment cooling is also within the scope of license renewal and is included in the aging management review for the Containment Ventilation system.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 3.6.0-1 Containment Structure - WCAP-14756-A, Final Safety Evaluation Report Response to Applicant Action Items Renewal Applicant Action Item Plant-Specific Response (9) (continued) No other concrete components are exposed to elevated temperature. Therefore, elevated temperature is not an aging mechanism that can lead to loss of material for Containment internal structural concrete components.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 3.6.0-1 Containment Structure - WCAP-14756-A, Final Safety Evaluation Report Response to Applicant Action Items Renewal Applicant Action Item Plant-Specific Response (10) Identify the codes, edition and/or date of The design, materials, fabrication, inspection, codes and standards which govern plant and proof testing of the containment complied containment design, inspection and repair. with the applicable parts of the following codes and standards:
- 1. ASME Boiler and Pressure Vessel Code,Section III - Nuclear Vessels;Section VIII Unfired Pressure Vessels;Section IX Welding Qualifications.
- 2. Building Code Requirements for Reinforced Concrete (ACI 318-63).
- 3. American Institute of Steel Construction Specifications:
- a. Specifications for the Design, Fabrication, and Erection of Structural Steel for Buildings, adopted April 17, 1963.
- b. Code of Standard Practice for Steel Buildings and Bridges, revised February 20,1963.
- 4. USAS N 6.2 - 1965, Safety Standard for Design, Fabrication, and Maintenance of Steel Containment Structures for Stationary Nuclear Power Reactors.
- 5. ACI 306-66, Specifications for Structural Concrete for Buildings.
- 6. ASTM C 150-64, Specifications for Portland Cement.
- 7. State of New York Department of Public Works Specification.
- 8. ASTM C 260-63T, Specifications for Air-Entrained Admixtures for Concrete.
- 9. ASTM A 15-64T, Specifications for Billet-Steel Bars for Concrete Reinforcement.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 3.6.0-1 Containment Structure - WCAP-14756-A, Final Safety Evaluation Report Response to Applicant Action Items Renewal Applicant Action Item Plant-Specific Response i
(10) (continued) 10. ASTM A 305-56T, Specifications for Minimum Requirements for Deformation of Deformed Bars for Concrete Reinforcement.
- 11. ASTM A408-64T, Specifications for Special Large Size Deformed Billet-Steel Bars for Concrete Reinforcement.
Additional codes and standards governing design, materials, inspection and repair for the Ginna Station Containment Structure are contained in Section 3.8.1.2.5 of the UFSAR.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 3.6.0-1 Containment Structure - WCAP-14756-A, Final Safety Evaluation Report Response to Applicant Action Items Renewal Applicant Action Item Plant-Specific Response (11) Specify whether freeze-thaw is an According to NUREG-1 801, freeze-thaw does applicable aging mechanism which will be not cause loss of material from reinforced managed by AMP 5.1 or AMP 5.2, as applicable. concrete in foundations and in above- and If not, provide the technical basis for exclusion. below-grade exterior concrete for plants located in a geographic region of negligible weathering conditions (weathering index
<100 day-inch/yr.). Loss of material from such concrete is not significant at plants located in areas where weathering conditions are severe (weathering index >500 day-inch/yr.)
or moderate (100-500 day-inch/yr.) provided that the concrete mix design meets the air content (entrained air 3-6%) and water-to-cement ratio (0.35-0.45) specified in ACI 318-63 or ACI 349-85.
Construction of concrete structures at Ginna Station was performed under three contracts:
Gilbert specification, GC-3799, "Technical Specification for Structural Concrete", Gilbert specification GC-3526, "General Construction Specification Service Building", and Rochester Gas and Electric (RGE)
Corporation specification, "Construction of Screen House, Heating Boiler Room, and Service Water Building Foundations, Walls, Floors to Elevation 253'-6".
Since the contract-specified air contents are within the range specified by current revisions of ACI 318, and the contract-specified water-to-cement ratio meets the recommendations of ACI 318-63 (L0.53),
loss of material and cracking of concrete due to freeze-thaw is not an applicable aging mechanism at Ginna Station.
Nevertheless, the current mandated inspections performed in accordance with the requirements of 10 CFR 50.55a and the ASME Section XI, Subsections IWE & IWL Inservice Inspection Program and the Structures Monitoring Program manage the
_______________________________________________________________ .1.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 3.6.0-1 Containment Structure - WCAP-14756-A, Final Safety Evaluation Report Response to Applicant Action Items Renewal Applicant Action Item Plant-Specific Response (11) (continued) concrete aging effects associated with freeze-thaw.
Therefore, the potential aging effects associated with freeze-thaw will be adequately managed during the period of extended operation in the unlikely event that this mechanisms should become active.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 3.6.0-1 Containment Structure - WCAP-14756-A, Final Safety Evaluation Report Response to Applicant Action Items Renewal Applicant Action Item Plant-Specific Response i
(12) Specify whether aggressive chemical NUREG-1801 provides guidance for attack is an applicable aging mechanism which determining whether aggressive chemical will be managed by AMP 5.3 or AMP 5.4, as attack is an applicable aging mechanism for applicable. If not, provide the technical basis for concrete structures. This guidance includes exclusion. threshold values for the pH of the environment, and chloride and sulfate concentrations. The environments of concern are air (as influenced by rainwater),
below-grade (as influenced by soil or groundwater) and lake water. These three environments are in direct contact with structures and components within the scope of license renewal. According to NUREG-1801, aggressive chemical attack is not significant if the concrete is not exposed to an aggressive environment. An aggressive environment is defined as pH < 5.5, or >500 ppm chlorides, or >1500 ppm sulfates. The below-grade and lake water environments at Ginna Station are therefore classified as non-aggressive.
Chemical analyses of rock and groundwater indicate pH values of 6.9-7.7, and chloride and sulfate concentrations < 50 ppm. Data collected by the National Atmospheric Deposition Program/National Trends Network (NADP/NTN) indicate that Ginna Station is exposed to an aggressive air environment based on analysis of rainwater pH. However, according to the Class I Structures License Renewal Industry Report, concrete used in Class I structures is not likely to be significantly affected by acid rain because of the properties of the concrete. In addition, if the concrete is exposed to an aggressive environment for intermittent periods only, degradation caused by aggressive chemicals is not significant. Rain is an intermittent event.
Therefore, degradation due to aggressive chemical attack by rainwater is not an aging effect requiring management. The PWR Containment Structures License Renewal Industry Report also states that exposure to acid rain and intermittent Page 3-230
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 3.6.0-1 Containment Structure - WCAP-14756-A, Final Safety Evaluation Report Response to Applicant Action Items Renewal Applicant Action Item Plant-Specific Response (12) (continued) exposure to aggressive chemicals will not cause significant degradation to PWR containments.
These facts notwithstanding, the current mandated inspections performed in accordance with the requirements of 10 CFR 50.55a and ASME Code Section XI, Subsection IWL, Examination Category L-A, and the Structures Monitoring Program manage the concrete aging effects associated with aggressive chemical attack.
Therefore, the potential aging effects associated with aggressive chemical attack will be adequately managed during the period of extended operation in the unlikely event that these mechanisms should become active.
(13) Provide details of the groundwater A groundwater monitoring process is being monitoring program and discuss potential implemented at Ginna Station and will seasonal variation in ground water chemistry. address potential seasonal variation in groundwater chemistry. There is no permanent dewatering system at Ginna Station.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 3.6.0-1 Containment Structure - WCAP-14756-A, Final Safety Evaluation Report Response to Applicant Action Items Renewal Applicant Action Item Plant-Specific Response (14) For prestressed concrete containments, Minor leakage of grease from tendon conduits discuss plant experience with respect to tendon was observed from the tendon fill-port piping grease leakage and, if applicable, how the in 2000. The tendon fill ports are located leakage will be managed; also discuss the around the exterior of the Containment wall at potential effects of grease leakage on the shear the base mat level. The ports are nominal 3" load capacity of the containment structure. diameter galvanized carbon steel pipes which are embedded in concrete. Each pipe is connected to the bottom of a tendon conduit at one end, and the other (fill) end projects vertically above the base mat for a short distance (about 12 inches). Each fill-port pipe is capped with a bronze valve. During original construction, liquefied grease was pumped into the bottom of each tendon conduit through the fill port and into a carbon steel reservoir (grease can) located at the top of each tendon. After a tendon was filled, the valve was shut and plugged with a threaded bronze plug. Ingress of ground water has occurred periodically over the years around the base of the Containment wall, at times submerging the fill ports and causing general corrosion of the exposed fill-port piping. The presence of the bronze valve created a galvanic couple which acted to accelerate the corrosion of the carbon steel pipe. Wall thinning progressed through-wall in some fill-ports, causing grease to leak out of the pipe stubs. After engineering assessment, repair of the fill-port pipe stubs was accomplished by encapsulating the short exposed length of pipe with a high-strength epoxy plastic sleeve, secured to the base mat surface with anchor bolts. The ingress of water has been brought under control. A visual inspection of the tendon top anchorage assemblies was undertaken in 2000-2001 to evaluate the grease levels in the upper reservoirs (grease cans) and to assess the condition of the wires and other anchorage hardware. Grease levels varied from tendon to tendon, but generally the grease levels were lower than desired and, in some cases, the tendon wires were exposed. Water Page 3-232
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 3.6.0-1 Containment Structure - WCAP-14756-A, Final Safety Evaluation Report Response to Applicant Action Items Renewal Applicant Action Item Plant-Specific Response (14) (continued) intrusion was observed in one grease can.
For those tendons with exposed wires, it was determined after careful inspection that the wires were covered with a film of grease and no evidence of corrosion was found. Gasket seals at the top and bottom of each can were replaced, and grease was added to these cans in the summer of 2001.
Management of grease leakage is accomplished by the Periodic Surveillance and Preventive Maintenance Program, the Structures Monitoring Program and the ASME Section XI, Subsections IWE & IWL Inservice Inspection Program. These programs include accepted elements that manage grease leakage. Further, since the aging management program follows NRC accepted inspection and maintenance procedures for prestressing systems, that also include the management of tendon grease leakage, there will be no detrimental effect on the shear load capacity of the containment structure due to grease leakage effects.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 3.6.0-1 Containment Structure - WCAP-14756-A, Final Safety Evaluation Report Response to Applicant Action Items Renewal Applicant Action Item Plant-Specific Response (15) Each license renewal applicant needs to Dissimilar metal welds and stainless steel describe its plant-specific program to address bellows assemblies at Ginna Station are the stress corrosion cracking (SCC) for exposed to Containment air (external dissimilar metal welds, and for stainless steel surfaces) and nitrogen or dry instrument air bellows assemblies, if the material is not (internal surfaces). These environments are shielded from a corrosive environment. For the not corrosive. Stainless steel components period of extended operation, ASME Section Xl, exposed to these environments are not IWE examination Categories E-B and E-F and susceptible to stress corrosion cracking; augmented VT-1 visual examination of bellows therefore cracking due to SCC is not an aging assemblies and dissimilar metal welds are effect requiring management for dissimilar required or a suitable alternative proposed. metal welds and stainless steel bellows assemblies. Plant-specific operating experience at Ginna Station confirms this conclusion.
Nevertheless, the ASME Section XI, Subsections IWE & IWL Inservice Inspection Program which includes augmented VT-1 examinations in certain cases is implemented at Ginna Station and would adequately manage cracking due to SCC in the unlikely event that such an aging effect were to develop.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 3.6.0-1 Containment Structure - WCAP-14756-A, Final Safety Evaluation Report Response to Applicant Action Items Renewal Applicant Action Item Plant-Specific Response i
(16) Discuss the plant-specific coatings While it is recognized that coatings provide monitoring and maintenance program and protection to metal surfaces, coatings specify whether it is credited as an AMP for themselves are not credited at Ginna Station containment steel elements. for management of aging effects since they perform no license renewal intended function.
Degradation of coatings typically is a result of aggressive chemical or thermal environments. As coatings degrade, the metal surfaces to which they are applied also degrade. Aging management is accomplished by assessing the integrity of the metal surfaces. If degradation of a coated steel component is identified, corrective action includes repair or reapplication of the coating.
Aging effects associated with aggressive environments are managed by the ASME Section XI, Subsections IWE & IWL Inservice Inspection Program, the Structures Monitoring Program and the Boric Acid Corrosion Program. These programs have been demonstrated to effectively manage age-related degradation of Containment steel surfaces at Ginna Station.
(17) For prestressed concrete containments, Degradation of the post-tensioning system at specify whether post-tensioning system Ginna Station will be managed in accordance degradation will be managed by AMP 5.6 with the requirements of ASME Section Xl, (Section XI, Subsection IWL, Requirements for Subsection IWL, Category L-B, 1992 Edition Class CC Concrete Components of Light-Water with 1992 Addenda along with the additional Cooled Power Plants, Examination Category requirements delineated in the amendment to L-B, Unbonded Post-Tensioning System, 1992 10 CFR § 50.55a (see SECY-96-080).
Code Edition with 1992 Addenda of the ASME Code) and the additional requirements delineated in 10 CFR 50.55a(b)2(ix). If not, provide the technical basis for exclusion.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 3.6.0-1 Containment Structure - WCAP-14756-A, Final Safety Evaluation Report Response to Applicant Action Items Renewal Applicant Action Item Plant-Specific Response (18) Specify whether settlement of the Settlement of structure is directly related to containment foundation is an applicable aging the physical properties of the foundation mechanism which will be managed by AMP 5.7. material. For structures located on rock strata If not, provide the technical basis for exclusion. or a suitable foundation, any settlement should have occurred during or immediately following construction. For structures not located on rock strata or a suitable foundation, inspections may be performed to verify that excessive settlement has not occurred. The majority of the settlement should have occurred prior to the end of the current licensing period.
There is no permanent dewatering system at Ginna Station. The Cable Tunnel is founded on steel foundation piles driven to bedrock or selected backfill. The Standby Auxiliary Feedwater Building is supported by 12" caissons that are socketed into competent rock. All other structures are founded on competent bedrock, which is fine-grained sandstone with nearly horizontal bedding planes and joints of limited vertical extent.
The Containment cylinder is founded on bedrock which acts as an integral part of the containment structure by the use of post-tensioned rock anchors.
Structural inspections indicate no visible evidence of settlement since construction of the station. During the Systematic Evaluation Program, the NRC concluded that settlement of foundations and buried equipment is not a safety concern for Ginna Station.
Therefore, settlement is not an applicable aging mechanism for Containment structure concrete components. This notwithstanding, the ASME Section XI, Subsections IWE &
IWL Inservice Inspection Program and the Structures Monitoring Program effectively manage aging effects due to settlement of the Containment structure during the period of extended operation.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 3.6.0-1 Containment Structure - WCAP-14756-A, Final Safety Evaluation Report Response to Applicant Action Items I
Renewal Applicant Action Item Plant-Specific Response (19) Identify whether erosion of the porous The NRC has issued Information Notices (IN) concrete sub-foundation layer is an applicable 97-11 and 98-26 informing nuclear facilities of aging mechanism; if applicable, provide an the possibility of degradation to structure aging management review and describe the foundations due to the erosion of porous credited aging management program. concrete sub-foundations. According to NUREG-1801, erosion of porous concrete sub-foundations must be analyzed. According to the Ginna Station review of NRC INs 97-11 and 98-26, the structure foundations at Ginna Station are constructed of normal concrete and not the subject porous type. Ginna Station is not one of the nine nuclear plants identified by the NRC in October 1996 as having the potential of degradation due to porous concrete. Since porous concrete is not used at Ginna Station, erosion of porous concrete sub-foundations is not applicable aging mechanism.
(20) The GTR listed only six (6) attributes to Aging management programs credited for form the basis for each aging management managing effects of aging for concrete and program. However, the "Draft Standard Review steel Containment components and the Plan for the Review of License Renewal post-tensioning system contain the ten (10) applications for Nuclear Power Plans," dated attributes identified in the SRR These April 21, 2000, identifies ten (10) elements programs are described in Appendix B of the (attributes) as appropriate for an acceptable LRA.
AMP. The GTR predates the Draft standard Review Plan for the review of Licensing Renewal applications for Nuclear Power Plans, and states in Section 4.0 that the report only presents program attributes for the AMPs, and that plant-specific details of the AMPs will be developed during the preparation of license renewal applications. Therefore, applicants for license renewal will be responsible for developing and describing the plant-specific AMPs and addressing each of the ten elements specified in the Draft Standard Review Plan.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 3.6.0-1 Containment Structure - WCAP-14756-A, Final Safety Evaluation Report Response to Applicant Action Items Renewal Applicant Action Item Plant-Specific Response (21) The WOG GTR indicates that the license Fatigue is the progressive degradation of renewal applicant may update an existing design materials subjected to application of cyclic fatigue analysis to account for the additional loads that are less than the maximum years of plant operation or manage the effects of allowable static loads. Concrete components the aging mechanism through aging at Ginna Station were designed in management programs. The GTR uses AMP 5.5 accordance with ACI 318 and therefore have for managing the effects of fatigue during the excellent low-cycle fatigue properties. ACI renewal license period, and basically endorses standards limit the maximum design stress to the ASME Code Section XI surveillance and less than 50% of the static stress of the testing program. For components where CLB concrete. The concrete fatigue strength is fatigue TLAAs exist, this option would allow the about 55% of its static strength at extremely CLB fatigue Section III cumulative usage factors high cycles (>107 cycles) of loading.
(CUF) to be exceeded during the period of Therefore, fatigue is not an aging mechanism extended operation. The staff has not endorsed that can lead to cracking for Containment this option on a generic basis at this time. An structure concrete components at Ginna applicant wishing to pursue this option would Station.
have to obtain staff review and approval on a case-by-case basis. For components where CLB For steel components, fatigue is the fatigue TLAAs do not exist (are not addressed in cumulative effect of microstructural localized 10 CFR 54.21), aging effects due to fatigue can plastic deformation in the material section that be addressed by either a Section III fatigue occurs with each cycle of applied stress of analysis (including the additional years for the sufficient magnitude. Class I steel structures period of extended operation) or by adequately were designed in accordance with American managing these effects for the period of Institute of Steel Construction (AISC) Code extended operation. where consideration was given to the number of stress cycles, the expected range of stress, and type and location of structural members/detail. The maximum stress and the maximum range of stress are specified in the code. The stress permitted in structural steel Class I structure components provide an adequate safety margin against fatigue failure. Additionally, more margin is available since the actual cyclic loading is lower than that assumed in the analysis, which typically uses bounding conditions.
Typically, loads applied to structural members are constant or static; non-constant loading (i.e., due to wind, etc.) is infrequent. Fatigue has been evaluated with respect to its effects on the ability of Class I steel structures to perform their intended safety function during the period of extended operation. Since the Class I structures were designed in
_______________________________________________________________ +/-
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 3.6.0-1 Containment Structure - WCAP-14756-A, Final Safety Evaluation Report Response to Applicant Action Items Renewal Applicant Action Item Plant-Specific Response (21) (continued) accordance with the AISC Code, the stress ranges in steel components and connections will be limited and therefore cracking due to fatigue is not an aging effect requiring management.
Cracking due to metal fatigue is treated as a Time-Limited Aging Analysis (TLAA) at Ginna Station ifincluded in the current licensing basis (CLB).
(22) Specify the containment structure There are no TLAA's for prediction of components and provide plant-specific details of cumulative fatigue usage of Containment the TLAAs for prediction of cumulative fatigue structure components at Ginna Station.
usage through the period of extended operation.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 3.6.0-1 Containment Structure - WCAP-14756-A, Final Safety Evaluation Report Response to Applicant Action Items Renewal Applicant Action Item Plant-Specific Response (23) Specify those containment structure Containment Liner and Penetrations components for which fatigue is an applicable aging mechanism, but no CLB fatigue analysis The Containment liner, liner penetrations, and based on a 40-year plant life exists. In addition liner steel components were originally to implementation of AMP 5.5, the requirements designed to comply with ASME Code Section of 10 CFR 50.55a should be met. Il1-1965 for pressure boundary and the AISC Code for structural steel. The Containment liner and penetrations, including the personnel and equipment hatch penetrations, were designed as Class B vessels. The winter 1965 Addenda of ASME Section III, Subsection B, N-1314(a) requires that the Containment vessel satisfy the provisions of Subsection A, N-415.1, "Vessels Not Requiring Analysis for Cyclic Operation," in order that the Subsection B rules may be applied. ASME Section III, N-415.1 states that a fatigue analysis is not required provided the service loading of the vessel or component meets six (6) conditions. An analysis has been performed which demonstrates that all six conditions are met for the liner and penetrations and that the ASME Section III Code rules for fatigue are met for the extended period of operation.
Containment Liner Channel Anchors A fatigue analysis of the fillet weld attaching the channel anchors to the liner was performed as part of the original design. The liner anchorage was originally designed for 100,000 stress cycles. This corresponds to more than four full stress cycles on a daily basis for 60 years. Fluctuations of temperature and pressure in the Containment on a daily basis are not significant enough in magnitude to cause four cycles of design basis stress at the liner anchorage weld every day. Therefore the original fatigue analysis remains valid for the period of extended operation.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 3.6.0-1 Containment Structure - WCAP-14756-A, Final Safety Evaluation Report Response to Applicant Action Items Renewal Applicant Action Item Plant-Specific Response (23) (continued) Containment Stainless Steel Tendon Bellows The fatigue usage factor for the Containment stainless steel tendon bellows was calculated for the period of extended operation using the allowable radial and vertical displacements.
The CUF was calculated to be .004 at the end of 60 years. Therefore the structural integrity of the bellows will be maintained throughout the period of extended operation.
Tendon Wire Fatigue Analysis Fatigue tests were conducted by the Swiss Federal Testing Station (EMPA) in 1960 on individual 7-mm tendon wires and on 18-wire tendons. The tendons and wires were cycled between 0.7 UTS and 0.8 UTS (ultimate tensile strength) for over 2 million cycles without the failure of a single wire. An analysis has been performed which demonstrates that the original seismic fatigue evaluation for the tendon wires remains valid throughout the period of extended operation.
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Renewal Applicant Action Item Plant-Specific Response i
(24) For prestressed concrete containments, In accordance with ACI 318-63 and the Ginna provide plant-specific details of the TLAA for Station UFSAR the design of the Containment prediction of tendon prestress losses through Structure post-tensioning system accounts for the period of extended operation. prestress losses caused by the following mechanisms/processes:.
"*Elastic shortening of concrete
"*Creep of concrete
"-Shrinkage of concrete.
"*Stress-relaxation in steel tendon wires.
"*Frictional loss due to curvature in the tendons and contact with tendon conduits.
No allowance is made for seating of the BBRV anchor since no slippage occurs in the anchor during transfer of the tendon load to the structure.
Loss of prestress of the Containment structure post-tensioning system is a Time-Limited Aging Analysis and is discussed in Section 4.5 of the LRA.
i (25) The GTR identified structural connections There are no Containment structural as containment structure components that connections unique to Ginna Station. The require aging management in Table 2-1. parts or subcomponents that could be However, there is no definition or description of considered structural connections in GTR structural connections in GTR Section 2.0. A Table 2-1 are addressed under penetrations, definition and a description of the AMP for equipment and personnel hatches. Therefore, structural connections are needed. a description of the AMP for structural connections need not be provided.
(26) The GTR identified embedments as The embedments are part of subcomponents containment structure components that require given in GTR Table 2-1 (e.g., equipment and aging management in Table 2-1. However, there personnel hatches). Additional embedments is no definition or description of embedments in are associated with the liner and other GTR Section 2. A definition and a description of miscellaneous equipment attached to the the AMP for embedments are needed. containment structure. This equipment is identified elsewhere in the LRA, along with the aging management program. Therefore, a description of the AMP for embedments need not be provided here.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 3.6.0-1 Containment Structure - WCAP-14756-A, Final Safety Evaluation Report Response to Applicant Action Items Renewal Applicant Action Item Plant-Specific Response (27) The GTR does not commit to inspection of Structural components inaccessible for inaccessible areas when there is no indication of inspection were evaluated for potential aging degradation of adjacent accessible areas, effects based on their environment as part of except when the potential for degradation is the aging management review. Several "event driven"; i.e., some unusual event has structural components that are inaccessible occurred which has the potential to degrade for visual inspection require aging inaccessible areas of the containment management at Ginna Station. Examples structures. Therefore, the GTR cannot be include buried concrete, embedded steel, and referenced by license renewal applicants for structural components blocked by installed managing aging of inaccessible areas. Individual equipment or structures. Structural license renewal applicants are required to components inaccessible for inspection are describe a program for inspection of managed by inspecting accessible structures inaccessible areas or adopt a program endorsed with similar materials and environments for by the staff in similar applications. aging effects that may be indicative of age-related degradation of inaccessible structural components. The programs credited for managing aging effects of inaccessible structural components are the Structures Monitoring Program and the ASME Section X1, Subsections IWE & IWL Inservice Inspection Program. These programs are described in Appendix B.
These aging management programs have been implemented at Ginna Station to meet the USNRC requirements as given in SECY-96-080 to improve the quality and effectiveness of containment inspections, and to inspect, and as needed to take corrective action for defects, in critical areas that include inaccessible areas. These programs will be continued into the extended period of operation as defined by the licensee renewal program. Therefore, the programs used at Ginna Station to inspect inaccessible areas follows the program endorsed by the staff in the amendment to 10 CFR § 50.55a (see via SECY-96-080).
It is also noted that these aging management programs address inaccessible areas as follows:§
- via the requirements of 10 CFR § 50.55a amendment (SECY-96-080);
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 3.6.0-1 Containment Structure - WCAP-14756-A, Final Safety Evaluation Report Response to Applicant Action Items Renewal Applicant Action Item Plant-Specific Response (27) (continued)
- when "serious degradation in inaccessible areas" is identified from other utility plant inspections that may be an area of concern for this plant;
"*when indicators exist that degradation may be occurring in an inaccessible area;
"*when an event occurs that could affect an inaccessible area making it susceptible to degradation.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 3.6.0-1 Containment Structure - WCAP-14756-A, Final Safety Evaluation Report Response to Applicant Action Items Renewal Applicant Action Item Plant-Specific Response i
(28) The aging effects in concrete due to Leaching of calcium hydroxide is typically leaching of calcium hydroxide and alkali observed in concrete that is exposed to aggregate reaction are identified in the GTR as flowing water. Containment concrete not requiring aging management. This is structures and components are not exposed unacceptable because plant-specific evaluation to flowing water at Ginna Station. In addition, of their applicability is needed. Therefore, if concrete structures and components at Ginna these aging mechanisms (leaching of calcium Station are constructed of dense, well-cured hydroxide and alkali aggregate reaction) are concrete with an amount of cement suitable applicable, applicants would be required to for strength development, and a propose a plant specific aging management water-to-cement ratio that is characteristic of program. Alternatively, applicant can credit the low-permeability concrete. This is consistent ASME Code, Section Xl, Examination Category with the guidance provided by the ACI L-A as an adequate aging management 201.2R. Therefore, degradation caused by program. leaching of calcium hydroxide is not significant and leaching is not an applicable aging mechanism for Containment structure concrete components at Ginna Station.
Concrete components at Ginna Station were constructed using non-reactive aggregates that were tested for potential reactivity in accordance with ASTM C227 and ASTM C295, which are established industry standards. Therefore, reaction with aggregates is not an aging mechanism that can lead to degradation of Containment structure concrete components.
These facts notwithstanding, the ASME Section Xl, Subsections IWE & IWL Inservice Inspection Program and the Structures Monitoring Program are aging management programs implemented at Ginna Station which manage aging effects due to leaching of calcium hydroxide and reaction with aggregates. These programs provide assurance that all potential concrete aging effects will be adequately managed throughout the period of extended operation.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 3.6.0-2 Reactor Coolant System Supports - WCAP-14422, Rev. 2-A, Final Safety Evaluation Report Response to Applicant Action Items Renewal Applicant Action Item Plant-Specific Response 4:
(1) Definition of "local" and "adjacent" Concrete adjacent to the support is addressed (Section 3.1) in the Containment Structure aging management review. Concrete local to the The Westinghouse Owners Group did not embedment is addressed with the RCS clearly define the term "local" in its report. supports.
However, the aging management programs could be the same for all concrete structures and structural components, therefore, the license renewal applicants must describe the aging management program for adjacent concrete structures and any differences from the aging management program for the local concrete structures.
(2) Detailed description of the Reactor Coolant The configurations of the RCS component System supports (Section 3.1) supports at Ginna Station are identical to those described in the GTR, namely:
A license renewal applicant will have to justify any differences between its Reactor Coolant
- Reactor Vessel Supports - Configuration 1 System support system and the figures and
- Steam Generator Supports - Configuration 3 descriptions of the supports systems contained - Pressurizer Supports - Configuration 2 in the Westinghouse Owners Group report.
- Reactor Coolant Pump Supports Configuration 2
- RCS Pressurizer Surge Line Support Variable spring hanger (3) Discrepancies and Omissions (Section 3.1)
The Westinghouse Owners Group report contains many discrepancies and omissions. A license renewal applicant needs to resolve these discrepancies and omissions in its application.
- 1. Wear plates and bearing pads are included 1. Table 2.4.2-12 includes wear plates and as support components and are within the bearing pads scope of this Westinghouse Owners Group report but are not identified in Table 2-1 as parts and sub-components requiring an aging management review.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 3.6.0-2 Reactor Coolant System Supports - WCAP-14422, Rev. 2-A, Final Safety Evaluation Report Response to Applicant Action Items Renewal Applicant Action Item Plant-Specific Response
- 2. Sketches of Reactor Coolant Pump support 2. The Ginna RCP supports are represented configuration 4 and Pressurizer support by Configuration 2, not Configuration 4.
configuration 2 are not provided in the Figure 2-11 in the GTR depicts Configuration Westinghouse Owners Group report. 2 by eliminating the upper supports with the PZR bolted into the concrete floor.
- 3. Section 3.2.9 of the Westinghouse Owners 3. ASTM A36 structural carbon steel material Group report indicates that ASTM A36 steel is included in the scope of the aging is used in Steam Generator and Reactor management review as it is utilized in SG Coolant Pump supports, however, ASTM supports. RCP supports do not utilize A36 A36 steel is not included in the list of steel.
material for the primary component supports (Table 2-4).
- 4. The 1963 AISC manual (Ref. 3) states that 4. Of the materials listed, only ASTM A36 is the following steel materials are commonly used in Ginna RCS supports. This material used for steel construction but they are not has been included in the material matrix listed in Table 2-4 of the Westinghouse table in the aging management review Owners Group report. They are ASTM A7, report.
A36, A242, A373, A440, and A441 structural steel and ASTM A325 bolts.
- 5. There are no specific descriptions and 5. A description and sketch of the Ginna PZR sketches for the pressurizer surge line surge line supports has been included in the supports. aging management review report. This support (RCU-1) is a variable spring hanger and the only support in the line.
(4) Strain Aging Embrittlement (Section 3.3.1.4) RCS supports at Ginna Station:
Temper embrittlement and strain aging a. do not contain cast austenitic stainless or embrittlement are the most common forms of low carbon steels; thermal embrittlement that are seen in ferritic materials as stated in Section 3.2.4 of the b. are not loaded beyond the elastic limit Westinghouse Owners Group report. The during normal operation, and Westinghouse Owners Group report has determined that temper embrittlement is not a c. operate at temperatures below 450 0F.
concem for the ferritic materials of Reactor Coolant System supports. However, the Consequently, it can be concluded that thermal Westinghouse Owners Group report does not and strain-aging embrittlement are not address the aging effects from strain aging applicable aging mechanisms at Ginna Station embrittlement but states that thermal and therefore no aging management of embrittlement is not applicable. The license age-related degradation caused by these renewal applicants will address the applicability mechanisms is required.
of the aging effects due to strain energy embrittlement to their plants.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 3.6.0-2 Reactor Coolant System Supports - WCAP-1 4422, Rev. 2-A, Final Safety Evaluation Report Response to Applicant Action Items Renewal Applicant Action Item Plant-Specific Response (5) Low Fracture Toughness (Section 3.3.1.6) Plant-specific evaluation of low fracture toughness and lamellar tearing as applicable to Appendix C of NUREG-0577 addresses this Ginna RCS supports was performed and item and groups many Westinghouse Owners submitted to the NRC in 1978. The primary Group member plants as Group I "plants conclusion of this analysis was that low requiring further evaluation." Although Table B9 fracture toughness and lamellar tearing are not of NUREG-1557 indicated that "low fracture a concern for the design and installation of the toughness is not significant for containment RCS supports at Ginna Station.Furthermore, internal structures," in general, these two Ginna Station is not listed as a Group I plant in documents only addressed the containment Appendix C of NUREG-0577.Therefore, no internal structures as a whole and did not aging management program is required for the specifically address the Reactor Coolant effects of low fracture toughness and lamellar System support components. Westinghouse tearing during the extended period of Owners Group recognizes this concern and operation.
states in Section 3.2.9 of its report that "Utilities with potential problems were required to demonstrate that the suspect structures have adequate fracture toughness to comply with the criteria defined in NUREG-0577." However, it further states that "low fracture toughness does not cause detrimental aging effects that must be addressed by maintenance programs." The staff does not believe that the Westinghouse Owners Group report provides sufficient information to support this conclusion. A license renewal applicant will address, if its plant is listed as Group 1 in Appendix C of NUREG-0577, that its plant had performed an analysis and the steel components of its RCS supports have adequate fracture toughness that no maintenance program is necessary.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 3.6.0-2 Reactor Coolant System Supports - WCAP-14422, Rev. 2-A, Final Safety Evaluation Report Response to Applicant Action Items Renewal Applicant Action Item Plant-Specific Response (6) Fatigue (Section 3.3.1.7) The only difference between materials used for Ginna RCS supports and those shown in Table A license renewal applicant will have to justify 2-4 of the GTR is ASTM A36 carbon steel any differences between the materials used for which is used in the steam generator supports.
its Reactor Coolant System supports and the However, A36 is used in a design configuration values listed in Table 2-4 of the Westinghouse where the loading is not cyclic and therefore Owners Group report. fatigue is not a concern.The rigid bumpers in the upper support system of the steam generators also utilized materials that are not listed in Table 2-4 of the GTR. These are listed below:
"*Clevis Pin - Al 93
"*Tangential Plate - A537
"*Guide Shaft - A193
- Tie Rod - A193
"*Rod Head - A668
"*Boss - A537
"*Tangential End and Rod End Bearing Plate A36
"*Stop Nut - A193 These materials are also not subjected to cyclic or fluctuating loads and fatigue is therefore not a concern.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 3.6.0-2 Reactor Coolant System Supports - WCAP-1 4422, Rev. 2-A, Final Safety Evaluation Report Response to Applicant Action Items Renewal Applicant Action Item Plant-Specific Response (7) Irradiation of Concrete (Section 3.3.2.3) Fast neutron fluence and gamma ray dose at the cavity liner/concrete interface 6.0 feet The Westinghouse Owners Group report states above the core midplane (at the location of the that concrete degradation from irradiation will RPV supports) have been evaluated. The be addressed by plant-specific evaluation. The maximum neutron fluence at 54 EFPY is 9.49 x staff agrees with this suggestion and the 1017 n/cm 2 . The maximum gamma ray dose at license renewal applicant must develop the same location at 54 EFPY is 4.94 x 109 rad.
plant-specific program(s) to evaluate this These values are below the 1018 n/cm 2 fast concern. neutron fluence and 1010 rad gamma ray thresholds for concrete damage due to irradiation (cracking and change in properties).
Therefore, degradation of the concrete material associated with the RPV supports due to neutron irradiation damage is not a concern.
The Ginna RCS supports are loaded in compression during normal operation.
Furthermore, these supports (other than the RPV supports) are located sufficiently far from the reactor vessel core that effect of neutron embrittlement is not a concern. Therefore, no aging management program is required for neutron irradiation embrittlement.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 3.6.0-2 Reactor Coolant System Supports - WCAP-1 4422, Rev. 2-A, Final Safety Evaluation Report Response to Applicant Action Items Renewal Applicant Action Item Plant-Specific Response i
(8) Elevated Temperature of Concrete Cracking due to elevated temperature is not an (Section 3.2.4) aging effect requiring management at Ginna Station since concrete temperatures are The Westinghouse Owners Group report states maintained below American Concrete Institute that concrete operating temperature should not (ACI) code thresholds due to normal exceed 150°F and local area temperature containment and supplemental cooling.
should be kept under 200 0 F. The Cooling systems that limit long term concrete Westinghouse Owners Group report further temperatures to less than 150OF with local states that reactor pressure vessel supports areas limited to 200OF include the Reactor could be subjected to high temperatures that Compartment Fan Cooling System and the could potentially result in a local temperature RPV Support Pad Cooling System. The air above 200°F if supplemental cooling is not temperature exiting the reactor compartment is provided. For those support configurations monitored to assure compliance with these where the local temperature at concrete design requirements. Component cooling surfaces could exceed 200 0 F, special design water circulates continuously through the features are incorporated based on air or water cooling channels in the RPV support pads.
cooling to keep local temperature below 200'F. These cooling systems are within the scope of These temperatures are specified in the ASME license renewal and will continue to be Code. Therefore, elevated temperature is not a operated through the period of extended concern for concrete. operation, providing assurance that concrete operating temperature limits will be maintained Because the operating temperature of concrete at Ginna Station.
components are kept below the limits specified by the code by means of supplemented cooling, the staff considers that the aging effects of elevated temperature are applicable to the Reactor Coolant System supports and are being managed by supplemented cooling features.
The license renewal applicants will address the concern that the aging effects associated with elevated temperature are applicable and demonstrate that the existing design features in the plants are capable of preventing any unacceptable degradation during the period of extended operation.
Page 3-251
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 3.6.0-2 Reactor Coolant System Supports - WCAP-14422, Rev. 2-A, Final Safety Evaluation Report Response to Applicant Action Items Renewal Applicant Action Item Plant-Specific Response (9) SRP-LR (Section 3.4) The aging management program descriptions contained in Appendix B of the License The attributes of the aging management Renewal Application (LRA) address all programs provided in the Westinghouse required attributes. The program descriptions Owners Group report do not address all also contain any relevant Ginna-specific elements as listed in Table Al-1 of Appendix A operating experience.
of the SRP-LR. The applicants should address the missing review elements and describe the plant-specific experience, if any, related to aging degradation of the Reactor Coolant System supports in their applications.
(10) Details of leakage walk-ons and leakage Details of Ginna leakage identification and monitoring program (Section 3.4.2) monitoring programs are contained in plant operating procedures. The Ginna Station Boric A license renewal applicant must provide the Acid Corrosion Program, as implemented by necessary details to perform leakage plant procedures, contains details of the identification walkdowns and the details of the leakage identification walkdowns and leakage monitoring program(s), especially the monitoring requirements, including frequencies, for Aging Management Program frequencies, associated with identifying boric 1-1 and Aging Management Program 1-2. acid leakage. A description of the Boric Acid Corrosion Program is contained in Appendix B of the LRA. Plant procedures are available for on-site review.
(11) Baseline Inspection (Section 3.4.2) Although not characterized as "baseline inspections" at the time they were performed, All structures and structural components need inspections that serve as baseline inspections a baseline inspection to document the have been performed and documented for the condition of the structures and structural RCS Supports under the Ginna Station ASME components. Therefore, the renewal applicants Section X1, Subsection IWF Inservice must have plant-specific baseline inspection Inspection Program.
results for all structures and structural components, or a planned inspection to obtain such results and validate the aging management programs prior to entering the period of extended operation.
Page 3-252
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 3.6.0-2 Reactor Coolant System Supports - WCAP-1 4422, Rev. 2-A, Final Safety Evaluation Report Response to Applicant Action Items Renewal Applicant Action Item Plant-Specific Response (12) Inspection of inaccessible areas Structural components inaccessible for (Section 3.4.2) inspection were evaluated for potential aging effects based on their environment as part of For RCS supports located in inaccessible the aging management review. Several areas, a license renewal applicant must structural components that are inaccessible for provide an inspection program to inspect these visual inspection require aging management at RCS supports or provide technical justification Ginna Station. Examples include buried for not performing inspection. concrete, embedded steel, and structural components blocked by installed equipment or structures. Structural components inaccessible for inspection are managed by inspecting accessible structures with similar materials and environments for aging effects that may be indicative of age-related degradation of inaccessible structural components. The programs credited for managing aging effects of inaccessible structural components are the ASME Section XI, Subsections IWE & IWL Inservice Inspection Program and the Structures Monitoring Program . These programs are described in Appendix B.
Page 3-253
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 3.6.0-2 Reactor Coolant System Supports - WCAP-14422, Rev. 2-A, Final Safety Evaluation Report Response to Applicant Action Items Renewal Applicant Action Item Plant-Specific Response (13) Surveillance Frequency for AMP-1.2 Two aging management programs were (Section 3.4.3) identified for concrete embedments. The Boric Acid Corrosion Program is credited for AMP-1.2 specifies inspection frequency in managing degradation from boric acid leaks accordance with the requirements of and the Structures Monitoring Program for Subsection IWF-2410 (Inspection Program) managing general loss of material and change and Table IWB-2412-1, each 10-year interval in material properties. Both aging management following the first interval, 10-year inspection programs, described in Appendix B, meet or program, with IWB-2412. The staff considers exceed the frequencies recommended in ACI the frequency proposed by Westinghouse 349.3R.
Owners Group not to be adequate. The proposed frequency is in accordance with ASME standards, but the inspections are to the requirements of ACI Standards, therefore, the frequency of inspection should also follow the recommendations of the ACI standards.
Inspection frequencies recommended by ACl 349.3R-96 are every 10 years for below grade structures and controlled interiors and every 5 years for all other structures. Section 4.2.4.1 of NUREG/CR-6424 has the same recommendation for inspection frequencies. A license renewal applicant must address this concem in its application.
(14) Acceptance criteria for leakage walkdowns Acceptance criteria for leakage walkdowns and (Section 3.4.4) monitoring are included in the Boric Acid Corrosion Program, which is described in In accordance with the Westinghouse Owners Appendix B of the LRA.
Group report, leakage walkdowns and monitoring are plant-specific. Therefore, a license renewal applicant will have to provide the necessary qualitative or quantitative acceptance criteria for leakage walkdowns and monitoring.
Page 3-254
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 3.6.0-2Reactor Coolant System Supports - WCAP-1 4422, Rev. 2-A, Final Safety Evaluation Report Response to Applicant Action Items Renewal Applicant Action Item Plant-Specific Response (15) Acceptance Criteria for AMP-1.2 The inspection criteria for the aging (Section 3.4.4) management program to control effects of aggressive chemical attack and corrosion of AMP-1.2 specifies acceptance criteria in concrete embedments at Ginna Station are accordance with several ACl standards. These described in the Structures Monitoring ACl standards are ACI 201.2R-77, Program . These criteria are based on ACI and AC1224.1 R-89, and ACI 224R-89. The staff has ASME Section Xl guidance. ACl 201.2R-77, reviewed these ACI standards and concluded AC1224.1R-89, ACl 224R-89, and ACl that, except for ACI 224.1 R, they are mainly for 349.3R-96 are recommended as guides in the design and construction rather than aging GTR for establishing acceptance criteria. ACl effects management because those concrete 201.2R-7 and ACI 224.1 R-89 are mainly for properties are built-in by design and design and construction rather than aging construction. However, they do contain management. However, acceptance criteria attributes that can be used to develop contained in ACl 349.3R-96 that can be used inspection acceptance criteria for AMP-1.2. For as inspection acceptance criteria for the leakage walkdowns and leakage monitoring, concrete embedment area associated with the acceptance criteria are the same as that RCS supports are described below:
listed for AMP-1.1. The staff has also reviewed ACI 349.3R-96, which is referenced in the Acceptance criteria without further evaluation Westinghouse Owners Group report for surveillance technique, and concluded it has a. Absence of leaching and chemical attack acceptance criteria that can be modified and used as the inspection acceptance criteria for b. Popouts and voids less than 20 mm (3/4 in.)
AMP-1.2. These criteria include acceptance in diameter or equivalent surface area without further evaluation, acceptance after review, and conditions requiring further c. Scaling less than 5 mm (3/16 in.) in depth evaluation. The license renewal applicants will provide a description of the inspection d. Spalling less than 10 mm (3/8 in.) in depth acceptance criteria in their application for the and 100 mm (4 1/4 in.) in any dimension staff to review.
- e. Absence of any signs of corrosion in reinforcing steel system, anchorage components, exposed embedded metal surfaces and corrosion stains around the embedded metal
- f. Passive cracks (i.e., absence of recent growth or other degradation mechanisms at the crack) less than 0.4 mm (0.015 in.) in maximum width below any surface enhanced widening Page 3-255
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 3.6.0-2 Reactor Coolant System Supports - WCAP-1 4422, Rev. 2-A, Final Safety Evaluation Report Response to Applicant Action Items Renewal Applicant Action Item Plant-Specific Response (15) (continued) g. Absence of excessive deflections, settlements or other physical movements that may affect structural performance
- h. Absence of detached embedments or loose bolts
- i. Absence of indications of degradation due to vibratory loads from piping and equipment Acceptance criteria requiring review
- a. Appearance of leaching or chemical attack
- b. Popouts and voids less than 50 mm (2 in.) in diameter or equivalent surface area
- c. Scaling less than 30 mm (1 1/8 in.) in depth
- d. Spalling less than 20 mm (3/4 in.) in depth and 200 mm (8 in.) in any dimension
- e. Corrosion staining of undefined source of concrete surfaces
- f. Passive cracks less than 1 mm (0.04 in.) in maximum width
- g. Passive settlements or deflections within the original design limits Criteria defining further evaluation
- a. Observed concrete surface conditions which exceed the acceptance criteria limits given for the cases requiring review
- b. Conditions found to be detrimental to the structural or functional integrity as a result of review Page 3-256
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 3.6.O-2Reactor Coolant System Supports - WCAP-1 4422, Rev. 2-A, Final Safety Evaluation Report Response to Applicant Action Items Renewal Applicant Action Item Plant-Specific Response i
(16) Plant-Specific Programs. Recommendations Responses to specific items are arranged in the from Section 5 of the Westinghouse Owners same order as in the preceding column.
Group report (Section 3.6)
- Identification and evaluation of any " Fatigue is the only TLAA associated with the plant-specific Time-Limited Aging Analyses RCS supports. A discussion of TLAA's is applicable to their Reactor Coolant System presented in Section 4.0 of the LRA. No supports. explicit fatigue evaluation was performed during the design of RCS supports for Ginna Station. Review of the CLB and other analyses related to RCS supports did not identify any other TLAA.
Identification and evaluation of current-term "*Current-term programs implemented within programs implemented within the current the current licensing term at Ginna Station to licensing term to address technical issues address technical issues that were identified from industry practices and United States from industry practices and NRC directives Nuclear Regulatory Commission (NRC) will be continued into the license renewal directives [that] should be continued into the term, without modifications.
license renewal term. Modifications to or elimination of these programs have to be justified.
-i Identification and justification of plant-specific " Plant-specific programs for aging programs that deviate from the recommended management of Ginna RCS supports do not aging management programs. deviate from those recommended by the GTR such that the effectiveness of the program is reduced.
- Identification of any specific program " Loss of preload due to creep and necessary to ensure that proper preload is stress-relaxation is not an aging effect retained for the component supports within requiring management at Ginna Station.
the scope of this report. Nevertheless, inspections performed under the ASME Section XI, Subsection IWF Inservice Inspection Program and the Structures Monitoring Program (described in Appendix B of the LRA) provide assurance that proper pre-load is retained for RCS supports within the scope of this AMR report.
- Identification of any evidence of aging "*The Structures Monitoring Program covers degradation in inaccessible areas during the the identification of aging degradation in current licensing term, that is considered to inaccessible areas during the current potentially affect system intended functions. A licensing term. A plan of action to resolve or plan of action to address any identified address identified problems is also provided.
potential degradation should be provided.
Page 3-257
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Table 3.6.0-2 Reactor Coolant System Supports - WCAP-1 4422, Rev. 2-A, Final Safety Evaluation Report Response to Applicant Action Items Renewal Applicant Action Item Plant-Specific Response
"*Verification that the plant is bounded by this "*Based on review of the installed GTR. The actions applicants must take to configuration, operating environment, and verify that this plant is bounded will be materials of construction of the RCS provided in an implementation procedure. supports, Ginna Station is bounded by the Reactor Coolant System Supports GTR Plant-specific data are given in this report show that indeed this is true.
"*Plant-specific evaluation of potential "*It has been demonstrated that potential degradation due to irradiation of the degradation due to irradiation of the RCS components within the scope of this report. supports components within the scope of this AMR is very low. Consequently, no aging management program is needed.
Page 3-258
Table 3.6-1 Structures and Component Supports - Aging Management Programs Evaluated in NUREG-1801 that are Relied on for License Renewal Aging Further Aging Management Evaluation Component Effect/Mechanism Programs Recommended Discussion (1) Penetration sleeves, Cumulative fatigue TLAA evaluated in Yes, TLAA A fatigue analysis of penetration sleeves is contained in the penetration bellows, and damage (CLB accordance with 10 Ginna Station CLB and has been evaluated as a TLAA (see dissimilar metal welds fatigue analysis CFR 54.21(c) Section 4.6). However, penetration bellows and dissimilar exists) metal welds are not incorporated into Ginna Station's current licensing basis as TLAAs.
(2) Penetration sleeves, Cracking due to Containment ISI Yes, detection of The Containment Program implements and formally adopts the bellows, and dissimilar metal cyclic loading, or and Containment aging effects is to requirements of the ASME Section XI, Subsections IWE & IWL welds. crack initiation and leak rate test be evaluated Inservice Inspection Program as part of the Ginna Station growth due to SCC Inservice Inspection Program. Included in the scope of the IWE program are the exposed portions of the containment liner, the liner for the fuel transfer penetration, all other penetrations, associated bolting, moisture barriers, and all airlocks, seals, gaskets and penetration bellows previously included in the scope of Appendix J. The ASME Section XI, Subsections IWE
& IWL Inservice Inspection Program includes inspections and leak rate tests which would indicate the presence of significant degradation from to cracking due to cyclic loading or crack initiation and growth due to SCC. That notwithstanding, SCC is not an applicable aging mechanism for penetration sleeves, bellows and dissimilar metal welds. The carbon steel components within penetrations are not susceptible to SCC.
The stainless steel components require both a high temperature (>1400 F) and exposure to an aggressive chemical environment (e.g. exposure to chlorides). The bellows at Ginna Station are not exposed to aggressive chemical environments.
A review of plant specific operating experience did not identify any occurrences of bellows failures due to SCC. Furthermore a review of industry operating experience indicated that SCC of bellows was typically caused by poor design controls leading to the inadvertent introduction of contaminants.
Page 3-259 Application for Renewed Operating License
Table 3.6-1 Structures and Component Supports - Aging Management Programs Evaluated in NUREG-1801 that are Relied on for License Renewal Aging Further Aging Management Evaluation Component Effect/Mechanism Programs Recommended Discussion (3) Penetration sleeves, Loss of material Containment ISI No Consistent with NUREG-1801. The ASME Section XI, penetration bellows, and due to corrosion and Containment Subsections IWE & IWL Inservice Inspection Program ensures dissimilar metal welds leak rate test that the containment ISI and leak rate tests adopt and implement the requirements of ASME Section XI, Subsections IWE & IWL. Included in the scope of the IWE program are the exposed portions of the containment liner, the liner for the fuel transfer penetration, all other penetrations, associated bolting, moisture barriers, and all airlocks, seals, gaskets and penetration bellows previously included in the scope of Appendix J. Inspections and leak rate tests performed in accordance with the containment inspection procedures verify the containment pressure boundary integrity and may be credited for detecting loss of material due to corrosion.
(4) Personnel airlock and Loss of material Containment ISI No Consistent with NUREG-1 801. The ASME Section XI, equipment hatch due to corrosion and Containment Subsections IWE & IWL Inservice Inspection Program ensures leak rate test that the containment ISI and leak rate tests adopt and implement the requirements of ASME Section Xl, Subsections IWE & IWL. Included in the scope of the IWE program are the exposed portions of the containment liner, the liner for the fuel transfer penetration, all other penetrations, associated bolting, moisture barriers, and all airlocks, seals, gaskets and penetration bellows previously included in the scope of Appendix J. This program may be credited for managing the aging effects of loss of material due to corrosion.
Additionally, the Periodic Surveillance and Preventive Maintenance Program also requires visual inspections of hatches, hinges, locks, and closure mechanisms as well as elastomeric seals associated with the containment air locks and is also credited for managing the aging effect loss of material due to corrosion. The program may also be credited with managing the aging effects of loss of leak tightness in the closed position due to mechanical wear of locks, hinges and closure mechanisms.
Page 3-260 Application for Renewed Operating License
Table 3.6-1 Structures and Component Supports - Aging Management Programs Evaluated in NUREG-1801 that are Relied on for License Renewal Aging Further Aging Management Evaluation Component Effect/Mechanism Programs Recommended Discussion (5) Personnel airlock and Loss of leak Containment leak No Consistent with NUREG-1801. The ASME Section XI, equipment hatch tightness in closed rate test and Plant Subsections IWE & IWL Inservice Inspection Program ensures position due to Technical that the containment ISI and leak rate tests adopt and mechanical wear of Specifications implement the requirements of ASME Section Xl, Subsections locks, hinges and IWE & IWL. Included in the scope of the IWE program are the closure mechanism exposed portions of the containment liner, the liner for the fuel transfer penetration, all other penetrations, associated bolting, moisture barriers, and all airlocks, seals, gaskets and penetration bellows previously included in the scope of Appendix J. This program may be credited for managing the aging effects of loss of material due to corrosion.
Additionally, the Periodic Surveillance and Preventive Maintenance Program also requires visual inspections of hatches, hinges, locks, and closure mechanisms as well as elastomeric seals associated with the containment air locks and is also credited for managing the aging effects of loss of material due to corrosion. The program may also be credited with managing the aging effects of loss of leak tightness in the closed position due to mechanical wear of locks, hinges and closure mechanisms.
Page 3-261 Application for Renewed Operating License
Table 3.6-1 Structures and Component Supports - Aging Management Programs Evaluated in NUREG-1801 that are Relied on for License Renewal Aging Further Aging Management Evaluation Component Effect/Mechanism Programs Recommended Discussion (6) Seals, gaskets, and Loss of sealant and Containment ISI No Consistent with NUREG-1801. The ASME Section XI, moisture barriers leakage through and Containment Subsections IWE & IWL Inservice Inspection Program ensures containment due to leak rate test that the containment ISI and leak rate tests adopt and deterioration of joint implement the requirements of ASME Section Xl, Subsections seals, gaskets, and IWE & IWL. Included in the scope of the IWE program are the moisture barriers exposed portions of the containment liner, the liner for the fuel transfer penetration, all other penetrations, associated bolting, moisture barriers, and all airlocks, seals, gaskets and penetration bellows previously included in the scope of Appendix J. This program may be credited for managing the aging effects of loss of sealant and leakage through containment due to deterioration of joint seal, gaskets, and moisture barriers.
The Periodic Surveillance and Preventive Maintenance Program also requires visual inspections of hatches, hinges, locks, and closure mechanisms as well as elastomeric seals associated with the containment air locks and may also be credited for managing the aging effects of loss of sealant and leakage through containment due to deterioration of joint seal, gaskets, and moisture barriers.
Page 3-262 Application for Renewed Operating License
Table 3.6-1 Structures and Component Supports - Aging Management Programs Evaluated in NUREG-1801 that are Relied on for License Renewal Aging Further Aging Management Evaluation Component Effect/Mechanism Programs Recommended Discussion (7) Concrete elements: Aging of accessible Containment ISI Yes, if aging Consistent with NUREG-1801. The ASME Section XI, foundation, walls, dome. and inaccessible mechanism is Subsections IWE & IWL Inservice Inspection Program concrete areas due significant for identifies the evidence that an aging mechanism is present and to leaching of inaccessible areas active and also provides confirmation and verification of the calcium hydroxide, absence of all types of aging effects. Indication of aging effects aggressive may be absent if the materials of construction, design chemical attack, specifications, and operational environment preclude an aging and corrosion of mechanism but, due to the long lead time necessary for some embedded steel effects to manifest themselves, it is prudent to periodically assess the condition of SSCs regardless of the likelihood that a particular aging mechanism is applicable. The degradation of inaccessible concrete can create symptoms of aging effects that are detectable in accessible areas. Conversely, if aging effects are present in accessible areas it is sensible to extrapolate those effects into inaccessible areas and perform additional evaluations.
Containment accessible and inaccessible concrete has been evaluated for the following aging mechanisms:
Aging Mechanism: Aggressive Chemical Attack Aging Effect: Loss of Material, Changes in Material Properties Evaluation: Concrete degradation in air due to aggressive rainwater is insignificant and the below-grade/lake water environment is non-aggressive. Additionally, recent structural inspections revealed no evidence of degradation owing to aggressive chemical attack; therefore, loss of material and change in material properties due to aggressive chemical attack are not probable aging effects at Ginna Station and have not been observed to date. The Structures Monitoring Program requires periodic monitoring of ground/lake water to verify chemistry remains non-aggressive.
Page 3-263 Application for Renewed Operating License
(
Table 3.6-1 Structures and Component Supports - Aging Management Programs Evaluated in NUREG-1801 that are Relied on for License Renewal Aging Further Aging Management Evaluation Component Effect/Mechanism Programs Recommended Discussion (7) (continued) Aging Mechanism: Corrosion of Embedded Steel Aging Effect: Loss of Material, Cracking, Loss of Bond Evaluation: Since the embedded steel is not exposed to an environment which is considered aggressive, loss of material, cracking, and loss of bond due to corrosion of embedded steel are not probable aging effects at Ginna Station and have not been observed to date.
Aging Mechanism: Leaching of Calcium Hydroxide Aging Effect: Change in Material Properties Evaluation: The original construction specifications met the intent of ACI 201.2R. Change in material properties due to leaching of calcium hydroxide is not a probable aging effect at Ginna Station and has not been observed to date.
Operating experience has shown that concrete has not experienced unanticipated aging effects at Ginna Station. That notwithstanding, the identification of the above aging effects by the ASME Section Xl, Subsections IWE & IWL Inservice Inspection Program, as well as the resistance provided by the materials of construction provide adequate assurance that all types of concrete aging effects will be identified and managed through out the extended period of operation.
Page 3-264 Application for Renewed Operating License
Table 3.6-1 Structures and Component Supports - Aging Management Programs Evaluated in NUREG-1801 that are Relied on for License Renewal Aging Further Aging Management Evaluation Component Effect/Mechanism Programs Recommended Discussion (8) Concrete elements: Cracks, distortion, Structures No, if within the Consistent with NUREG-1801. Cracks, distortion, and increase foundation and increases in Monitoring scope of the in component stresses due to settlement of concrete component stress applicant's foundations are considered in the Structures Monitoring level due to structures Program . All structures at Ginna Station are either founded on settlement monitoring program bedrock, steel foundation piles that are driven to bedrock, or have foundations that consist of caissons extending to bedrock. Structural inspections indicate no visible evidence of settlement since construction of the station. During the Systematic Evaluation Program, the NRC concluded that settlement of foundations and buried equipment is not a safety concem for Ginna Station. Cracking, distortion, and an increase in component stress levels due to settlement are not probable aging effects at Ginna Station and have not been observed to date. That notwithstanding, the Structures Monitoring Program monitors for cracks and distortion and contains inspection criteria to verify these aging effects are not developing.
(9) Concrete elements: Reduction in Structures No, if within the Consistent with NUREG-1801. Reduction in foundation foundation foundation strength Monitoring scope of the strength due to erosion of porous concrete subfoundations is due to erosion of applicant's not an aging effect requiring management at Ginna. Ginna porous concrete structures Station's structure foundations are constructed of normal subfoundation monitoring program concrete and not the subject porous type, nor are foundations subject to flowing water. That notwithstanding, the Structures Monitoring Program monitors for settlement and cracking. The identification of indications of settlement by the Structures Monitoring Program , as well as the resistance provided by the materials of construction, provide adequate assurance that reductions in foundation strength for any reason will be identified and managed through out the extended period of operation.
Page 3-265 Application for Renewed Operating License
Table 3.6-1 Structures and Component Supports - Aging Management Programs Evaluated in NUREG-1801 that are Relied on for License Renewal Aging Further Aging Management Evaluation Component Effect/Mechanism Programs Recommended Discussion (10) Concrete elements: Reduction of Plant specific Yes, for any Consistent with NUREG-1 801. For plant areas of concern, foundation, dome, and wall strength and portions of concrete temperatures are normally maintained below the specified modulus due to containment that limits; therefore, loss of material, cracking, and change in elevated exceed specified material properties due to elevated temperature at Ginna temperature temperature limits Station have not been observed to date. (Note: The SSCs relied upon to maintain the concrete surrounding containment penetrations and the reactor vessel support pad within specified temperature limits are within the scope of the License Renewal Rule, i.e. penetration cooling and component cooling water.) That notwithstanding, the ASME Section XI, Subsections IWE & IWL Inservice Inspection Program monitors for loss of material, cracks, and changes in material properties and contains inspection criteria to verify these aging effects are not developing.
(11) Prestressed containment: Loss of prestress TLAA evaluated in Yes, TLAA Consistent with NUREG-1 801. Loss of prestress is addressed tendons and anchorage due to relaxation, accordance with 10 as a TLAA in Section 4.3. Additionally, the ASME Section XI, components shrinkage, creep, CFR 54.21(c) Subsections IWE & IWL Inservice Inspection Program and elevated implements and formally adopts the requirements of the ASME temperature Section XI, Subsections IWE & IWL Inservice Inspection Program as part of the Ginna Station Inservice Inspection Program. Included in the scope of the IWL program are all exterior exposed accessible areas and exterior suspect areas of the concrete containment, and the post-tensioning system.
Tendon Surveillance Program procedure PT-27.2 performs periodic liftoff tests, grease analysis, and visual inspection of the top tendon anchorage hardware and thus provides reasonable assurance that loss of prestress due to stress relaxation, shrinkage, creep and elevated temperature and loss of material due to corrosion are effectively managed.
Page 3-266 Application for Renewed Operating License
Table 3.6-1 Structures and Component Supports - Aging Management Programs Evaluated in NUREG-1801 that are Relied on for License Renewal Aging Further Aging Management Evaluation Component EffectlMechanism Programs Recommended Discussion (12) Steel elements: liner Loss of material Containment ISI Yes, if corrosion is Consistent with NUREG-1801. The ASME Section XI, plate, containment shell due to corrosion in and Containment significant for Subsections IWE & IWL Inservice Inspection Program includes accessible and leak rate test inaccessible areas inspections and leak rate tests which would indicate the inaccessible areas presence of significant degradation due to loss of material from all applicable corrosion mechanisms. Additionally, plant operating experience has shown that borated water spills in containment have the potential to impact the containment liner.
Accordingly, the Boric Acid Corrosion Program is also credited with assessing and managing loss of material in the containment liner. Additional (non-pressure retaining) structural steel evaluations not accounted for by other items in this table are found in Table 3.6-2 Line Number (15) and Table 3.6-2 Line Number (17)
(13) Steel elements: protected Loss of material Protective coating No Protective coatings are not credited with managing the effects by coating due to corrosion in monitoring and of aging at Ginna Station. Ginna recognizes the benefits accessible areas maintenance derived from protective coatings. However coatings, in and of only themselves, do not perform License Renewal intended functions. That notwithstanding, steel elements in containment are inspected for corrosion by both the Structures Monitoring Program and the ASME Section XI, Subsections IWE & IWL Inservice Inspection ISI Program. When a steel coating is found degraded it is evaluated and repaired in accordance with station procedures.
Page 3-267 Application for Renewed Operating License
Table 3.6-1 Structures and Component Supports - Aging Management Programs Evaluated in NUREG-1801 that are Relied on for License Renewal Aging Further Aging Management Evaluation Component EffectlMechanism Programs Recommended Discussion (14) Prestressed containment: Loss of material Containment ISI No Consistent with NUREG-1801. The Ginna Station ISI program tendons and anchorage due to corrosion of has been revised to implement and formally adopt ASME components prestressing Section XI, Subsections IWE & IWL Inservice Inspection.
tendons and Included in the scope of the IWL program are all exterior anchorage exposed accessible areas and exterior suspect areas of the components concrete containment, and the post-tensioning system. Tendon Surveillance Program procedure PT-27.2 performs periodic liftoff tests, grease analysis, and visual inspection of the top tendon anchorage hardware and thus provides reasonable assurance that loss of prestress due to stress relaxation, shrinkage, creep and elevated temperature as well as loss of material due to corrosion, are effectively managed.
Ginna Station has a tendon anchor system that includes the use of rock anchors. Rock anchors basically consist of high strength steel wires and button heads grouted directly into the bedrock under containment. The Structures Monitoring Program requires periodic measurement of the galvanic potential between the anchor and the environment. Should the rock anchor/ tendon assemblies indicate electrical current flow of sufficient magnitude to support corrosion, mitigative measures can be taken to protect the rock anchors by application of a cathodic potential.
Additionally, due to operating experience (grease loss, water intrusion), the Periodic Surveillance and Preventive Maintenance Program requires visual inspection of the containment tendon grease cans for evidence of loss of seal (water intrusion) due to changes in gasket elastomeric properties, loss of grease due to leakage, and structural soundness of the exposed fill port pipe segments encapsulated in epoxy plastic. Thus, by ensuring the tendons are not exposed to moisture, the Periodic Surveillance and Preventive Maintenance Program also ensures loss of material due to corrosion of prestressing tendons and anchorage components is effectively managed.
Page 3-268 Application for Renewed Operating License
Table 3.6-1 Structures and Component Supports - Aging Management Programs Evaluated in NUREG-1801 that are Relied on for License Renewal Aging Further Aging Management Evaluation Component EffectlMechanism Programs Recommended Discussion (15) Concrete elements: Scaling, cracking, Containment ISI No Consistent with NUREG-1801. The ASME Section Xl, foundation, dome, and wall and spalling due to Subsections IWE & IWL Inservice Inspection Program freeze-thaw; identifies the evidence that an aging mechanism is present and expansion and active and also provides confirmation and verification of the cracking due to absence of all types of aging effects. Indication of aging effects reaction with may be absent if the materials of construction, design aggregate specifications, and operational environment preclude an aging mechanism but, due to the long lead time necessary for some effects to manifest themselves, it is prudent to periodically assess the condition of SSCs regardless of the likelihood that a particular aging mechanism is applicable. The degradation of inaccessible concrete can create symptoms of aging effects that are detectable in accessible areas. Conversely, if aging effects are present in accessible areas it is sensible to extrapolate those effects into inaccessible areas and perform additional evaluations.
Containment concrete elements have been evaluated for the following aging mechanisms:
Aging Mechanism: Freeze-Thaw Aging Effect: Loss of Material Evaluation: The contract-specified air contents are within the range specified by current revisions of ACI 318, and the contract-specified water-to-cement ratio meets the recommendations of ACI 318-63 (L.0.53). Therefore, loss of material and cracking of concrete due to freeze-thaw are not probable aging effects at Ginna Station and have not been observed to date.
Aging Mechanism: Reaction with Aggregates Aging Effect: Cracking, Expansion Evaluation: During construction the aggregates were tested for potential reactivity in accordance with ASTM C227 and ASTM C295, cracking and expansion due to reaction with aggregates are not probable aging effects at Ginna Station and have not been observed to date.
Page 3-269 Application for Renewed Operating License
f Table 3.6-1 Structures and Component Supports - Aging Management Programs Evaluated in NUREG-1801 that are Relied on for License Renewal Aging Further Aging Management Evaluation Component Effect/Mechanism Programs Recommended Discussion (15) (continued) Operating experience has shown that concrete has not experienced unanticipated aging effects at Ginna Station. That notwithstanding, the identification of the above aging effects by the ASME Section XI. Subsections IWE & IWL Inservice Inspection Program, as well as the resistance provided by the materials of construction provide adequate assurance that all types of concrete aging effects will be identified and managed through out the extended period of operation.
Page 3-270 Application for Renewed Operating License
Table 3.6-1 Structures and Component Supports - Aging Management Programs Evaluated in NUREG-1801 that are Relied on for License Renewal Aging Further Aging Management Evaluation Component Effect/Mechanism Programs Recommended Discussion (16) All Groups except Group All types of aging Structures No, if within the Consistent with NUREG-1801. The Structures Monitoring 6: accessible interior/exterior effects Monitoring scope of the Program identifies the evidence that an aging mechanism is concrete & steel components applicant's present and active and also provides confirmation and structures verification of the absence of all types of aging effects.
monitoring program Indication of aging effects may be absent if the materials of construction, design specifications, and operational environment preclude an aging mechanism but, due to the long lead time necessary for some effects to manifest themselves, it is prudent to periodically assess the condition of SSCs regardless of the likelihood that a particular aging mechanism is applicable. The degradation of inaccessible concrete can create symptoms of aging effects that are detectable in accessible areas. Conversely, if aging effects are present in accessible areas it is sensible to extrapolate those effects into inaccessible areas and perform additional evaluations.
Accessible interior and exterior concrete have been evaluated for the following aging mechanisms:
Aging Mechanism: Freeze-Thaw Aging Effect: Loss of Material Evaluation: The contract-specified air contents are within the range specified by current revisions of ACI 318, and the contract-specified water-to-cement ratio meets the recommendations of ACI 318-63 L< 0.53). Therefore, loss of material and cracking of concrete due to freeze-thaw are not probable aging effects at Ginna Station and have not been observed to date.
Aging Mechanism: Elevated Temperature Aging Effect: Loss of Material, Cracking, Changes in Material Properties Evaluation: For plant areas of concern, temperatures are normally maintained below the specified limits; therefore, loss of material, cracking, and change in material properties due to elevated temperature are not probable aging effects at Ginna Station and have not been observed to date. (Note: The SSCs Page 3-271 Application for Renewed Operating License
Table 3.6-1 Structures and Component Supports - Aging Management Programs Evaluated in NUREG-1801 that are Relied on for License Renewal Aging Further Aging Management Evaluation Component Effect/Mechanism Programs Recommended Discussion (16) (continued) relied upon to maintain the concrete surrounding containment penetrations and the reactor vessel support pad within specified temperature limits are within the scope of the License Renewal Rule, i.e. penetration cooling and component cooling water.)
Aging Mechanism: Aggressive Chemical Attack Aging Effect: Loss of Material, Changes in Material Properties Evaluation: Concrete degradation in air due to aggressive rainwater is insignificant and the below-grade/lake water environment is non-aggressive. Additionally, recent structural inspections revealed no evidence of degradation owing to aggressive chemical attack; therefore, loss of material and change in material properties due to aggressive chemical attack are not probable aging effects at Ginna Station and have not been observed to date. The Structures Monitoring Program requires periodic monitoring of ground/lake water to verify chemistry remains non-aggressive.
Aging Mechanism: Corrosion of Embedded Steel Aging Effect: Loss of Material, Cracking, Loss of Bond Evaluation: Since the embedded steel is not exposed to an environment which is considered aggressive, loss of material, cracking, and loss of bond due to corrosion of embedded steel are not probable aging effects at Ginna Station and have not been observed to date.
Aging Mechanism: Reaction with Aggregates Aging Effect: Cracking, Expansion Evaluation: During construction the aggregates were tested for potential reactivity in accordance with ASTM C227 and ASTM C295, cracking and expansion due to reaction with aggregates are not probable aging effects at Ginna Station and have not been observed to date.
________________________ i ________________ I._______________ J ________________ I _________________________________________________
Page 3-272 Application for Renewed Operating License
Table 3.6-1 Structures and Component Supports - Aging Management Programs Evaluated in NUREG-1801 that are Relied on for License Renewal Aging Further Aging Management Evaluation Component Effect/Mechanism Programs Recommended Discussion (16) (continued) Aging Mechanism: Settlement Aging Effect: Cracking, Distortion, Increase in Component Stress Level Evaluation: All structures at Ginna Station are either founded on bedrock, steel foundation piles that are driven to bedrock, or have foundations that consist of caissons extending to bedrock. Structural inspections indicate no visible evidence of settlement since construction of the station. During the Systematic Evaluation Program, the NRC concluded that settlement of foundations and buried equipment is not a safety concern for Ginna Station. Cracking, distortion, and an increase in component stress levels due to settlement are not probable aging effects at Ginna Station and have not been observed to date.
Aging Mechanism: Leaching of Calcium Hydroxide Aging Effect: Change in Material Properties Evaluation: The original construction specifications met the intent of ACI 201.2R. Change in material properties due to leaching of calcium hydroxide is not a probable aging effect at Ginna Station and has not been observed to date.
Operating experience has shown that concrete has not experienced unanticipated aging effects at Ginna Station. That notwithstanding, the identification of the above aging effects by the Structures Monitoring Program , as well as the resistance provided by the materials of construction provide adequate assurance that all types of concrete aging effects will be identified and managed through out the extended period of operation.
Accessible interior and exterior steel components have been evaluated for the following aging mechanisms:
Aging Mechanism: General Corrosion Aging Effect: Loss of Material Evaluation: Carbon and low-alloy steel surfaces, which are exposed to typical plant environments, can experience general Page 3-273 Application for Renewed Operating License
Table 3.6-1 Structures and Component Supports - Aging Management Programs Evaluated in NUREG-1801 that are Relied on for License Renewal Aging Further Aging Management Evaluation Component Effect/Mechanism Programs Recommended Discussion (16) (continued) corrosion. Additionally, structural steel can be subject to boric acid corrosion.
Corrosion of accessible interior and exterior structural steels is an aging effect that requires management at Ginna Station.
Operating experience has shown that corrosion can be initiated and/or accelerated by unique factors. Although Ginna Station is not located in an area subject to high ambient chloride or sulfate ion concentrations, localized factors such as the deposition of bird excrement on exterior structural steel have been shown to create adverse conditions that promulgate corrosion. Accordingly, the Structures Monitoring Program identifies and evaluates corrosion of interior and exterior structural steel. Additionally, accessible carbon low alloy structural steel located in areas that contain borated water systems are subject to the requirements of Boric Acid Corrosion Program.
Page 3-274 Application for Renewed Operating License
Table 3.6-1 Structures and Component Supports - Aging Management Programs Evaluated in NUREG-1801 that are Relied on for License Renewal Aging Further Aging Management Evaluation Component Effect/Mechanism Programs Recommended Discussion (17) Groups 1-3, 5, 7-9: Aging of Plant-specific Yes, ifan Consistent with NUREG-1 801. Inaccessible wall and concrete inaccessible concrete inaccessible aggressive foundations are considered in the Structures Monitoring components, such as exterior concrete areas due below-grade Program. Results of inspections for accessible concrete are walls below grade and to aggressive environment exists evaluated and, ifaging effects are noted, the Structures foundation chemical attack, Monitoring Program evaluates the symptom and possible and corrosion of causes with respect inaccessible areas. The Structures embedded steel Monitoring Program requires periodic monitoring of ground water to verify chemistry remains non-aggressive. Concrete degradation in air due to aggressive rainwater is insignificant and the below-grade/lake water environment is non-aggressive. Additionally, recent structural inspections revealed no evidence of degradation owing to aggressive chemical attack; therefore, degradation due to chemical attack is not a probable aging effect at Ginna Station.
The concrete at Ginna Station was designed in accordance with ACl 301-66 or ACl 318-63. ACl 301-66 refers to ACI 318 for concrete reinforcement. Designing concrete to ACI 318 also provides for sufficient concrete cover over embedded steel to provide ample corrosion protection. Chemical analyses performed on the rock and groundwater indicate these environments are non-aggressive. Since the embedded steel is not exposed to an environment which is considered aggressive, corrosion of embedded steel is not a probable aging effect at Ginna Station and has not been observed to date.
Page 3-275 Application for Renewed Operating License
Table 3.6-1 Structures and Component Supports - Aging Management Programs Evaluated in NUREG-1801 that are Relied on for License Renewal Aging Further Aging Management Evaluation Component Effect/Mechanism Programs Recommended Discussion (18) Group 6: all accessible/ All types of aging Inspection of No Ginna Station water control structures include the circulating inaccessible concrete, steel, effects, including Water-Control water system discharge canal, the canal's interface with the and earthen components loss of material due Structures or pump screen house, and a stone revetment which protects the to abrasion, FERC/US Army site from surge flooding. The water control structure cavitation, and Corps of Engineers inspections are performed in accordance with the Periodic corrosion dam inspections Surveillance and Preventive Maintenance Program and the and maintenance Structures Monitoring Program . Although the components and aging attributes monitored are consistent with NUREG-1801, the program assignment is not. Accordingly the components that comprise water control structures (Group 6 items) are detailed and evaluated in Table 3.6-2 Line Number (7) .
(19) Group 5: liners Crack initiation and Water Chemistry No Consistent with NUREG-1801. The Water Chemistry Control growth from SCC Program and Program is credited with managing the aging effects of crack and loss of material Monitoring of spent initiation and growth from SSC and loss of material due to due to crevice fuel pool water crevice corrosion for the spent fuel pool liner and refueling corrosion level transfer canal liner.
Plant Technical Specification, 3.7.11 Spent Fuel Pool (SFP)
Water Level, as well as plant operating procedures provide monitoring and control of the spent fuel pool water level.
(20) Groups 1-3, 5, 6: all Cracking due to Masonry Wall No Consistent with NUREG-1 801. Masonry wall inspections are masonry block walls restraint, shrinkage, incorporated into the Structures Monitoring Program . The creep, and Structures Monitoring Program effectively manages cracking aggressive due to restraint, shrinkage and creep. Concrete degradation in environment air due to aggressive rainwater is insignificant and the below-grade/lake water environment is non-aggressive.
Additionally, recent structural inspections revealed no evidence of degradation owing to aggressive chemical attack; therefore, degradation due to chemical attack is not a probable aging effect for concrete and masonry block walls. That notwithstanding, the Structures Monitoring Program monitors for indications of chemical attack on masonry walls.
Page 3-276 Application for Renewed Operating License
Table 3.6-1 Structures and Component Supports - Aging Management Programs Evaluated in NUREG-1801 that are Relied on for License Renewal Aging Further Aging Management Evaluation Component Effect/Mechanism Programs Recommended Discussion (21) Groups 1-3, 5, 7-9: Cracks, distortion, Structures No, if within the Consistent with NUREG-1 801. Cracks, distortion, and increase foundation and increases in Monitoring scope of the in component stresses due to settlement of concrete component stress applicant's foundations are considered in the Structures Monitoring level due to structures Program . All structures at Ginna Station are either founded on settlement monitoring program bedrock, steel foundation piles that are driven to bedrock, or have foundations that consist of caissons extending to bedrock. Structural inspections indicate no visible evidence of settlement since construction of the station. During the Systematic Evaluation Program, the NRC concluded that settlement of foundations and buried equipment is not a safety concern for Ginna Station. Cracking, distortion, and an increase in component stress levels due to settlement are not probable aging effects at Ginna Station and have not been observed to date. That notwithstanding, the Structures Monitoring Program monitors for cracks and distortion and contains inspection criteria to verify these aging effects are not developing.
(22) Groups 1-3, 5-9: Reduction in Structures No, if within the Consistent with NUREG-1 801. Reduction in foundation foundation foundation strength Monitoring scope of the strength due to erosion of porous concrete subfoundations is due to erosion of applicant's not an aging effect requiring management at Ginna. Ginna porous concrete structures Station's structure foundations are constructed of normal subfoundation monitoring program concrete and not the subject porous type, nor are foundations subject to flowing water. That notwithstanding, the Structures Monitoring Program monitors for settlement and cracking. The identification of indications of settlement indication by the Structures Monitoring Program , as well as the resistance provided by the materials of construction, provide adequate assurance that reductions in foundation strength for any reason will be identified and managed through out the extended period of operation.
Page 3-277 Application for Renewed Operating License
Table 3.6-1 Structures and Component Supports - Aging Management Programs Evaluated in NUREG-1801 that are Relied on for License Renewal Aging Further Aging Management Evaluation Component Effect/Mechanism Programs Recommended Discussion (23) Groups 1-5: concrete Reduction of Plant-specific Yes, for any Consistent with NUREG-1801. For plant areas of concern, strength and portions of concrete temperatures are normally maintained below the specified modulus due to that exceed limits; therefore, loss of material, cracking, and change in elevated specified material properties due to elevated temperature at Ginna temperature temperature limits Station have not been observed to date. (Note: The SSCs relied upon to maintain the concrete surrounding containment penetrations and the reactor vessel support pad within specified temperature limits are within the scope of the License Renewal Rule, i.e. penetration cooling and component cooling water.) That notwithstanding, the Structures Monitoring Program monitors for loss of material, cracks, and changes in material properties and contains inspection criteria to verify these aging effects are not developing.
(24) Groups 7, 8: liners Crack Initiation and Plant-specific Yes All tanks within the scope of License Renewal receive their growth due to SCC; aging management evaluation with the system they serve.
Loss of material Thus, this line item is not applicable to class 1 structures at due to crevice Ginna Station.
corrosion Page 3-278 Application for Renewed Operating License
Table 3.6-1 Structures and Component Supports - Aging Management Programs Evaluated in NUREG-1801 that are Relied on for License Renewal Aging Further Aging Management Evaluation Component Effect/Mechanism Programs Recommended Discussion (25) All Groups: support Aging of Structures No, if within the Consistent with NUREG-1801. (Note: Equipment included in members: anchor bolts, component Monitoring scope of the component support groups B13.1 and B13.2 (ASME class 1,2 and concrete surrounding anchor supports applicant's 3 supports), as well as the effects of boric acid corrosion on all bolts, welds, grout pad, bolted structures groups, are also discussed as separate items in the following connections, etc. monitoring program sections. Group B13.3 is applicable to BWRs.) The aging effects associated with component supports are considered in the Structures Monitoring Program . Additionally, component supports submerged in raw water are considered in the Periodic Surveillance and Preventive Maintenance Program.
Component supports include those structural elements that are connected to civil structures and which extend to a system or system components for the purpose of providing support or restraint. Inclusive in this boundary definition are any vibration dampeners or other passive connective appurtenances intrinsic to the functioning of the support. The group also includes spray or drip shields attached to equipment as well as electrical system rack, panels and enclosures. Component supports are located throughout the plant. Included in the evaluation of the component supports are supports for both safety-related components and non-safety related components whose failure could affect a safety function (typically referred to as seismic 11/1).
Component supports including support members; anchor bolts, concrete surrounding anchor bolts, welds, grout pad, bolted connections, etc. have been evaluated for the following aging mechanisms:
Aging Mechanism: Environmental Corrosion Aging Effect: Loss of material Evaluation: Carbon and low-alloy steels, which are exposed to typical plant environments, can experience general corrosion.
The Structures Monitoring Program and Periodic Surveillance and Preventive Maintenance Program identify and evaluate corrosion of component supports.
Page 3-279 Application for Renewed Operating License
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Table 3.6-1 Structures and Component Supports - Aging Management Programs Evaluated in NUREG-1801 that are Relied on for License Renewal Aging Further Aging Management Evaluation Component Effect/Mechanism Programs Recommended Discussion (25) (continued) Aging Mechanism: Service-induced cracking or other concrete aging mechanisms Aging Effect: Reduction in concrete anchor capacity due to local concrete degradation Evaluation: Operating experience has shown that service induced cracking can occur in grouted foundations. The Structures Monitoring Program and Periodic Surveillance and Preventive Maintenance Program identify and evaluate cracking and other concrete aging mechanisms for component supports.
Aging Mechanism: Degradation of vibration isolation elements Aging Effect: Reduction/loss of isolation function Evaluation: Operating experience has shown that elastomer materials can degrade over time. The Structures Monitoring Program identifies and evaluates the degradation of vibration isolation elements.
Aging Mechanism: Metal Fatigue Aging Effect: Cracking Evaluation: The metals and welds used in component supports and support fasteners are subjected to both service induced and, potentially, unanticipated or upset loads. Cracking due to fatigue in component supports is not a probable aging effect at Ginna Station and has not been observed to date. That notwithstanding, the Structures Monitoring Program and Periodic Surveillance and Preventive Maintenance Programs look for metal discontinuities and cracks that may be evidence of fatigue.
(26) Groups B1.1, B13.2, and Cumulative fatigue TLAA evaluated in Yes, TLAA A fatigue analysis for structures and component supports is not B1.3: support members: damage (CLB accordance with 10 incorporated into Ginna Station's current licensing basis.
anchor bolts, welds fatigue analysis CFR 54.21(c) Consequently, this line item is not applicable to Ginna Station.
exists)
Page 3-280 Application for Renewed Operating License
Table 3.6-1 Structures and Component Supports - Aging Management Programs Evaluated in NUREG-1801 that are Relied on for License Renewal Aging Further Aging Management Evaluation Component Effect/Mechanism Programs Recommended Discussion (27) All Groups: support Loss of material Boric acid corrosion No Consistent with NUREG-1801. The Boric Acid Corrosion members: anchor bolts, welds due to boric acid Program monitors for loss of material due to boric acid corrosion corrosion in all plant areas that contain systems that use boric acid. In addition to support members, the program also monitors and evaluates structural members, fasteners and welds that could be potentially exposed to borated water leaks.
(28) Groups B1.1, B1.2, and Loss of material ISI No Consistent with NUREG-1801. The ASME Section XI, B1.3: support members: due to Subsection IWF Inservice Inspection Program monitors all anchor bolts, welds, spring environmental elements of safety related supports for degradation and fouling.
hangers, guides, stops, and corrosion; loss of Visual examinations inspect for corrosion, deformation, vibration isolators mechanical misalignment, improper clearances, damage to sliding function due to surfaces, and missing detached, or loose support items.
corrosion, Additionally, some non-ASME supports are also included in the distortion, dirt, scope of the Inservice Inspection Program (e.g. selected high overload, etc. energy line pipe supports).
(29) Group B1.1: high Crack initiation and Bolting integrity No Consistent with NUREG-1801. The Bolting Integrity Program strength low-alloy bolts growth due to SCC includes the use of ISI to evaluate and monitor crack initiation and growth due to SSC in high strength low-alloy steel bolts used in NSSS component supports.
Page 3-281 Application for Renewed Operating License
Table 3.6-2 Structures and Component Supports - Component Types Subject to Aging Management not Evaluated in NUREG-1801 Component Types Material Environment AERMs Program/Activity Discussion (1) Aluminum-Indoor Aluminum Indoor (No Air No Aging Effects Structures The aging effects of flood barriers and electrical Conditioning) Monitoring conduit in areas containing safety related This generic asset Program equipment are considered in the Structures includes aluminum used Monitoring Program. Aluminum components in in flood barriers and indoor air have no aging effects. That aluminum conduit notwithstanding, the Structures Monitoring protected from the Program verifies that no unforeseen aging weather. mechanisms are causing aluminum degradation.
(2) ARCH-EXT ARCH-EXT Outdoor (exposed Hardening and Structures The aging effects of architectural materials used in includes the to the weather) Shrinkage due to Monitoring areas containing vital equipment are considered in The generic asset non-load bearing Weathering, Loss Program the Structures Monitoring Program. Architectural ARCH-EXT represents building elements of Material due to materials provide weather resistance and the non-safety weather not relied upon in General habitability control. By verifying the material barrier system that the safety Corrosion, condition of building roofing and siding systems provides shelter for analysis which Cracking due to the Structures Monitoring Program ensures safety related equipment provide normal Restraint, adverse weather conditions will not introduce from the elements and habitability control Shrinkage and unanticipated failures in vital plant systems.
allows for building and weather Creep habitability control. Load proofing, e.g.,
bearing frame members building siding, are evaluated in the built up roof building SS(CS)-INT systems, asset. windows, etc.
Materials vary but generally include:
Carbon Steel siding, Elastomer roof membranes, Build-up roof materials (insulating materials, stone, etc.) Glass Windows, Aluminum Flashing, Scuppers, downspouts, etc.
Page 3-282 Application for Renewed Operating License
Table 3.6-2 Structures and Component Supports - Component Types Subject to Aging Management not Evaluated in NUREG-1801 Component Types Material Environment AERMs Program/Activity Discussion (3) Cast Iron - Indoor Cast Iron Indoor (No Air Loss of Material Structures The aging effects of indoor cast iron components (Control Building Conditioning) due various Monitoring are considered in the Structures Monitoring de-watering Flapper mechanisms. Program Program. The Structures Monitoring Program Valve) verifies the integrity of the Control Building HELB pressurization wall. The wall includes the flapper valve that is credited in the CLB with ensuring a service or fire water line break in the control building mechanical equipment room will not cause a water buildup that has an adverse effect on the adjoining battery room.
(4) Cast Iron - Outdoor Cast Iron Outdoor (exposed Loss of Material Structures The aging effects of outdoor cast iron components (Duct Bank manhole to the weather) due various Monitoring are considered in the Structures Monitoring covers, roof drain pipes) mechanisms. Program Program. The Structures Monitoring Program evaluates essential yard components to ensure that structural integrity is not degraded.
(5) Elastomer- Indoor Elastomers (Butyl Indoor (No Air Cracking due to Structures The aging effects of indoor elastomers used in civil rubber, Neoprene, Conditioning) Thermal Stress, Monitoring features and non-NSSS system component (Door seals, flood Nitrile Rubber, Cracking due to Program supports are considered in the Structures barrier seals, refueling Silicone Rubber, Ultraviolet Monitoring Program. The Structures Monitoring cavity seal, seismic joint etc.) Radiation and Systems Program mandates inspection of the civil features (gap) filler, Ozone, Change in Monitoring and SSCs relied upon in the CLB to protect the racks/panels/electrical Material pubic. These inspections include verification of the enclosure gaskets and Properties due to integrity of elastomer materials used in the seals, etc.)This generic Thermal Stress associated SSC.
asset includes all elastomer (e.g., The aging effects of indoor elastomers used in vibration isolator cabinet seals and spray shields for electrical equipment mounts) that enclosures are considered in the Systems is indoor (i.e., protected Monitoring Program.
from the weather). Also The condition of the refueling cavity seal is included in this evaluated prior to installation.
evaluation are cabinet door seals, gaskets, and other seals. Fire barrier sealing material is evaluated as a separate commodity group.
Page 3-283 Application for Renewed Operating License
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Table 3.6-2 Structures and Component Supports - Component Types Subject to Aging Management not Evaluated in NUREG-1801 Component Types Material Environment AERMs Program/Activity Discussion (6) Elastomer- Outdoor Elastomers (Butyl Outdoor (exposed Cracking due to Structures The aging effects of outdoor elastomers used in (Flood barrier door rubber, Neoprene, to the weather) Thermal Stress, Monitoring civil features is considered in the Structures seals) Nitrile Rubber, Cracking due to Program Monitoring Program. The Structures Monitoring Silicone Rubber, Ultraviolet Program mandates inspection of the civil features etc.) Radiation and and SSCs relied upon in the CLB to protect the Ozone, Change in public. These inspections include verification of the Material integrity of elastomer materials used in the Properties due to associated SSC.
Thermal Stress Page 3-284 Application for Renewed Operating License
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Table 3.6-2 Structures and Component Supports - Component Types Subject to Aging Management not Evaluated in NUREG-1801 Component Types Material Environment AERMs Program/Activity Discussion (7) Structures and Concrete Submerged in raw All types of aging Periodic Concrete - Exterior above and below grade and Component Supports Submerged, water, outdoor effects, including Surveillance and exposed to flowing water: The Structures A6. Group 6 Structures Concrete-Outdoor (exposed to the loss of material Preventive Monitoring Program and Periodic Surveillance and (Water-Control (exposed to the weather) due to abrasion, Maintenance Preventive Maintenance Program identifies the Structures) weather) cavitation, and evidence that an aging mechanism is present and corrosion. Structures active and also provides conformation and All accessible/ Monitoring verification of the absence of all types of aging inaccessible concrete, Program effects. Indication of aging effects may be absent if steel and earthen the materials of construction, design specifications, components. and operational environment preclude an aging (Embedded steel, mechanism but, due to the long lead time reinforcement, and the necessary for some effects to manifest embedded portion of themselves, it is prudent to periodically assess the anchor bolts are condition of SSCs regardless of the likelihood that included.) a particular aging mechanism is present. The degradation of inaccessible concrete can create symptoms of aging effects that are detectable in accessible areas. Conversely, if aging effects are present in accessible areas it is sensible to extrapolate those effects into inaccessible areas and perform additional evaluations.
Concrete used in water control structures has been evaluated for the following aging mechanisms:
Aging Mechanism: Freeze-Thaw Aging Effect: Loss of Material Evaluation: The contract-specified air contents are within the range specified by current revisions of ACI 318, and the contract-specified water-to-cement ratio meets the recommendations of ACI 318-63 (< 0.53). Therefore, loss of material and cracking of concrete due to freeze-thaw are not probable aging effects at Ginna Station and have not been observed to date.
_____________________ l. ________________ _________________ ________________ I ________________ I Page 3-285 Application for Renewed Operating License
Table 3.6-2 Structures and Component Supports - Component Types Subject to Aging Management not Evaluated in NUREG-1801 Component Types Material Environment AERMs Program/Activity Discussion (7) (continued) Aging Mechanism: Leaching of Calcium Hydroxide Aging Effect: Change in Material Properties (Increase in porosity and permeability, loss of strength)
Evaluation: The original construction specifications met the intent of ACI 201.2R. Change in material properties due to leaching of calcium hydroxide is not a probable aging effect at Ginna Station and has not been observed to date.
Aging Mechanism: Reaction with Aggregates, Aging Effect: Cracking, Expansion Evaluation: During construction the aggregates were tested for potential reactivity in accordance with ASTM C227 and ASTM C295, cracking and expansion due to reaction with aggregates are not probable aging effects at Ginna Station and have not been observed to date.
Aging Mechanism: Corrosion of Embedded Steel Aging Effect: Loss of Material, Cracking, Loss of Bond Evaluation: Since the embedded steel is not exposed to an environment which is considered aggressive, loss of material, cracking, and loss of bond due to corrosion of embedded steel are not probable aging effects at Ginna Station and have not been observed to date. The concrete at Ginna Station was designed in accordance with ACl 301-66 or ACI 318-63. ACI 301-66 refers to ACI 318 for concrete reinforcement. Designing concrete to ACI 318 also provides for sufficient concrete cover over embedded steel to provide ample corrosion protection. Chemical analyses performed on the rock and groundwater indicate these environments are non-aggressive. Since the embedded steel is not exposed to an environment that is considered aggressive, corrosion of embedded steel is not a probable aging effect at Page 3-286 Application for Renewed Operating License
Table 3.6-2 Structures and Component Supports - Component Types Subject to Aging Management not Evaluated in NUREG-1801 Component Types Material Environment AERMs Program/Activity Discussion (7) (continued) Ginna Station and has not been observed to date.
Aging Mechanism: Aggressive Chemical Attack Aging Effect: Loss of Material (spalling and scaling), Changes in Material Properties (Increase in porosity and permeability, cracking)
Evaluation: Concrete degradation in air due to aggressive rainwater is insignificant and the below-grade/lake water environment is non-aggressive. Additionally, recent structural inspections revealed no evidence of degradation owing to aggressive chemical attack; therefore, loss of material and change in material properties due to aggressive chemical attack are not probable aging effects at Ginna Station and have not been observed to date.The Structures Monitoring Program requires periodic monitoring of ground/lake water to verify chemistry remains non-aggressive.
Aging Mechanism: Settlement Aging Effect: Cracking, Distortion, Increase in Component Stress Level Evaluation: All structures at Ginna Station are either founded on bedrock, steel foundation piles that are driven to bedrock, or have foundations that consist of caissons extending to bedrock.
Structural inspections indicate no visible evidence of settlement since construction of the station.
During the Systematic Evaluation Program, the NRC concluded that settlement of foundations and buried equipment is not a safety concern for Ginna Station. Cracking, distortion, and an increase in component stress levels due to settlement are not probable aging effects at Ginna Station and have not been observed to date. That notwithstanding, the Structures Monitoring Program monitors for cracks and distortion and contains inspection criteria to verify these aging effects are not developing.
Page 3-287 Application for Renewed Operating License
Table 3.6-2 Structures and Component Supports - Component Types Subject to Aging Management not Evaluated in NUREG-1801 Component Types [ Material Environment [ AERMs Program/Activity Discussion (7) (continued) Aging Mechanism: Erosion of porous concrete subfoundations Aging Effect: Reduction in foundation strength, cracking, differential settlement Porous concrete - Material not used at Ginna Station Evaluation: Reduction in foundation strength due to erosion of porous concrete subfoundations is not an aging effect requiring management at Ginna. Ginna Station's structure foundations are constructed of normal concrete and not the subject porous type, nor are foundations subject to water flowing under them. That notwithstanding, the Structures Monitoring Program monitors for settlement and cracking. The identification of indications of settlement by the Structures Monitoring Program, as well as the resistance provided by the materials of construction, provide adequate assurance that reductions in foundation strength for any reason will be identified and managed through out the extended period of operation.
Aging Mechanism: Cavitation Aging Effect: Loss of Material/Abrasion Evaluation: Flow velocities at the Screen House and Discharge canal are less than the values at which cavitation occur. Additionally, recent under water inspections of water control structures indicate no unusual concrete degradation due to abrasion or cavitation. Under water inspections are performed as a repetitive task as part of the Periodic Surveillance and Preventive Maintenance Program.
The Periodic Surveillance and Preventive Maintenance Program also inspects for silting and fouling of water control structures.Divers and submarine mounted cameras are used to inspect
________________________ J ____________________ __________________ __________________ __________________________________________________
Page 3-288 Application for Renewed Operating License
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Table 3.6-2 Structures and Component Supports - Component Types Subject to Aging Management not Evaluated in NUREG-1801 Component Types Material Environment AERMs I Program/ActivityI Discussion (7) (continued) the under water surfaces of the Screen House, discharge canal, canal valves and weir gates, and the intake tunnels and structure. Results of these inspections are reviewed by qualified engineers as part of the Structures Monitoring Program.
Operating experience has shown that Water Control Structures have not experienced unanticipated aging effects at Ginna Station. That notwithstanding, the identification of the above aging effects by the Structures Monitoring Program and Periodic Surveillance and Preventive Maintenance Program, as well as the resistance provided by the materials of construction provide adequate assurance that all types of concrete aging effects will be identified and managed Carbon Steel Submerged in raw Loss of Material Periodic through out the extended period of operation.
Components water, outdoor due to General Surveillance and (exposed to the Corrosion Preventive Aging Mechanism: General Corrosion weather) Maintenance Aging Effect: Loss of Material Evaluation: Carbon and low-alloy steel surfaces Structures that are exposed to outdoor and submerged Monitoring environments can experience loss of material from Program general, pitting and crevice corrosion. The Structures Monitoring Program and Periodic Surveillance and Preventive Maintenance Programs evaluate all carbon steel surfaces used for water control structures to ensure the aging Rock Outdoor (exposed Erosion, Structures effect is not progressing at an unanticipated rate.
to the weather) Movement Monitoring Program Large armor stones are used in a revetment that protects the plant from storm surges. The revetment received a site specific review from the Army Corp of Engineers for NRC use in the review of Systematic Evaluation Program topics: 11-3.A, 11-3.B and 11-3.C; "Hydrology, Flooding, and Ultimate Heat Sink." The Structures Monitoring Program and Periodic Surveillance and Preventive Maintenance Programs execute the Page 3-289 Application for Renewed Operating License
(
Table 3.6-2 Structures and Component Supports - Component Types Subject to Aging Management not Evaluated in NUREG-1801 Component Types Material Environment AERMs Program/Activity Discussion (7) (continued) recommendations made by the Corp by performing surveys and inspections of the armor stone and cap rocks to ensure erosion and stone movement do not compromise the effectiveness of the water control structure.
Thus, although Ginna Station does not utilize Reg.
Guide 1.127, "Inspections of Water-Control Structures Associated with Nuclear Power Plants" or utilize the Army Corp of Engineers for inspections and maintenance, the activities performed by the Structures Monitoring Program and Periodic Surveillance and Preventive Maintenance Program satisfy all the appropriate criteria and provide assurance that the intended function of water control structures will be maintained through the period of extended operation.
No masonry walls or earthen water control structures are used at Ginna Station.
(8) IB-LEAD-INT Lead Indoor (No Air No Aging Effects Structures Lead components in indoor air have no aging Conditioning) Monitoring effects. That notwithstanding, the Structures This generic asset Program Monitoring Program verifies that no unforeseen represents the lead in aging mechanisms are causing Lead degradation.
the shielded enclosure constructed over the primary sample containment isolation valves. The lead is held in place by a steel frame.
Page 3-290 Application for Renewed Operating License
Table 3.6-2 Structures and Component Supports - Component Types Subject to Aging Management not Evaluated in NUREG-1801 Component Types Material Environment AERMs Program/Activity Discussion (9) TUNNEL-SS(CS) Carbon Steel Buried (Below Loss of Material Structures Buried carbon steel components can experience PILES-BURIED Grade) due to General Monitoring loss of material from corrosion. The Cable Tunnel Corrosion, Loss of Program is founded on steel piles driven to bedrock. These This generic asset Material due to piles are inaccessible. The Structures Monitoring represents the carbon Pitting Corrosion, Program evaluates the effects of pile aging by steel piles the Cable Loss of Material monitoring the tunnel for signs of settlement which Tunnel is founded on. due to MIC would indicate foundation degradation. Site operating experience on sheet piles, below grade on one side and exposed to air on the other, has shown that only minimal loss of material has occurred since construction. Additionally, inspections of opportunity performed on other buried carbon steel components provide valuable information that may be used to infer the condition of inaccessible carbon steel piles. Thus, it can be concluded that the Structures Monitoring Program provides reasonable assurance that the aging effects of carbon steel piles will be managed through the period of extend operation.
(10) Pipe and valve Carbon Steel Buried (Below Loss of Material Structures Buried carbon steel components can experience Grade) due to General Monitoring loss of material from corrosion. The Cable Tunnel This asset represents Corrosion, Loss of Program has a cofferdam around its escape hatch. The the Cable Tunnel Material due to hatch cover has a rain collection gutter that drains escape hatch cofferdam Pitting Corrosion, through an open valve to the transformer yard.
drain. Loss of Material Immediately preceding potential floods the valve is due to MIC closed by operations. The pipe and valve are loosely covered in gravel. The Structures Monitoring Program is used evaluate the material condition of the valve and pipe and to verify valve operation.
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Table 3.6-2 Structures and Component Supports - Component Types Subject to Aging Management not Evaluated in NUREG-1801 Component Types Material Environment AERMs Program/Activity Discussion (11) Wood Wood Indoor (No Air No Aging Effects Structures Wood is used as a platform base for a non-safety Conditioning) Monitoring fan assembly. An angle iron support frame retains This asset represents Program the wood. Wood is also used as an electrical cable wood used for spacer. Research has shown that dry wood, component supports. maintained under cover and not exposed to parasites, will not rot or decay. Plant operating experience confirms this conclusion. That notwithstanding, wood used indoors is evaluated by the Structures Monitoring Program to ensure no unforeseen aging effects are causing degradation.
(12) FAST(HSLAS)-INT High Strength Indoor (No Air Loss of Material Structures The non-nuclear safety pressurization walls are Low Alloy Steel Conditioning) due to General Monitoring comprised of carbon steel sheet piles welded The generic asset Corrosion Program together to form a diaphragm.To provide stiffness FAST(HSLAS)-INT structural steel members are fastened with bolts represents the exposed horizontally across the diaphragm. The diaphragm portion of high strength is welded to the to the building structural steel carbon steel fasteners frame to provide the required barrier. These walls used in the construction are located between the Turbine Building and the of the Control Building Control and Diesel Generator Buildings. Inspection and Diesel Building high of the wall for signs of degradation is included in energy line break the Structures Monitoring Program.
pressure resistant wall.
(High strength fasteners used in NSSS component supports receive a separate evaluation.)
Page 3-292 Application for Renewed Operating License
Table 3.6-2 Structures and Component Supports - Component Types Subject to Aging Management not Evaluated in NUREG-1801 Component Types Material Environment AERMs Program/Activity Discussion (13) CV-BLOCK Masonry Block Indoor (No Air Cracking due to Structures Masonry wall inspections are incorporated into the Conditioning) restraint, Monitoring Structures Monitoring Program. The Structures This generic asset shrinkage and Program Monitoring Program effectively manages cracking includes all masonry creep due to restraint, shrinkage and creep.
block walls used in the Containment Vessel protected from the weather. The containment elevator shaft is block. Mortar is included in this asset evaluation.
(14) CV-INSULATION Plastic (PVC) Indoor (No Air No Aging Effects None Required The liner insulation is Vinylcel as manufactured by Conditioning) Johns-Manville. This material is a closed-cell This generic asset polyvinyl chloride foam insulation with low includes the conductivity, low water absorption, and high Containment Vessel strength. UFSAR Section 3.8.1.6.8, Liner thermal insulation Insulation, and Section 6.1.2.8.4, Vinylcel panels. Insulation, provide an extensive discussion concerning the selection criteria and properties of the containment liner insulation. Insulation panels are occasionally removed to gain access to the containment liner. Operating experience confirms that no aging management is required. That notwithstanding containment insulation will continue to be evaluated during inspections of opportunity in accordance with the guidance given in the Structures Monitoring Program.
Page 3-293 Application for Renewed Operating License
Table 3.6-2 Structures and Component Supports - Component Types Subject to Aging Management not Evaluated in NUREG-1801 Component Types Material Environment AERMs Program/Activity Discussion (15) CV-SS(CS)-INT Carbon Low Alloy Indoor (No Air Loss of Material Structures Carbon and low-alloy steel surfaces, which are Steel Conditioning) due to General Monitoring exposed to typical plant environments, can This generic asset Corrosion, Loss of Program experience general corrosion. Additionally, includes all carbon Material due to structural steel can be subject to boric acid structural steel of the Pitting Corrosion Boric Acid corrosion.
Containment Vessel that Corrosion is protected from the Corrosion of structural steels is an aging effect that weather. Columns, requires management at Ginna Station. Operating posts, beams, experience has shown that corrosion can be baseplates, bracing, initiated and/or accelerated by unique factors.
crane support girders, Accordingly, the Structures Monitoring Program crane rails, and the identifies and evaluates corrosion of structural exposed faces of plates steels used in containment. Additionally, and structural members accessible carbon low alloy structural steel located are included. The in areas that contain borated water systems are containment liner is subject to the requirements of Boric Acid Corrosion evaluated as a separate Program.
asset. This evaluation does not include carbon structural steel used as component supports.
(16) CV-FAST(CS)-INT Carbon Low Alloy Indoor (No Air Loss of Material Structures Carbon and low-alloy steel surfaces, which are Steel Conditioning) due to General Monitoring exposed to typical plant environments, can This generic asset Corrosion, Loss of Program experience general corrosion. Additionally, includes the exposed Material due to structural steel can be subject to boric acid portion of carbon steel Pitting Corrosion Boric Acid corrosion.
threaded fasteners for Corrosion the Containment Vessel Corrosion of structural steels is an aging effect that that are protected from requires management at Ginna Station. Operating the weather. The experience has shown that corrosion can be exposed portion of high initiated and/or accelerated by unique factors.
strength low alloy steel Accordingly, the Structures Monitoring Program fasteners are evaluated identifies and evaluates corrosion of structural in the component steels used in containment. Additionally, supports commodity accessible carbon low alloy structural steel located group in areas that contain borated water systems are subject to the requirements of Boric Acid Corrosion Program Page 3-294 Application for Renewed Operating License
( !
Table 3.6-2 Structures and Component Supports - Component Types Subject to Aging Management not Evaluated in NUREG-1801 Component Types Material Environment AERMs Program/Activity Discussion (17) CV-SS(SS)-INT Stainless Steel Indoor (No Air No Aging Effects Periodic The stainless steel refueling cavity liners are Conditioning) Surveillance and normally maintained dry and have no aging effects This generic asset Preventive that require management. That notwithstanding includes all stainless Maintenance plant operating experience shows that when the structural steel of the cavity is flooded for refueling, borated water weeps Containment Vessel that from welded seams and equipment attachment is protected from the points. The corrosion consequences for carbon weather. Included in this steel components in containment from these leaks evaluation is the warrant inspections. The Periodic Surveillance and refueling cavity and fuel Preventive Maintenance Program works to transfer liners (including implement the Boric Acid Corrosion Program and attachments). Stainless effectively manages the consequences of leaks steel expansion bellows from the cavity liners.
used with containment penetrations are addressed with the penetration.
Page 3-295 Application for Renewed Operating License
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Section 3.6 References
- 1. WCAP-14756-A, Aging Management Evaluation for Pressurized Water Reactor Containment Structure, May 2001.
- 2. WCAP-1 4422, Rev. 2-A, License Renewal Evaluation: Aging Management for Reactor Coolant System Supports, December 2000.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information 3.7 Aging Management of Electrical and Instrument and Controls Systems The results of the aging management review of the Electrical and Instrument and Control System components are provided in this section and summarized in Table 3.7-1 and Table 3.7-2. Table 3.7-1 shows the aging management of system components evaluated in NUREG-1 801 that are relied on for license renewal of the Electrical and Instrument and Control System components at Ginna. Included in the table is a discussion column. The discussion column will provide a conclusion indicating if the aging management evaluation results are consistent with NUREG-1 801 along with any clarifications or explanations required to support the stated conclusion if that conclusion is different than those of the NUREG. For a determination to be made that a table line item is "Consistent with NUREG-1 801" several criteria must be met. First the plant specific component is reviewed against the GALL to ensure that the component, materials of construction and internal or external service environment are comparable to those described in a particular GALL item. Second, for those that are comparable, the results of the plant aging management review- aging effect evaluation are compared to the aging effects/mechanisms in the GALL. Finally, the programs credited in the GALL for managing those aging effects are compared to the programs invoked in the plant evaluation. If, using good engineering judgment, it could be reasonably concluded that the plant evaluation is in agreement with the GALL evaluation a line item was considered consistent with NUREG-1 801. There are cases where components and component material/environment combinations and aging effects are common between a NUREG-1 801 line item and the plant evaluation but the aging management program selections differ. In those cases the discussion column will indicate the plant aging management program selection but no conclusion will be made that the line item is consistent with the GALL. Table 3.7-2 contains the Electrical and Instrument and Control System components aging management review results that are not addressed in NUREG-1 801. A plant component is considered not addressed by the NUREG if the component type is not evaluated in the GALL or has a different material of construction or operating environment than evaluated in the GALL. This table includes the component types, materials, environments, aging effects requiring management, the programs and activities for managing aging, and a discussion column. To avoid confusion, no attempt was made to interrelate material/environment/aging effects from one NUREG-1 801 chapter to another. Note that these tables only include those components, materials and environments that are applicable to a PWR.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Materials The materials of construction of a component have a major influence on the evaluation of aging effects applicable to the component. Sources of information used to identify materials of construction include original equipment specifications, vendor technical manuals and drawings, fabrication drawings, cable circuit schedules, design records and field walkdowns/verifications. The tables below account for the materials of construction for the components requiring an aging management review. Since similar materials are susceptible to the same aging effects/mechanisms, the tables itemize the component types (i.e., groupings) while factoring in the materials of construction. Specific materials of construction were not used to determine the scope of components in the Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Program. The program includes all in-scope, electrical cables and connections within specified plant spaces. Therefore further review of materials of construction was not required for electrical cables and connections.
Environment As previously described, the environment(s) to which components are exposed are critical in the determination of potential aging mechanisms and effects. A review of plant design documentation was performed to quantify the environmental conditions to which Ginna Station equipment is exposed. This review identified that some equipment is exposed to a variety of environments. This can include normal operating conditions and post accident conditions. Since aging mechanisms and effects will be primarily driven by the environmental conditions to which equipment is exposed on a daily basis, under normal operating conditions, these conditions will differ from the design parameters which are established based upon the worst case scenario (e.g., LOCA conditions). Ginna Station equipment environments may be categorized into basic external and internal environments detailed in Section 3.1.2.
Since passive electrical components do not have internal environments in the same sense that mechanical components do, the term self-heating is used to describe the effect of resistive heating that may occur in electrical components. This heating can result in the component having a service temperature that is hotter than the ambient conditions.
Self heating is only applicable to components that carry significant current, and therefore instrumentation circuits, such as Resistance Temperature Detectors (RTDs),
Thermocouples, and related loop wiring is considered not to be subject to self-heating.
For those components that are subject to self-heating (i.e., power cables, phase bus) the ultimate temperature is a result of the ambient temperature and the square of the ratio of the actual current to the ampacity of the conductor. Plant design guidelines do not Page 3-298
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information normally permit a cable to be loaded to more than 80% of ampacity. Ginna Station has identified specific plant spaces that may lead to cables exceeding 80% of ampacity due to cable tray fill deratings. These areas have been included in the scope of the Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Program.
For the review of passive electrical commodities, the environments discussed in Section 3.1.2, were summarized for analysis as follows:
Environment Temperature Radiation Inside Containment Sheltered 60-120°F 1 R/hr outside of loop areas, 10 R/hr inside loop areas (assumed)
Outside Containment Sheltered 60-104°F 10 mR/hr Underground Bus Duct Not Applicable Outdoor Exposed Not Applicable Most organic materials used as electrical insulators were first evaluated against the environment "Inside Containment Sheltered" to determine if aging effects required management. This environment is considered conservative for normal ambient temperature and radiation dose.
Ginna includes limited installations of underground passive electrical components (cables). Of those, Ginna has four medium voltage power cables installed in underground duct banks. The functions of these cables were reviewed and determined that a failure of the cable would not prevent the satisfactory accomplishment of any intended function.
Therefore a further review of this environment was not required.
The switchyard components and a limited number of cables are addressed in the environment of "Outdoor Exposed." These components are subject to normal environmental conditions, including precipitation.
Aging Effects Requiring Management After the components requiring aging management review were identified and grouped by materials of construction and environment, a review of industry and plant-specific operating experience was performed. The purpose of this review was to assure that all applicable aging effects were identified, and to evaluate the effectiveness of existing aging management programs. This experience review was performed utilizing various Page 3-299
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information industry and plant-specific programs and databases. Industry operating experience sources included NRC Generic Publications (including Information Notices, Circulars, Bulletins, and Generic Letters), INPO Significant Operating Event Reports (SOER), EPRI Technical Reports, and other information sources, such as the Sandia Aging Management Guidelines for Electrical Cable and Terminations, Westinghouse Generic Technical Reports (GTRs), and the Generic Aging Lessons Learned (GALL) report. Plant specific operating experience sources included Semi-annual and Annual Reports to AEC/NRC, Abnormal Occurrence and Licensee Event Reports (LERs),
Non-Conformance Reports (NCRs), Corrective Action Reports (CARs), Refueling, Inspection and Overhaul Reports (RIOs), Inservice Inspection (ISI) Reports, Identified Deficiency Reports (IDRs), and ACTION Reports (ARs) from 1969 to the present.
Information from these sources was compiled in various databases. Based upon the material of construction, the applicable environments, and operating experience the potential aging effects requiring management for each of the components was identified as documented in Table 3.7-1 and Table 3.7-2.
The most common aging effect for passive electrical components is electrical failure due to thermal/thermoxidative degradation of organics. Thermal life was evaluated using methodology similar to Appendix G of the Sandia Aging Management Guideline for Electrical Cables and Connections (Reference 1). In many cases, conservative assumptions were used to simplify the analysis. Thermal life was not used to determine the scope of components in the Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Program. The program includes all in-scope, electrical cables and connections within specified plant spaces, and adequately addresses aging effects due to thermal conditions. Therefore further review of the thermal aging results was not required for electrical cables and connections.
Electrical failure due to radiolysis and radiation induced oxidation is considered a significant aging effect only for those passive electrical components installed in containment. For these components, the moderate damage threshold for the materials were reviewed against expected radiation environments. Although the review identified very few susceptible materials, the results of the review were not used to determine the scope of the components in the Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Program. The program includes all in-scope, electrical cables and connections within specified plant spaces, and adequately addresses aging effects due to radiation. Therefore further review of the radiation induced aging effects was not required for electrical cables and connections.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Moisture induced electrical failure for in-scope passive electrical components is not considered to be a significant aging effect at Ginna Station. While medium voltage cables are known to experience water-treeing in a wet environment, the only environment that may qualify as wet is the underground duct banks. As discussed previously, there are no in-scope medium voltage cables in the underground duct banks. Industry and plant operating experience does not support moisture as a significant stressor for other passive electrical components. The results of this review were not used to determine the scope of the components in the Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Program. The program includes all in-scope, electrical cables and connections within specified plant spaces, and adequately addresses aging effects due to moisture. Therefore further review of moisture induced aging effects was not required for electrical cables and connections.
Aging effects for components not included in NUREG-1801 were identified and included on Table 3.7-2. Using basic materials properties and operating experience as a basis, these aging effects were not determined to require aging management for the period of extended operation. Nonetheless, the relevant aging effects were listed in Table 3.7-2 in the column, "AERMs" indicating that they are aging effects requiring management.
Therefore the aging effects listed in the AERM column have been provided for conservatism to indicate those aging effects that have been evaluated for the components listed in the table.
Time-Limited Aging Analysis In addition to those identified in NUREG-1 801, any additional time-limited aging analyses (TLAA) identified as appropriate to the system are identified in Section 4.0.
A description of the aging management activities for this area are provided in Appendix B, along with the demonstration that the identified aging effects will be managed for the period of extended operation.
Therefore, based on the demonstrations provided in Appendix B, the effects of aging associated with the Electrical and Instrumentation and Control System components will be adequately managed so that there is reasonable assurance that the intended function(s) will be maintained consistent with the current licensing basis during the period of extended operation.
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R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Conclusion The programs and activities selected to manage the aging effects of the Electrical and Instrument and Control Systems are identified in Table 3.7-1 and Table 3.7-2. A description of these aging management activities is provided in Appendix B, along with the demonstration that the identified aging effects will be managed for the period of extended operation.
Therefore, based on the demonstrations provided in Appendix B, the effects of aging associated with the Electrical and Instrument and Control Systems components will be adequately managed so that there is reasonable assurance that the intended function(s) will be maintained consistent with the current licensing basis during the period of extended operation.
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Table 3.7-1 Electrical and Instrumentation and Controls Systems - Aging Management Programs Evaluated in NUREG-1801 that are Relied on for License Renewal Aging Further Aging Management Evaluation Component Effect/Mechanism Programs Recommended Discussion (1) Electrical equipment Degradation due to Environmental Yes, TLAA Evaluation of Time-Limited Aging Analyses for EQ equipment subject to 10 CFR 50.49 various aging qualification of is provided in application Section 4.4.
environmental qualification mechanisms electric (EQ) requirements components (2) Electrical cables and Embrittlement, Aging management No Consistent with NUREG-1801. The Electrical Cables and connections not subject to 10 cracking, melting, program for Connections Not Subject to 10 CFR 50.49 Environmental CFR 50.49 EQ requirements discoloration, electrical cables Qualification Requirements Program will adequately manage swelling, or loss of and connections the potential aging effects for this component. An analysis of dielectric strength not subject to 10 material/environment combinations for normal plant leading to reduced CFR 50.49 EQ environments indicates that a large majority of components insulation requirements have no aging effects that require management throughout the resistance (IR); period of extended operation. Exceptions to this include PVC electrical failure cables in containment subject to heating above ambient caused by thermal/ temperatures (self-heating), and cables installed in adverse thermoxidative localized equipment environments. However, due to plant degradation of specific operating experience, all material/environment organics; radiolysis combinations will be included in the scope of the program and photolysis using an encompassing approach.
(ultraviolet [UV]
sensitive materials only) of organics; radiation-induced oxidation; moisture intrusion Page 3-303 Application for Renewed Operating License
Table 3.7-1 Electrical and Instrumentation and Controls Systems - Aging Management Programs Evaluated in NUREG-1801 that are Relied on for License Renewal Aging Further Aging Management Evaluation Component Effect/Mechanism Programs Recommended Discussion (3) Electrical cables used in Embrittlement, Aging management No Not consistent with NUREG-1 801. Rochester Gas and Electric instrumentation circuits not cracking, melting, program for (RG&E) believes that invoking the NUREG-1801 XI.E1, subject to 10 CFR 50.49 EQ discoloration, electrical cables Electrical Cables and Connections Not Subject to 10 CFR requirements that are sensitive swelling, or loss of used in 50.49 Environmental Qualification Requirements Program to to reduction in conductor dielectric strength instrumentation manage the effects of aging in accessible non-EQ cable and insulation resistance (IR) leading to reduced circuits not subject connectors provides reasonable assurance that these SC's will IR; electrical failure to 10 CFR 50.49 perform their intended function during the period of extended caused by thermal/ EQ requirements operation. The aging effects of cable and connector insulation thermoxidative may have very long incubation periods. In essence, routine degradation of maintenance, calibration and repair activities on the active organics; radiation- components in an instrument loop initially work to mask induced oxidation; indications of cable and cable and connector insulation moisture intrusion degradation. Only after the active portions of a loop can no longer be adjusted to compensate for cable and connector degradation would the passive portions of the instrument loop become suspect. Surveillance provides meaningful information, but that information is primarily used to cause changes to the active portions of an instrument loop. The predominate cause of non-event driven degradation in cable and connector insulation is thermal aging. External inspection of cables and connectors and their host environments identifies the possibility of thermal aging long before instrument loop adjustments can't compensate for current leakage. Because of this, RG&E feels that the only legitimate way to ensure the continued functioning of the long-lived passive components are those inspection activities performed under the XI.E1 program, Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements.
Page 3-304 Application for Renewed Operating License
Table 3.7-1 Electrical and Instrumentation and Controls Systems - Aging Management Programs Evaluated in NUREG-1801 that are Relied on for License Renewal Aging Further Aging Management Evaluation Component Effect/Mechanism Programs Recommended Discussion (4) Inaccessible Formation of water Aging management No Not Applicable. Medium voltage cables and connections medium-voltage (2 kV to 15 trees, localized program for subject to an aging management review are not installed in kV) cables (e.g., installed in damage leading to inaccessible environments that lead to the formation of water trees therefore conduit or direct buried) not electrical failure medium-voltage no aging management program is required. That subject to 10 CFR 50.49 EQ (breakdown of cables not subject notwithstanding, should aging effects be observed in requirements insulation); water to 10 CFR 50.49 accessible medium voltage cables the Electrical Cables and trees caused by EQ requirements Connections Not Subject to 10 CFR 50.49 Environmental moisture intrusion Qualification Requirements Program will require that inaccessible cables be evaluated to ensure no adverse aging effects are developing (5) Electrical connectors not Corrosion of Boric acid corrosion No Consistent with NUREG-1801. Corrosion of connectors due to subject to 10 CFR 50.49 EQ connector contact Boric Acid Corrosion is an aging effect requiring management.
requirements that are exposed surfaces caused by The Boric Acid Corrosion Program effectively manages to borated water leakage intrusion of borated corrosion of contact surfaces caused by the intrusion of water borated water.
Page 3-305 Application for Renewed Operating License
Table 3.7-2 Electrical and Instrumentation and Controls Systems - Component Types Subject to Aging Management not Evaluated in NUREG-1801 Component Types Material Environment AERMs Program/Activity Discussion (1) Electrical Aluminum, Copper Indoor (No Air Embrittlement, One-Time Based on a materials analysis, the high reliability of Phase Bus (bus, solid and Conditioning), cracking, melting, Inspection Program the phase bus at Ginna, and partial upgrades flexible connectors Outdoor discoloration, performed on the 4KV system for redundancy and straps), Steel swelling, or loss of purposes, there is no reason to believe that there (bolts, washers, dielectric strength are aging effects requiring management for nuts, etc.), Rigid leading to reduced electrical phase bus. A review of industry operating Bus parts insulation experience indicated that plants have experienced (porcelain resistance (IR); failures of electrical phase bus. The review insulators, etc) electrical failure concluded that most failures were not due to aging caused by thermal/ mechanisms. However as a confirmatory measure, thermoxidative an inspection was performed on selected portions degradation of (both indoors and outdoors) of the 4KV bus duct at organics; moisture Ginna Station during the 2002 RFO. The results intrusion. showed that the internal components were in "like new" condition. Because no aging effects requiring management were identified, Ginna Station considers the One-Time Inspection results adequate to demonstrate that no additional aging management programs are required through the period of extended operation.
Page 3-306 Application for Renewed Operating License
Table 3.7-2 Electrical and Instrumentation and Controls Systems - Component Types Subject to Aging Management not Evaluated in NUREG-1801 Component Types Material Environment AERMs Program/Activity Discussion (2) Switchyard Bus Copper, Copper Outdoor Loss of material Not Applicable Rochester Gas and Electric's Energy Delivery Alloy, Steel (bolts, due to corrosion Department performs inspection and maintenance washers, nuts, etc.) leading to of the Switchyard Bus components. Switchyard Bus increased components subject to aging management review resistance. contain materials that when exposed to plant operating environments could potentially lead to aging effects requiring management. Plant Operating Experience reviews have not identified any case where aging effects requiring management have developed however, evidence of aging effects may in fact be removed (masked) by ongoing routine Energy Deliver Department maintenance activities. That notwithstanding, the Energy Delivery Department inspections identify if the evidence of an aging mechanism is present and active and also provides the confirmation and verification of the absence of all types of aging effects. Indication of aging effects may be absent if the materials of construction and operational environment preclude an aging effect but, due to the long lead time necessary for some effects to manifest themselves, it is prudent to periodically assess the condition of SSCs regardless of the likelihood that a particular aging mechanism is applicable.
Plant operating experience reviews show that the activities performed by the Energy Delivery Department on the Switchyard Buses are effective in managing Switchyard Bus components. The Maintenance Rule activities monitor the effectiveness of the Energy Delivery Department Activities by tracking system level performance indicators.
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Table 3.7-2 Electrical and Instrumentation and Controls Systems - Component Types Subject to Aging Management not Evaluated in NUREG-1801 Component Types Material Environment AERMs Program/Activity Discussion (3) High Voltage Porcelain, Cement, Outdoor Cracks; Loss of Not Applicable Rochester Gas and Electric's Energy Delivery Insulators Steel material due to Department performs inspection and maintenance corrosion; loss of of the High Voltage Insulators. High Voltage dielectric strength Insulator components subject to aging management leading to reduced review contain materials that when exposed to plant insulation operating environments could potentially lead to resistance (IR) aging effects requiring management. Plant Operating Experience reviews have not identified any case where aging effects requiring management have developed however, evidence of aging effects may in fact be removed (masked) by ongoing routine Energy Deliver Department maintenance activities. That notwithstanding, the Energy Delivery Department inspections identify if the evidence of an aging mechanism is present and active and also provides the confirmation and verification of the absence of all types of aging effects. Indication of aging effects may be absent if the materials of construction and operational environment preclude an aging effect but, due to the long lead time necessary for some effects to manifest themselves, it is prudent to periodically assess the condition of SSCs regardless of the likelihood that a particular aging mechanism is applicable.
Plant operating experience reviews show that the activities performed by the Energy Delivery Department on the High Voltage Insulators are effective in managing Phase Bus components. The Maintenance Rule activities monitor the effectiveness of the Energy Delivery Department Activities by tracking system level performance indicators.
Page 3-308 Application for Renewed Operating License
R. E. Ginna Nuclear Power Plant Application for Renewed Operating License Technical and Administrative Information Section 3.7 References
- 1. SAND 96-0344, "Aging Management Guideline for Commercial Nuclear Power Plants Electrical Cable and Terminations," Sandia National Laboratories for the U. S.
Department of Energy, September 1996.
- 2. NUREG-1 801, "Generic Aging Lessons Learned (GALL)," U. S. Nuclear Regulatory Commission, April 2001.
- 3. NUREG-1 800, "Standard Review Plan for the Review of License Renewal Applications for Nuclear Power Plants," U. S. Nuclear Regulatory Commission, April 2001.
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