Text

Response to Request for Additional Information, License Renewal Application
	ML102590047
Person / Time
Site:	Columbia
Issue date:	09/13/2010
From:	Gambhir S; Energy Northwest
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	GO2-10-135
	Download: ML102590047 (56)
	v • d • e

Sudesh K. Gambhir ENERGY Columbia Generating Station P.O. Box 968, PE04 NORTHWEST Richland, WA 99352-0968 Ph. 509.377.8313 I F. 509.377.2354 sgambhir@ energy-northwest.com September 13, 2010 G02-10-135 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001

Subject:

COLUMBIA GENERATING STATION, DOCKET NO. 50-397 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION LICENSE RENEWAL APPLICATION

References:

1) Letter, G02-1 0-011, dated January 19, 2010, WS Oxenford (Energy Northwest) to NRC, "License Renewal Application"

2) Letter dated July 15, 2010, NRC to WS Oxenford (Energy Northwest), "Request for Additional Information for the Review of the Columbia Generating Station, License Renewal Application,"

(ADAMS Accession No. ML101900125)

Dear Sir or Madam:

By Reference 1, Energy Northwest requested the renewal of the Columbia Generating Station (Columbia) operating license. Via Reference 2, the Nuclear Regulatory Commission (NRC) requested additional information related to the Energy Northwest submittal.

Transmitted herewith in the Attachment is the Energy Northwest response to the Request for Additional Information (RAI) contained in Reference 2. Enclosure 1 contains Amendment 7 to the License Renewal Application (LRA) that was submitted in Reference 1.

Ao-)'ýS (upIVC

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION LICENSE RENEWAL APPLICATION Page 2 of 2 Commitment 49 in LRA Appendix A, Table A-1 is revised. It is provided in Enclosure 1, LRA Amendment 7. No new commitments are included in this response.

If you have any questions or require additional information, please contact Abbas Mostala at (509) 377-4197.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the date of this letter.

4ecffully, Sl;Gambhir Vice President, Technical Services

Attachment:

Response to Request for Additional Information

Enclosure:

License Renewal Application Amendment 7 cc: NRC Region IVAdministrator NRC NRR Project Manager NRC Senior Resident Inspector/988C EJ Leeds - NRC NRR EFSEC Manager RN Sherman - BPA/1 399 WA Horin - Winston & Strawn EH Gettys - NRC NRR (w/a)

BE Holian - NRC NRR RR Cowley - WDOH

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION LICENSE RENEWAL APPLICATION Page 1 of 11 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION RAI B.2.A-1:

Request for Additional Information (RAI) for the Following Aging Management Programs (AMPs):

B.2.12 ("Chemistry Program Effectiveness Inspection")

B.2.14 ("Cooling Units Inspection")

B.2.16 ("Diesel Starting Air Inspection")

B.2.17 ("Diesel System Inspection")

B.2.18 ("Diesel Driven Fire Pumps Inspection")

B.2.27 ("Flexible Connection Inspection")

B.2.30 ("Heat Exchangers Inspection")

B.2.37 ("Lubricating Oil Inspection")

B.2.41 ("Monitoring and Collection Systems Inspection")

B.2.48 ("Service Air Inspection")

B.2.51 ("Supplemental Piping/Tank Inspection")

Background:

Underthe program element "parameters monitored or inspected," the above applicant AMPs state that "inspections will be performed by qualified personnel using established NDE techniques." GALL Report AMP XI.M32 ("One-Time Inspection") states that inspections are to be performed by qualified personnel following procedures consistent with the requirements of the ASME Code and 10 CFR 50, Appendix B.

Issue:

The above applicant AMPs do not specifically state that the inspections will be performed in conformance with the requirements of the ASME Code nor does the AMP or AMP basis document reference a site-specific procedure outlining the training requirements for individuals assigned to perform inspections in accordance with the above AMPs.

Request:

Provide details regarding the training requirements for individuals assigned to perform the inspections described in the above AMPs. Describe how this training program is consistent with training requirements stipulated in the applicable ASME Code requirements.

Energy Northwest Response:

During the implementation phase, following receipt of the renewed license, approved procedures will be revised or developed to direct these one-time inspections. Per plant procedures, personnel performing VT-1, 2, or 3 inspections, ultrasonic examinations in

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION LICENSE RENEWAL APPLICATION Page 2 of 11 accordance with ASME Section XI, and other nondestructive examinations in accordance with ASME Section V or Section XI are required to be trained in accordance with ASME requirements. Training requirements for all inspections will be consistent with the requirements of 10 CFR 50 Appendix B.

RAI B.2.A-2:

RAI for the Following AMPs:

B.2.12 ("Chemistry Program Effectiveness Inspection")

B.2.14 ("Cooling Units Inspection")

B.2.16 ("Diesel Starting Air Inspection")

B.2.17 ("Diesel System Inspection")

B.2.18 ("Diesel Driven Fire Pumps Inspection")

B.2.27 ("Flexible Connection Inspection")

B.2.30 ("Heat Exchangers Inspection")

B.2.37 ("Lubricating Oil Inspection")

B.2.41 ("Monitoring and Collection Systems Inspection")

B.2.48 ("Service Air Inspection")

B.2.51 ("Supplemental Piping/Tank Inspection")

Background:

Under the "detection of aging effects" program element, the above applicant AMPs states that "a sample population will be determined by engineering evaluation based on sound statistical sampling methodology." The GALL Report AMP XI.M32 ("One-Time Inspection") states under this program element that "the inspection includes a representative sample of the system population, and where practical, focuses on the bounding or lead components most susceptible to aging due to time in service, severity of operating conditions, and lowest design margin."

Issue:

Although the above AMPs require the inspection of a representative sample of the material and environment combinations for systems within the scope of the program, they present no details of the proposed sampling plan and provide no assurance that a representative population of sufficient size and scope will be inspected.

Request:

Describe the sampling methodology, including how the population for each of the material-environment-aging effect combinations is being selected, and what type of engineering, design, or operating experience considerations would be used to select the sample of components for both the scheduled and supplemental inspections.

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION LICENSE RENEWAL APPLICATION Page 3 of 11 Energy-Northwest Response:

During the implementation phase, following receipt of the renewed license (but prior to the period of extended operation), engineering evaluations will be conducted to determine a sample population for each material and environment combination within the scope of the individual One-Time Inspections. This sample population will be based on sound statistical sampling methodology and, where practical, focused on the components most susceptible to aging. The basis for determining the susceptibility of components to aging would include their time in service, the severity of conditions during normal plant operations, and design margins.

To provide consistency in determination of sample size, guidance that will be used during one-time inspections is provided below. This sample size guidance is intended for use in determining an appropriate sample size for those inspection programs that do not already specify sample size criteria.

For mechanical AMPs, required piping inspections will be performed as part of the inspections of discrete components included in a pipe run. For example, if a pipe run includes a population of four valves requiring inspection, the sample size guidance indicates that inspection of one valve is sufficient. Visual inspection of the piping would be performed by looking upstream and downstream of the valve when it is disassembled for inspection. If volumetric inspection (e.g., ultrasonic testing (UT)) of the piping is required by the AMP, then UT inspection of the accessible piping on both sides of the valve would be performed. An inspection of the adjacent pipe segments is performed for each discrete component included in the inspection sample.

Supplemental inspections sample size would be determined as part of the corrective action program.

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION LICENSE RENEWAL APPLICATION Page 4 of 11 No Yes No No Inspection of piping and piping components adjacent to discretb components wilIbe performed along with thediscrete component inspections. Accessible piping will be visually inspected and non-destructive testing performed as required by the aging management program.

Once the'sample size of each populatiori is deterrmined; the actual*sample locations are determined based on the bounding or lead component'most susceptible to aging due to time in service and severity of operating conditions, e.g. low or stagnant flow. Opportunistic inspections can be substituted and applied to the sample as long as they represent, the materiallenvironm ent exj ectedfor. that sanm p le.

'1) Populatioh is basedton component material, internal/exteralrenvirgnment, and comrponent type.,

2) A fraction.is always rounded up to a.whole value when determining sample size:

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION LICENSE RENEWAL APPLICATION Page 5 of 11 RAI B.2.49-1:

Background:

A review of operating experience at Columbia Generating Station (CGS) has indicated that multiple failures of small-bore socket welds have occurred at CGS due to cracking.

Issue:

GALL Report,Section XI.M35 recommends the use of the One-Time Inspection of ASME Code Class 1 small-bore piping only for those plants that have not experienced cracking of ASME Code Class 1 small-bore piping resulting from stress corrosion or thermal and mechanical loading. For those plants that have experienced cracking, the GALL Report recommends periodic inspection of the subject piping to be managed by a plant-specific AMP.

Request:

Since cracking of socket welds has occurred at CGS, either provide a plant-specific AMP that includes periodic inspections to manage cracking, or provide justification why a plant-specific AMP that includes periodic inspections is not necessary for ASME Code Class 1 small-bore piping.

Energy Northwest Response:

Since cracking of socket welds has occurred at CGS, an aging management program, Small Bore Class 1 Piping Program, will be developed and implemented to detect and characterize cracking of small bore Class 1 piping components that are exposed to reactor coolant.

This periodic program will provide physical evidence as to whether, and to what extent, cracking due to stress corrosion cracking or thermal or mechanical loading has occurred in small bore Class 1 piping components. It will also verify, by inspections for cracking, that reduction of fracture toughness due to thermal embrittlement requires no additional aging management for small Class 1 cast austenitic stainless steel valve bodies. The Small Bore Class 1 Piping Program will be a condition monitoring program with no actions to prevent or mitigate aging effects. The program will include visual and volumetric inspection of a representative sample of small bore Class 1 piping, including butt welds and socket welds.

The Small Bore Class 1 Piping Program is a new program that will be implemented prior to the period of extended operation. Inspection activities will start during the fourth 10-year inservice inspection interval and continue through the period of extended operation. The Small Bore Class 1 Piping Program will credit portions of the Inservice Inspection Program. The Small Bore Class 1 Piping Program will verify the effectiveness of the BWR Water Chemistry Program in mitigating cracking of small bore piping and piping components.

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION LICENSE RENEWAL APPLICATION Page 6 of 11 The Small Bore Class 1 Piping Program is added in Amendment 7 to the LRA and is provided in Enclosure 1 to this letter.

RAI B.2.26-3:

Background:

Standard Review Plan-License Renewal (SRP-LR) Table 3.3-1, item 84 addresses copper alloy with >15% zinc piping, piping components, piping elements, and heat, exchanger components exposed to raw water, treated water, or closed cycle cooling water, and recommends GALL AMP XI.M33, ("Selective Leaching of Materials Program") to manage loss of material due to selective leaching for these components.

GALL AMP XI.M33 recommends managing selective leaching of components using a one-time visual inspection and hardness measurement of selected components to determine whether selective leaching is occurring. GALL AMP XI.M27, "Fire Water System Program," does not include activities to manage loss of material due to selective leaching.

License renewal application (LRA) Section B.2.26 states an enhancement to the

,acceptance criteria" program element of its Fire Water System Program to include

,hardness testing, or equivalent, on its sprinkler heads as part of its NFPA testing in

-order to manage selective leaching. The sprinkler heads are constructed of copper

alloy with >15% zinc.

Issue:

GALL AMP XI.M33 states that if selective leaching is occurring, an engineering evaluation is initiated to determine acceptability of the affected components for further service. GALL AMP XI.M33 also states that, if necessary, the evaluation will include a root cause analysis. In its review of the "detection of aging effects" and "acceptance criteria" program elements described in LRA Section B.2.26 for the Fire Water System Program, and in its basis document the staff noted that the AMP did not include any details regarding how to detect whether selective leaching has occurred or any acceptance criteria or follow-up evaluations that will be performed if selective leaching is identified.

Request:

Describe how selective leaching of copper alloy with >15% zinc components will be detected by hardness testing or equivalent methods, and define the acceptance criteria and follow-up actions that will be implemented if selective leaching is identified as part of the activities performed in the Fire Water System Program.

Energy Northwest Response:

NUREG-1801, Rev 1,Section XI.M33 states in the Detection of Aging Effects element that the one-time visual inspection and hardness measurement includes close

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION LICENSE RENEWAL APPLICATION Page 7 of 11 examination of a select set of components to determine whether selective leaching has occurred and whether the resulting loss of strength and/or material will affect the intended functions of these components during the period of extended operation.

The Fire Water Program conducts visual inspections of sprinklers and the associated piping and fittings, as well as inspections of hose station valves and fire hydrants, conducted at periodic intervals to look for corroded, damaged, or obstructed components. To provide confirmation of the condition of sprinkler heads, either all sprinkler heads will be replaced or hardness testing of a sample number of sprinkler heads will be performed as part of the Fire Water Program. This hardness testing, or an NRC-approved alternative, will determine whether, and to what extent, a loss of material due to selective leaching has occurred or is likely to occur that could result in a loss of intended function. Sampling (or replacement) of 50-year service life sprinkler heads will be performed in accordance with the acceptance criteria to be established based on applicable NFPA standards and field service laboratory testing procedures. Evaluation of the test results by engineering will determine if they are acceptable. Follow-up actions as a result of the engineering evaluation will be implemented as described in LRA Section B.1.3.

RAI 3.3.2.2.13-1:

,

Background:

GALL Report Section IX.F defines wear as follows: "Wear is defined as the removal of surface layers due to relative motion between two surfaces or under the influence of hard abrasive particles. Wear occurs in parts that experience intermittent relative motion, frequent manipulation, or in clamped joints where relative motion is not intended:

but may occur due to a loss of the clamping force." SRP-LR, Section 3.3.2.2.13, states that loss of material due to wear could occur in the elastomer seals and components exposed to air-indoor uncontrolled (internal or external). The GALL Report recommends further evaluation to ensure that these aging effects are adequately managed. Acceptance criteria are described in Branch Technical Position RLSB-1 (Appendix A.1 of this SRP-LR.)

LRA Section 3.3.2.2.13 refers to LRA Table 3.3.1, item 3.3.1-34, and addresses loss of material due to wear for elastomer seals and components exposed to air-indoor uncontrolled (internal or external). The applicant addressed the further evaluation criteria of the SRP-LR by stating that wear of elastomer seals and components exposed to air-indoor was not identified as an aging effect requiring management. The applicant also stated that loss of material due to wear is the result of relative motion between two surfaces in contact and wear occurs during the performance of an active function as a result of improper. design, application or operation; or to a very small degree with insignificant consequences, and therefore, loss of material due to wear is not an aging effect requiring management for elastomers exposed to air-indoor uncontrolled.

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION LICENSE RENEWAL APPLICATION Page 8 of 11 Issue:

The staff does not have sufficient information to determine that improper design, application or operation is not a factor resulting in loss of material due to wear in elastomeric components exposed to air-indoor uncontrolled.

Request:

Provide justification why improper design, application or operation is not a factor resulting in the loss of material due to wear for elastomeric components exposed to an air-indoor uncontrolled environment, or provide an AMP to manage this aging effect.

Energy Northwest Response:

Consistent with the Statements of Consideration for 10 CFR 54, Section Ill.d.(i),

improper design, faulty manufacturing processes, improper application, faulty maintenance, improper operation, or personnel errors may cause events that result in significant wear of components, but this cause of degradation is not aging related.

Therefore, loss of material due to wear is not an applicable aging effect for the elastomeric components that are subject to an air-indoor uncontrolled environment.

'RAI 3.3.2.3-1:

Background:

The GALL Report does not contain a recommended AMR line item for elastomers exposed to lubricating oil or fuel oil. However, in LRA Tables 3.3.2-20 and 3.3.2-22, the applicant stated that for elastomer flexible connections exposed to lubricating or fuel oil there is no aging effect and no AMP is proposed.

Issue:

Resistance of natural rubber to lubricating or fuel oil is poor (P.A. Schweitzer, -

Corrosion Resistance Tables -Metals, Nonmetals, Coatings, Mortars, Plastics, Elastomers and Linings, and Fabrics, Fifth Edition, Marcel Dekker, 2004). The staff does not have sufficient information to determine that there is no aging effect for this material/environment combination.

Request:

Provide additional information (e.g., elastomer material type, fuel oil and lube oil composition) that would demonstrate that the plant-specific applications for the elastomer and fuel or lube oil environment does not have an aging effect requiring management during the extended period of operation or provide an AMP to manage any applicable aging effect.

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION LICENSE RENEWAL APPLICATION Page 9 of 11 Energy Northwest Response:

LRA Table 3.3.2 Diesel Lube Oil As stated in the basis documents for the diesel lubricating oil system, flex connections DLO-FLX-7A1, A2, B1, B2 & C are made of either Buna-N or fluoroelastomer rubber (elastomer) and steel. Industry literature indicates these synthetic elastomers have a good resistance to lubricating oils.

LRA Table 3.3.2 Fire Protection (Fuel Oil and Lube Oil)

As stated in the basis documents for the fire protection system flex connections DO-FLX-1 1 and 36 are constructed of Viton, which is a fluoropolymer elastomer. Industry literature indicates these synthetic elastomers have a good resistance to fuel oils.

As stated in the basis documents for the fire protection system, flex connections for lube oil are constructed of Viton, which is a fluoropolymer elastomer. Industry literature indicates these synthetic elastomers have a good resistance to lube oils.

In addition, as stated in the EPRI Mechanical Tools 1010639, Appendix C, page 2-6, elastomers are defined as rubber or polymers that have properties similar to those of rubber. They are used in nuclear plants in various capacities, such as joint sealants, flexible connections/hoses and moisture barriers. Furthermore, certain elastomers such

-as natural rubbers and ethylene-propylene-diene (EPDM) are not resistant to fuel oil or lubricating oil and only qualified elastomers, that are resistant, are used for oil or fuel oil service. Therefore, based on industry operating experience review, the assumption of proper design and application of the material, and considering that the oil and fuel oil environments do not typically include contaminants or conditions that would result in the degradation of glass (including fiberglass), thermoplastics, and elastomers, aging of these materials is not a concern in lubrication oil and fuel oil environments.

RAI 3.5.2.3-1:

Background:

GALL Report Table III A6, Group 6 Structures (Water-Control Structures), line item II1.A6-12 (TP-7) recommends that AMP XI.S6, ("Structures Monitoring Program"),

should be used to manage the aging effects of loss of sealing/deterioration of seals, gaskets, and moisture barriers (caulking, flashing, and other sealants) for elastomers exposed to various environments. In LRA Table 3.5.2-13, Bulk Commodities, Waterstops, the applicant stated that for elastomer waterstops exposed to air-indoor (within walls, floors, or foundation) there is no aging effect and no proposed AMP.

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION LICENSE RENEWAL APPLICATION Page 10 of 11 Issue:

Continuous waterstops are provided in the reactor building at all horizontal and vertical construction joints in exterior walls and interior walls between and including the top of the foundation mat. High temperature and radiation could exist in the reactor building walls, floors, and foundation mat, and could cause hardening and loss of strength for the elastomer components. The staff does not have sufficient information to determine that there is no aging effect for this material/environment combination.

Request:

Provide justification why hardening and loss of strength due to elastomer degradation for elastomer waterstops exposed to air-indoor (within walls, floors, or foundation) is not considered an aging~effect requiring aging management.

Energy Northwest Response:

EPRI 1015078 "Aging Effects for Structures and Structural Components (Structural Tools)" was used to identify potential aging effects for elastomer components including waterstops. The potential aging effects associated with elastomers as presented in the EPRI Structural Tools are cracking and change in material properties. These aging effects may be the result of thermal exposure or exposure to ionizing radiation and, in the case of rubber, the result of exposure to ultraviolet radiation and ozone. Potential aging effects due to exposure to ultraviolet radiation and ozone are eliminated regardless of material as the subject waterstops are only potentially exposed to an air-indoor environment. Per the EPRI Structural Tools the aging effects as a result of thermal exposure and ionizing radiation are not applicable if the temperature is less than 95 0 F and radiation level is less than 106 rads. Columbia's basis documents show that for the areas of the reactor building where the waterstops were installed the normal room temperature is 70 to 90°F and the maximum radiation dose is 3.6x10 7 rads.

The subject elastomer waterstops are located in the construction joints of the concrete components (walls and floor slabs) in the Emergency Core Cooling Systems (ECCS) pump rooms at the lower elevations of the reactor building. The minimum thickness for any wall for the ECCS pump rooms is 2'-6". This provides a minimum concrete cover of 1'-3" as the waterstops are installed in construction joints along the centerline of each wall. Using industry guidelines for radiation safety and shielding, the concept of half-value layer (HVL) will be used to determine the dose level at this elastomer (waterstop).

Based on this concept of HVL the minimum thickness of 15" of concrete cover at the waterstop locations is sufficient to reduce the ionizing radiation levels to below the threshold of the Structural Tools. Therefore, the actual installed locations of the waterstops will eliminate the potential aging effects associated with thermal exposure or ionizing radiation.

Additionally, as noted above, the waterstops are physically installed within the permanent concrete walls and as such are considered a subcomponent of the wall and implicitly included at the component (wall) level. This means that these waterstops will implicitly be monitored under the Structures Monitoring Program (SMP) when the

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION LICENSE RENEWAL APPLICATION Page 11 of 11 concrete walls are inspected. That is, the waterstops will be considered to be in similar condition of the walls they are installed within when the walls are inspected and monitored under the SMP. Note that the physical location of the waterstops makes them inaccessible for any type of inspection.

Time Limited Aging Analysis (TLAA):

RAI 4.2.1-1:

Background:

Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," states that the core neutron source should be determined by the power distribution, the power level, and the fuel management scheme.

Issue:

LRA Section 4.2.1 states that 54 effective full power year (EFPY) fluenceayalues were extrapolated from 51.6 EFPY fluence values.

'Request:

Please confirm that the flux used for this extrapolation was that assumed for post-Cycle 11, uprated operation. If it is not, please provide the basis for the assumed flux value and justify its use.

Energy Northwest Response:

The flux used for this extrapolation was that assumed for post-Cycle 11 uprated operation.

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION LICENSE RENEWAL APPLICATION Enclosure Page 1 of 2 License Renewal Application Amendment 7 Section no. -Page no. RAI response Section 3.1.2.1.3 3.1-5 RAI B.2.49-1 Section 3.1.2.2.4.1 3.1-7 RAI B.2.49 Section 3.1 .2.2.8.1 3.1-8. RAI B.2.49-1 Table 3.1.1 Item 3.1-21 RAI B.2.49-1.

3.1.1-48 3.1-24 Table 3.1.1 . Item3.11-55PAI B.2.49-1 3.1.1 -55 Table 3.1.2-3 andItems 1 3PAl 3.1.94 B.2.49-1 and 3 Table 3.1.2-3 Items 3.1-95 RAI B.2.49-1 13 and 15 Table 3.1.2-3 Items 3.1-96 RAI B.2.49-1 20 and 22 /

Table 3.1.2-3 Item 47 3.1-99 RAI B.2.49-1 Table 3.1.2-3 Items 3.1-101 RAI B.2.49-1 61, 63, and 68 Table 3.1.2-3 Item 70 3.1-102 RAI B.2.49-1 Table 3.1.2-3 Items 3.1-103 RAI B.2.49-1 79 and 80 Table 3.1.2-3 Items 3.1-104 R B.2.49-1 85 and 87 Table 3.1.2-3 Item 92 3.1-105 RAI B.2.49-1 Table 3.1.2-3 Items 3.1-108 RAl B.2.49-1 122 and 124 Table 3.1.2-3 Items 3.1-109 RAI B.2.49-1 129, 131, and 134 Table 3.1.2-3 Items 3.1-110 RAI B.2.49-1 137, 139, and 144 Appendix A Table of A-4 RAI B.2.49-1 Contents Section A.1.2.11 A-i1 RAI B.2.49-1 Section A.1.2.49 A-24 RAI B.2.49-1 Section A.1.2.49 A-24a RAI B.2.49-1 Section A.1.2.49 A-25 RAI B.2.49-1 Table A-1 Item 26 A-61 RAI B.2.49-1 Table A-1 Item 26 A-61a RAI B.2.49-1 Appendix B Table of B-5 RAI B.2.49-1 Contents

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION LICENSE RENEWAL APPLICATION Enclosure Page 2 of 2 License Renewal Application Amendment 7

-Setion n. *Page no. ' RAI response Table B-1 Number B-12 B.2.49-1 XI.M12 _XI.M12RAI Table B-1 Item B-14 RAI B.2.49-1 XI.M35 I .2 .

Table B-1 Item B-14a RAI B.2.49-1 XI.M35 Table B-2 B-24 RAI B.2.49-1 Section B.2.11 B-57 RAI B.2.49-1 Section B.2.49 B-187 though B- RAI B.2.49-1 191 Section B.2.49 B-187a though B- RAI B.2.49-1 191a Table C-11 C-27 RAI B.2.49-1

Columbia Generating Station ISection 3.1.2.1.3 1License Renewal Application Technical Information

Loss of Pre-load

Reduction of Fracture Toughness

Reduction in Heat Transfer Aging Management Programs The following aging management programs and activities address the aging effects requiring management for the reactor coolant pressure boundary systems:

Bolting Integrity Program

BWR Stress Corrosion Cracking Program

" BWR Water Chemistry Program

Chemistry Program Effectiveness Inspection

" Closed Cooling Water Chemistry Program

External Surfaces Monitoring Program

Flow-Accelerated Corrosion Program

" Heat Exchangers Inspection

Inservice Inspection (ISI) Program Program

" Small Bore Class 1 Piping 4...peetki 3.1.2.2 Further Evaluation of Aging Management as Recommended by NUREG-1801 For the Reactor Vessel, Internals, and Reactor Coolant System, those items requiring further evaluation are addressed in the following sections.

3.1.2.2.1 Cumulative Fatigue Damage Fatigue is a time-limited aging analysis as defined in 10 CFR 54.3. Time-limited analyses are required to be evaluated in accordance with 10 CFR 54.21(c). The evaluation of this time-limited analysis is addressed separately in Section 4.3.

3.1.2.2.2 Loss of Material due to General, Pitting, and Crevice Corrosion 3.1.2.2.2.1 BWR Top Head and Top Head Nozzles, PWR Steam Generator Shell Assembly The BWR Water Chemistry Program mitigates loss of material due to general, pitting, and crevice corrosion. The BWR Water Chemistry Program manages aging effects through periodic monitoring and control of contaminants. The Chemistry Program Effectiveness Inspection will provide a verification of the effectiveness of the BWR Water Chemistry Program to manage loss of material due to general, pitting, and crevice corrosion through examination of components.

Aging Management Review Results Page 3.1-5 ,a,,ae,-240 jAmendment77 I

Columbia Generating Station License Renewal Application Technical Information 3.1.2.2.3.2 Reactor Vessel Beltline Shell, Nozzle, and Welds Reduction of fracture toughness due to radiation embrittlement could occur for reactor vessel beltline region materials exposed to reactor coolant and neutron flux. The effects of embrittlement on the reactor vessel are discussed in Section 4.2. A reactor vessel materials surveillance program monitors radiation embrittlement of the steel reactor vessel beltline materials with stainless steel cladding. The Reactor Vessel Surveillance Program, and the results of its evaluation for license renewal, are presented in Appendix B.

3.1.2.2.4 Cracking due to Stress Corrosion Cracking (SCC) and Intergranular Stress Corrosion Cracking (IGSCC) 3.1.2.2.4.1 BWR Top Head Enclosure Vessel Flange Leak Detection Lines -- Program The reactor vessel flange leak detection nozzle and associated piping at Col bia is a Class 1 line that is normally dry. The stainless steel piping is evaluated f r a reactor coolant environment and is therefore susceptible to cracking due to S,. Cracking of the piping is managed by the Small Bore Class 1 Piping Inspeetien. The nickel-alloy nozzle is evaluated for a reactor coolant environment and is therefore susceptible to cracking due to SCC. Cracking of the nozzle is managed with a combination of the BWR Water Chemistry Program and the Chemistry Program Effectiveness Inspection.

In addition, loss of material for the stainless steel piping that forms the tube-in-a-tube seal cooler for the reactor recirculation pump is managed by the BWR Water Chemistry Program and the Chemistry Program Effectiveness Inspection.

3.1.2.2.4.2 Isolation Condenser Components Cracking of BWR isolation condenser components is not applicable since the Columbia design does not include an isolation condenser.

3.1.2.2.5 Crack Growth due to Cyclic Loading The associated items in Table 3.1.1 are applicable to PWRs only.

3.1.2.2.6 Loss of Fracture Toughness due to Neutron Irradiation Embrittlement and Void Swelling The associated items in Table 3.1.1 are applicable to PWRs only.

3.1.2.2.7 Cracking due to Stress Corrosion Cracking The associated items in Table 3.1.1 are applicable to PWRs only.

Aging Management Review Results Page 3.1-7 ja 2.. G

[Amendment 7

Columbia Generating Station License Renewal Application Technical Information 3.1.2.2.8 Cracking due to Cyclic Loading 3.1.2.2.8.1 Stainless Steel BWR Jet Pump Sensing Lines For Columbia, the jet pump instrumentation lines inside the vessel are not subject to AMR, as they do not perform an intended function. The lines outside of the vessel are part of the reactor coolant pressure boundary and are subject to AMR for a reactor coolant environment. Cracking of the stainless steel lines external to the vessel is managed with a combination of the BWR Water Chemistry Program and the Small Bore Class 1 Piping ..... tie-.

3.1.2.2.8.2 Isolation Condenser Components Cracking of BWR isolation condenser components is not applicable since the Columbia design does not include an isolation condenser.

3.1.2.2.9 Loss of Preload due to Stress Relaxation The associated items in Table 3.1.1 are applicable to PWRs only.

3.1.2.2.10 Loss of Material due to Erosion in Steam Generators The associated items in Table 3.1.1 are applicable to PWRs only.

3.1.2.2.11 Cracking due to Flow-Induced Vibration Cracking due to flow-induced vibration for stainless steel steam dryers exposed to reactor coolant is managed by the BWR Vessel Internals Program.

3.1.2.2.12 Cracking due to Stress Corrosion Cracking and Irradiation-Assisted Stress Corrosion Cracking (IASCC)

The associated items in Table 3.1.1 are applicable to PWRs only.

3.1.2.2.13 Cracking due to Primary Water Stress Corrosion Cracking (PWSCC)

The associated items in Table 3.1.1 are applicable to PWRs only.

3.1.2.2.14 Wall Thinning due to Flow-Accelerated Corrosion in Steam Generators The associated items in Table 3.1.1 are applicable to PWRs only.

3.1.2.2.15 Changes in Dimension due to Void Swelling The associated items in Table 3.1.1 are applicable to PWRs only.

3.1.2.2.16 Cracking due to Stress Corrosion Cracking and Primary Water Stress Corrosion Cracking The associated items in Table 3.1.1 are applicable to PWRs only.

Aging Management Review Results Page 3.1-8 j *vany,20n

[Amendment 7

Columbia Generating Station License Renewal Application Technical Information Table 3.1.1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System Evaluated in Chapter IV of NUREG-1 801 I Further Evaluation Discussion Item Co n C dit Aging Effect/ Aging Management Mechanism Programs Recommended D Number omponenommoiy 3.1.1-48 Steel and stainless steel Class 1 Cracking due to Inservice Inspection No Consistent with NUREG-1801.

piping, fittings and branch stress corrosion (IWB, IWC, and IWD),

connections < NPS 4 exposed to cracking, Water chemistry, and Cracking of piping and in-line reactor coolant intergranular stress One-Time Inspection of components is managed by the corrosion cracking ASME Code Class 1 BWR Water Chemistry Program (for stainless steel Small-bore Piping and the Small Bore Class 1 only), and thermal Piping Sp..ti.. Program and mechanical loading This item is also used for cracking of reactor vessel components (bottom head, closure flange, shell rings, nozzles) managed by the Inservice Inspection (ISI)

Program and the BWR Water Chemistry Program. A Note C is applied.

Aging Management Review Results Page 3.1-21 Amendment 7

Columbia Generating Station License Renewal Application Technical Information Table 3.1.1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System Evaluated in Chapter IVof NUREG-1801 I

Aging Management Further Evaluation Discussion Item Aging Effect/

Mechanism Programs Recommended D Number ComponentlCommodity 3.1.1-54 Copper alloy piping, piping Loss of material due Closed-Cycle Cooling No Not applicable.

components, and piping to pitting, crevice, Water System elements exposed to closed and galvanic The reactor coolant pressure cycle cooling water corrosion boundary does not have any copper alloy components.

3.1.1-55 Cast austenitic stainless steel Loss of fracture Inservice inspection No Consistent with NUREG-1801.

Class 1 pump casings, and valve toughness due to (IWB, IWC, and IWD).

bodies and bonnets exposed to thermal aging Thermal aging Loss of fracture toughness for reactor coolant >250 'C (>482 embrittlement susceptibility screening. Class 1 pump casings and valve OF) is not necessary, bodies is managed by the inservice inspection Inservice Inspection (ISI) requirements are Program.

sufficient for managing these aging effects. Reduction of fracture toughness ASME Code Case N-481 for CASS valve bodies less than also provides an 4 inches is included in this item alternative for pump and managed by the Small Bore casings. Class 1 Piping <_..e*

3.1.1-56 Copper alloy >15% Zn piping, Loss of material due Selective Leaching of No Not applicable.

piping components, and piping to selective leaching Materials elements exposed to closed The reactor coolant pressure cycle cooling water boundary does not include any copper alloy >15% Zn components.

Aging Management Review Results Page 3.1-24 Jai "iwa'2 010 jAm e nd m e nt 7

Columbia Generating Station License Renewal Application Technical Information Table 3.1.2-3 Aging Management Review Results - Reactor Coolant Pressure Boundary Aging Effect NUREG-Row Component Intended Material Environment Requiring Aging Management 1801 Table I Notes No. Type Function(s) Management Program Volume 2 Item Item Annubar Pressure boundary Stainless Steel Reactor (ICoolant Cracking - Flaw Growth Small PipingBore Class 4N/A1 1rspcctiar, N/A H Program 1 (Internal)

E PReactor 2 Annubar Pres Stainless Steel Coolant Cracking- BWR Water Chemistry IV.C1-1 3.1.1-48 A boundary (Internal) SCC/IGA Pressure Reactor Cracking - Small 3 Annubar boury Stainless Steel Coolant PipingBore Class 1 IV.C1-1 3.1.1-48 A boundary (Internal) SCC/IGA Pressure Reactor 4 Annubar Stainless Steel Coolant Loss of Material BWR Water Chemistry IV.C1-14 3.1.1-15 A boundary (Internal)

Pressure Reactor Chemistry Program 5 Annubar Stainless Steel Coolant Loss of Material Effectiveness IV.C1-14 3.1.1-15 A boundary (Internal) Inspection Pressure Air-Indoor 6 Annubar Boury Stainless Steel Uncontrolled None None IV.E-2 3.1.1-86 A Boundary (External)

Pressure Air-Indoor 7 Bolting boundary Stainless Steel Uncontrolled Loss of Pre-load Bolting Integrity IV.C2-8 3.1.1-52 B (External)

Air-ndooPressure Cracking -

8 Bolting boundary Steel Uncontrolled FatigueBolting Integrity IV.A1-6 3.1.1-1 E (External)

Aging Management Review Results Page 3.1-94 1Amendment 7 ...... 20..

Columbia Generating Station License Renewal Application Technical Information Table 3.1.2-3 Aging Management Review Results - Reactor Coolant Pressure Boundary Aging Effect NUREG-Row Component Intended Material Environment Requiring Aging Management 1801 Table I Notes No. Type Function(s) Management Program Item 2 Volume Item Pressure Air-Indoor 9 Bolting Boundary Steel Uncontrolled Cracking - SCC Bolting Integrity N/A N/A G (External)

Pressure Air-Indoor 10 Bolting boundary Steel Uncontrolled Loss of Material Bolting Integrity IV.C1-12 3.1.1-52 B (External)

Pressure Air-Indoor 11 Bolting Steel Uncontrolled Loss of Pre-load Bolting Integrity IV.C1-10 3.1.1-52 B boundary (External)

Condensing PressureReactor Cakn 12 12Condensing Unit Pressure boundary Stainless Steel Coolant Coolntenl Cracking Fatigue' - TLAA IV.C1-15 3.1.1-03 A (Internal) ____ ____

Reactor 13 .Condensing Unit Pressure Stainless Steel Coolant Cracking - Flaw Small Bore Class 1 N/A N/A H Program boundary (Internal) Growth Piping h.speeti-,. N.. N HI Reactor 14 Condensing Pressure Stainless Steel Coolant Cracking - BWR Water Chemistry IV.C1-1 3.1.1-48 A Unit boundary (Internal) SCC/IGA E 15 Condensing Pressure Stainless Steel Coolant Cracking - Small Bore Class 1 IV.C1-1 3.1.1-48 P Unit boundary (Internal) SCC/IGA Piping Inepe _*w.e Program I Reactor __j Condensing Pressure 16 Unit boury Stainless Steel Coolant Loss of Material BWR Water Chemistry IV.C1-14 3.1.1-15 A Unit boundary (Internal)

Condensing Pressure Reactor Chemistry Program 17 Unit boury Stainless Steel Coolant Loss of Material Effectiveness IV.C1-14 3.1.1-15 A

-Unit boundary(Internal) Inspection Aging Management Review Results Page 3.1-95 IAmendment 7

Columbia Generating Station License Renewal Application Technical Information Table 3.1.2-3 Aging Management Review Results - Reactor Coolant Pressure Boundary No. Type Function(s)

Materia Envirnmen RequiAging Effect Management Aging Management ManagmentItem Program NUREG-1801 Volume 2 Table I Item I Notes Condensing Pressure l Air-Indoor 18 Unit boury Stainless Steel Uncontrolled None None IV.E-2 3.1.1-86 A Unit boundary (External)

Flow Elements Pressure Reactor Cracking -

19 < Eles boury Stainless Steel Coolant Fatig- TLAA IV.C1-15 3.1.1-03 A

< 4 inches boundary (Internal) Fatigue Reactor Flow Elements Pressure Cracking - Flaw Small Bore Class 1 20

< 4 inches boundary Stainless Steel Coolant Growth Piping '-Vp _1.n <.

N/A N/A H ProgramI (Internal)

Flow Elements Reactor Pressure Cracking -

21

< 4 inches Stainless Steel Coolant BWR Water Chemistry IV.C1-1 3.1.1-48 A boundary SCC/IGA (Internal)

+/- 4 1- 4 + 4 4 -4 Reactor Flow Elements Pressure Cracking - Small Bore Class 1 22

< 4 inches Stainless Steel Coolant IV.C1-1 3.1.1-48 Ak Program boundary SCC/IGA Piping ' p- etl--

(Internal)

Reactor Flow Elements Pressure 23

< 4 inches Stainless Steel Coolant Loss of Material I BWR Water Chemistry IV.C1-14 3.1.1-15 A boundary (Internal)

Flow Elements Pressure Reactor Chemistry Program 24 < Eles boury Stainless Steel Coolant Loss of Material Effectiveness IV.C1-14 3.1.1-15 A

<4 inches boundary (Internal) Inspection Flow Elements Pressure Air-indoor 25 Stainless Steel Uncontrolled None None IV.E-2 3.1.1-86 A

< 4 inches boundary (External)

Flow Elements Pressure Stainless Steel Reactor Cracking - TLAA IV.C1-15 3.1.1-03 A 26 4 inches boundary (ICoolant Fatigue

___________________________________ (Internal)

Aging Management Review Results Page 3.1-96 A m e :dm::Et A

Columbia Generating Station License Renewal Application Technical Information Table 3.1.2-3 Aging Management Review Results - Reactor Coolant Pressure Boundary Aging Effect NUREG-Row Component Intended Material Environment Requiring Aging Management 1801 Table I Notes No. Type Function(s) Management Program Volume 2 Item Item Heat Closed cycle Reduction in Closed Cooling Water 43 Exchanger Heat transfer Stainless Steel cooling water Redtionan Cled intr VII.C2-3 3.3.1-52 B (tube in a tube) (External) heat transfer Chemistry Heat Closed cycle Reduction in Heat Exchangers 44 Exchanger Heat transfer Stainless Steel cooling water Reat echan VII.C2-3 3.3.1-52 E (tube in a tube) (External)

Heat Reactor 45 Exchanger Heat transfer Stainless Steel Coolant heat transfer BWR Water Chemistry V.D2-13 3.2.1-10 A (tube in a tube) (Internal)

Heat Reactor 46 Exchanger Heat transfer Stainless Steel Coolant Reduction in Heat Exchangers V.D2-13 3.2.1-10 A (tube in a tube) (Internal) heat transfer Inspection Heat Reactor Pressure Cracking - Flaw Small Bore Class 1 47 Exchanger Stainless Steel Coolant N/A N/A H Prgamf boundary Growth Piping 'n,--p-t--n <_

(tube in a tube) (Internal)

Heat Reactor Pressure Cracking -

48 Exchanger Stainless Steel Coolant BWR Water Chemistry IV.AI-1O 3.1.1-19 E boundary SCC/IGA (tube in a tube) (Internal)

Heat Pressure Reactor Cracking - Chemistry Program 49 Exchanger Peure Stainless Steel Coolant Cckig Effectiveness IV.Al-10 3.1.1-19 E (tube in a tube) boundary (Internal) SCC/IGA Inspection Heat Pressure Reactor 50 Exchanger boundary Stainless Steel Coolant Loss of material BWR Water Chemistry VII.A4-2 3.3.1-23 A (tube in a tube) (Internal)

Heat Pressure Reactor Chemistry Program 51 Exchanger Stainless Steel Coolant Loss of material Effectiveness VII.A4-2 3.3.1-23 A (tube in a tube) boundary (Internal) Inspection Aging Management Review Results Page 3.1-99 mendment 7

Columbia Generating Station License Renewal Application Technical Information Table 3.1.2-3 Aging Management Review Results - Reactor Coolant Pressure Boundary Row Component R Tntended I Material Environment Aging Effect Requiring Aging Management NUREG-1801 Table 1 Notes No. Type Function(s) Management Program Volume 2 Item Item Reactor Orifice < 4 Pressure Cracking - Flaw Small Bore Class 1 61 Stainless Steel Coolant N/A N/A H Program inches boundary Growth Piping In ep ee t . <.

(Internal) 4 4- 4 -I- 4 4 4 4 Reactor Orifice < 4 Pressure Cracking -

62 Stainless Steel Coolant BWR Water Chemistry IV.C1-1 3.1.1-48 A inches boundary SCC/IGA (Internal) 4 +/- 4 -I- 4 4 1 1 1 E Reactor Orifice < 4 Pressure Cracking - Small Bore Class 1 63 Stainless Steel Coolant IV.C1-1 3.1.1-48 Program inches boundary SCC/IGA Piping l+9peet. e (Internal) .--.

4 -~ 4 -I. 4 4 4 Reactor Orifice < 4 Pressure 64 Stainless Steel Coolant Loss of Material BWR Water Chemistry IV.C1-14 3.1.1-15 A inches boundary (Internal)

Orifice 65 < 4 Pressure Chemistry Program 65 Stainless Steel Coolant Loss of Material Effectiveness IV.C1-14 3.1.1-15 A inches boundary (Internal) Inspection Orifice < 4 Pressure Air-Indoor 66 Stainless Steel Uncontrolled None None IV.E-2 3.1.1-86 A inches boundary (External)

Orfce<4Reactor Cracking -

67 Orifice <s Throttling Stainless Steel Coolant Cacig- TLAA IV.C1-15 3.1.1-03 A inches (Internal) Fatigue Class 1 N/A N/A H 68 inches< 4 Orifice Throttling Stainless Steel Reactor Coolant Cracking - Flaw Growth PipingBore Small InepetieN Reactor Cracking -

69 Orifice s < Throttling Stainless Steel Coolant BWR Water Chemistry IV.C1-1 3.1.1-48 A inches_______ __(Internal) SCC/IGA Aging Management Review Results Page 3.1-101 jAmendment 7

Columbia Generating Station License Renewal Application Technical Information Table 3.1.2-3 Aging Management Review Results - Reactor Coolant Pressure Boundary Row Component Intended Material Environment TAging Effect Requiring Aging Management NUREG-1801 Table Notes No. Type Function(s) Management Program Volume 2 Item Item Reactor MaaeetIm Orifice < 4 Throttling Stainless Steel Coolant Cracking - Small Bore Class 1 VC 70 inches (Internal) SCC/IGA Piping .i.mepe:c ct*,._I'11 3.1.1-48 A :'ogam Orifie < 4Reactor 71 Oinche Throttling Stainless Steel Coolant Loss of Material BWR Water Chemistry IV.C1-14 3.1.1-15 A (Internal)

Orifice < 4 Reactor Chemistry Program 72 inches Throttling Stainless Steel Coolant Loss of Material Effectiveness IV.C1-14 3.1.1-15 A (Internal) Inspection Orifice < 4 Air-Indoor 73 inches Throttling Stainless Steel Uncontrolled None None IV.E-2 3.1.1-86 A (External) 74 Piping Structural Stainless Steel Treated water Loss of Material BWR Water Chemistry VII.A4-11 3.3.1-24 A integrity (Internal)

Structural Treated water Chemistry Program 75 Piping integri Stainless Steel (Interna Loss of Material Effectiveness VII.A4-1 1 3.3.1-24 A Inspection Structural SAir-Indoor 76 Piping integrity Stainless Steel Uncontrolled None None IV.E-2 3.1.1-86 A (External)

Structural Air-Indoor External Surfaces C 77 Piping integrity Steel Uncontrolled Loss of Material Monitoring VII.-8 3.3.1-58 0106 (Internal)

Aging Management Review Results Page 3.1-102 Amendment 7

Columbia Generating Station License Renewal Application Technical Information Table 3.1.2-3 Aging Management Review Results - Reactor Coolant Pressure Boundary Aging Effect NUREG-Row Component Intended Material Environment Requiring Aging Management 1801 Table 1 Notes No. Type Function(s) Management Program Volume 2 Item Item Piping &

Fittings < 4 Pressure Reactor Cracking -

78 inches (RV boundary Stainless Steel Coolant TLAA IV.C1-15 3.1.1-03 A flange leak off oundary (Internal) Fatigue lines)

Piping &

Fittings < 4 Pressure Reactor Cracking - Flaw Small Bore Class 1 79 inches (RV boury Stainless Steel Coolant N/A N/A H ProgramI flange leak off boundary (Internal) Piping in.p lines)

Piping &

Fittings < 4 Pressure Reactor Cracking - Small Bore Class 1 80 inches (RV b s Stainless Steel Coolant IV.A1C-10 3.1.1-19 E flange leak off boundary (Internal) SCC/IGA Piping lines)

Piping &

Fittings < 4 Pressure Reactor 81 inches (RV bounda Stainless Steel Coolant Loss of Material BWR Water Chemistry IV.C1-14 3.1.1-15 A flange leak off (Internal) lines)

Piping &

Fittings < 4 . Pressure Reactor Chemistry Program 82 inches (RV boundary Stainless Steel Coolant Loss of Material Effectiveness IV.C1-14 3.1.1-15 A flange leak off (Internal) Inspection lines)

Aging Management Review Results Page 3.1-103 jAmendment 7

Columbia Generating Station License Renewal Application Technical Information Table 3.1.2-3 Aging Management Review Results - Reactor Coolant Pressure Boundary NUREG-Row Component Intended Aging-Effect Aging Management 1801 Table 1 Type Function(s) Material Environment Requiring Volume2 Item Notes No. Item Management Piping &

Fittings < 4 Pressure Air-Indoor 83 inches (RV boundary Stainless Steel Uncontrolled None None IV.E-2 3.1.1-86 A flange leak off (External) lines)

Piping & PressurReactor Cracking-84 Fittings < Pressure Stainless Steel Coolant TLAA IV.C1-15 3.1.1-03 A inches boundary (Internal) Fatigue Piping & Pressure Reactor Cracking - Flaw Small Bore Class 1 85 Fittingsinhs< 4 boundary Stainless Steel (Itra)Growth Cracking Coolant Piping 4ir, pectcrfP N/A

.

N/A H rogram inches (Internal)

Piping & Pressure Reactor Cracking 86 Fittings <&4 boury Stainless Steel Coolant Cckig BWR Water Chemistry IV.C1-1 3.1.1-48 A inches (Internal)

Piping & Piig&

PrsueReactor Stainless Steel Pressure Coolant Cracking - Small Bore Class1 3.1.1-48 A 4 5

inches boundary SCC/IGA Piping 4nPpctr,. rogram

__j Piping & Pressure Reactor 88 Fittings < 4 boundary Stainless Steel Coolant Loss of Material BWR Water Chemistry IV.C1-14 3.1.1-15 A inches (Internal)

Piping & Pressure Reactor Chemistry Program 89 Fittings < 4 boundary Stainless Steel Coolant Loss of Material Effectiveness IV.C1-14 3.1.1-15 A inches (Internal) Inspection Piping & Pressure Air-Indoor 90 Fittings < 4 boundary Stainless Steel Uncontrolled None None IV.E-2 3.1.1-86 A inches (External)

Aging Management Review Results Page 3.1-104 ment 7" IAmend

Columbia Generating Station License Renewal Application Technical Information Table 3.1.2-3 Aging Management Review Results - Reactor Coolant Pressure Boundary Aging Effect NUREG-Row Component Intended Material Environment Requiring Aging Management 1801 Table Notes No. Type Function(s) Management Program Volume 2 Item Item Piping & Pressure Reactor Cracking -

91 Fittings < 4 Steel Coolant TLAA IV.C1-15 3.1.1-03 A inches boundary (Internal) Fatigue Piping & Pressure Reactor Cracking - Flaw Small Bore Class 1 92 Fittings < 4 Steel Coolant N/A N/A H Pirogram inches boundary (Internal) Growth Piping ',,pscet *,. L Piping & Pressure Reactor 93 Fittings < 4 boundary Steel Coolant Loss of Material BWR Water Chemistry IV.Al-11 3.1.1-11 C inches (Internal)

Piping & Pressure Reactor Chemistry Program 94 Fittings < 4 Steel Coolant Loss of Material Effectiveness IV.A1-11 3.1.1-11 C inches boundary (Internal) Inspection Piping & Pressure Reactor Loss of Material Flow-Accelerated A 95 Fittings < 4 Steel Coolant - FAC Corrosion IV.C1-7 3.1.1-45 0105 inches boundary (Internal)

Piping & Pressure Air-Indoor External Surfaces 96 Fittings < 4 boundary Uncontrolled Loss of Material MonitoringVII.-8 3.3.1-58 A inches (External)

Piping & Pressure Reactor Cracking -

97 Fittings > 4 boury Stainless Steel Coolant - TLAA IV.C1-15 3.1.1-03 A inches boundary (Internal) Fatigue Piping & Pressure Steel Coolan Reactor 98 Fittings >Ž4 Pesrrcig-Fa

tel Reato Coln Cracking FlawnesInservice Inspection N/A N/A H inches boundary (Internal) Growth Piping & Pressure Reactor Cracking - BWR Stress Corrosion 99 Fittings > 4 boundary Stainless Steel Coolant SCC/IGA Cracking IV.C1-9 3.1.1-41 A inches (Internal)

Aging Management Review Results Page 3.1-105 Amendment 7

Columbia Generating Station License Renewal Application Technical Information Table 3.1.2-3 Aging Management Review Results - Reactor Coolant Pressure Boundary Aging Effect NUREG-Mawterial Enirnend Aging Management 1801 Table 1 Notes No. Type Function(s) aRequiring Program Volume 2 Item Management Item RRC Pump Pressure Closed cycle Chemistry Program 118 Stuffin Stuffing Boxsunary CASS cooling water Loss of material Effectiveness VII.C2-10 3.3.1-50 E Box boundary (Internal) Inspection RRC Pump Pressure Air-Indoor 119 Casing Pressury CASS Uncontrolled None None IV.E-2 3.1.1-86 A Casing boundary (External)

RRC Pump Pressure Air-Indoor External Surfaces 120 RRC Pump Motor Flange boury Steel Uncontrolled Loss of Material Monitoring VII.I-8 3.3.1-58 A boundary (External)

Pressure ~Reactor Cakn Pressure Stainless Steel Coolant Cracking - TLAA IV.C1-15 3.1.1-03 A 121 Tubing boundary (Internal) Fatigue Reactor Pressure Cracking - Flaw Small Bore Class 1 122 Tubing boundary Stainless Steel Coolant N/A N/A H Pr ogram Growth Piping 1nspeelie-(Internal)

Reactor Pressure Stainless Steel Coolant CckigA- BWR Water Chemistry IV.C1-1 3.1.1-48 A 123 Tubing boundary (Internal) 5 Pressure boundary 124 Tg Stainless Sl Steel Reactor Coolant Cracking SCC/IGA- Small PipingBore Class 1 t.11-48 IV.C1-1 I,.C11 3.1.1-48 A" Pr Pr ogram 12 ouT day(Internal)----

big

__j Pressure Reactor 125 Tubing boundary Stainless Steel Coolant Loss of Material BWR Water Chemistry IV.C1-14 3.1.1-15 A (Internal)

Pressure Reactor Chemistry Program 126 Tubing boundary Stainless Steel Coolant Loss of Material Effectiveness IV.C1-14 3.1.1-15 A (Internal) Inspection Aging Management Review Results Page 3.1-108 Amendment 7

Columbia Generating Station License Renewal Application Technical Information Table 3.1.2-3 Aging Management Review Results - Reactor Coolant Pressure Boundary Ro Cmpnet InenedJAging Aging Aging ffectNUREG-Effect Management NUREG Table 1 Row Component Intended Material Environment Requiring 1801 Notes No. Type Function(s) Management Program Volume 2 Item Item Pressure Air-Indoor 127 Tubing boundary Stainless Steel Uncontrolled None None IV.E-2 3.1.1-86 A (External)

Valve Bodies Pressure Reactor Cracking -

128 < 4oies boundary CASS Coolant TLAA IV.C1-15 3.1.1-03 A

< 4 inches boundary (Internal) Fatigue Valve Bodies Pressure Reactor Cracking - Flaw Small Bore Class 1 129 < 4 inches boundary CASS CoolantP Growth.. . N/A N/A H Pr ogram (Internal) Piping .

Reactor 130 Valve Bodies Pressure CASS Coolant Cracking - BWR Water Chemistry IV.C1-1 3.1.1-48 A

< 4 inches boundary (Internal) SCC/IGA Valve Bodies Pressure CASS Reactor Cracking - Small Bore Class 1 IVC11 31148 Al__

< 4 inches boundary (Internal) SCC/IGA Piping ' ..... I^1 Pr6-g-ra rT r__

Valve Bodies Pressure Reactor 132 <4 inches boundary CASS Coolant Loss of Material BWR Water Chemistry IV.C1-14 3.1.1-15 A

< 4 inches boundary (Internal)

Reactor Chemistry Program 133 Valve Bodies Pressure CASS Coolant Loss of Material Effectiveness IV.C1-14 3.1.1-15 A

< 4inches boundary (Internal) Inspection Valve Bodies Pressure Reactor Reduction of Small Bore Class 1 134 < 4 inches boundary CASS Coolant Fracture Piping inTpcctf__ IV.C1-3 3.1.1-55 E Pr Valve Bodies Pressure (Internal)

Air-Indoor Toughness r

135 CASS Uncontrolled None None IV.E-2 3.1.1-86 A

<4 inches boundary (External)

Aging Management Review Results Page 3.1-109 Y

...

dment 7

Columbia Generating Station License Renewal Application Technical Information Table 3.1.2-3 Aging Management Review Results - Reactor Coolant Pressure Boundary Aging Effect NUREG-Row Component Intended Material Environment Requiring Aging Management 1801 Table 1 Notes No. Type Function(s) Management Program Volume 2 Item Item Valve Bodies Pressure Reactor Cracking -

136 < 4oies boury Stainless Steel Coolant Fatigue TLAA IV.C1-15 3.1.1-03 A

< 4 inches boundary (Internal) Fatigue Reactor 137 Valve Bodies

< 4 inches Pressure boundary Stainless Steel Coolant (Internal) Cracking growth- Flaw Small PipingBore eeti.... 1

.. s..Class N/A N/A H Program Reactor 138 Valve Bodies Pressure Stainless Steel Coolant Cracking - BWR Water Chemistry IV.C1-1 3.1.1-48 A

< 4 inches boundary (Internal) SCC/IGA Valve Bodies Pressure Reactor Cracking - Small Bore Class 1 139 < 4Binches boundary Stainless Steel Coolant SCC/IGA Small Bore IV.C1-1 3.1.1-48 A, P "ogramI (Internal) Piping P Valve Bodies Pressure Reactor 140 Stainless Steel Coolant Loss of Material BWR Water Chemistry IV.C1-14 3.1.1-15 A

< 4 inches boundary (Internal)

Valve Bodies Pressure Reactor Chemistry Program 141 Stainless Steel Coolant Loss of Material Effectiveness IV.C1-14 3.1.1-15 A

<4 inches boundary (Internal) Inspection Valve Bodies Pressure Air-Indoor 142 Stainless Steel Uncontrolled None None IV.E-2 3.1.1-86 A

< 4 inches boundary (External)

Valve Bodies Pressure Reactor Cracking -

143 < Bies boury Steel Coolant Fatig- TLAA IV.C1-15 3.1.1-03 A

< 4 inches boundary (Internal) Fatigue Reactor Valve Bodies Pressure Cracking - Flaw Small Bore Class 1 144

< 4 inches boundary Steel Coolant Growth Piping 4ptc, N/A N/A H Program (Internal)

__III_- II -

Aging Management Review Results Page 3.1-110 Amendment 7

Columbia Generating Station License Renewal Application Technical Information APPENDIX A TABLE OF CONTENTS A .1.2.25 Fire Protection Program ............................................................................. 16 A .1.2.26 Fire Water Program .................................................................................. 16 A.1.2.27 Flexible Connection Inspection .................................................................. 17 A.1.2.28 Flow-Accelerated Corrosion (FAC) Program ............................................ 17 A.1.2.29 Fuel Oil Chemistry Program ...................................................................... 17 A.1.2.30 Heat Exchangers Inspection .................................................................... 18 A.1.2.31 High-Voltage Porcelain Insulators Aging Management Program .............. 18 A.1.2.32 Inaccessible Medium-Voltage Cables Not Subject to 10 CFR 50.49 EQ R equirem ents Program ............................................................................. 18 A.1.2.33 Inservice Inspection (ISI) Program ........................................................... 19 A.1.2.34 Inservice Inspection (ISI) Program - IWE ................................................ 19 A.1.2.35 Inservice Inspection (ISI) Program - IWF ................................................ 20 A.1.2.36 Lubricating Oil Analysis Program ............................................................. 20 A.1.2.37 Lubricating Oil Inspection ........................................................................ 21 A.1.2.38 Masonry Wall Inspection .......................................................................... 21 A.1.2.39 Material Handling System Inspection Program ......................................... 21 A.1.2.40 Metal-Enclosed Bus Program .................................................................... 21 A.1.2.41 Monitoring and Collection Systems Inspection ......................................... 22 A.1.2.42 Open-Cycle Cooling Water Program ......................................................... 22 A.1.2.43 Potable Water Monitoring Program ........................................................... 23 A.1.2.44 Preventive Maintenance - RCIC Turbine Casing ..................................... 23 A.1.2.45 Reactor Head Closure Studs Program .................................................... 23 A.1.2.46 Reactor Vessel Surveillance Program ....................................................... 23 A.1.2.47 Selective Leaching Inspection .................................................................. 24 A.1.2.48 Service Air System Inspection .................................................................. 24 IPrograml A. 1.2.49 Small Bore Class 1 Piping ,ieeetie . .................. .............. 24 A.1.2.50 Structures Monitoring Program .................................................................. 25 A.1.2.51 Supplemental Piping/Tank Inspection ...................................................... 25 A.1.2.52 Thermal Aging and Neutron Embrittlement of Cast Austenitic Stainless Steel (C A SS ) Program ............................................................................. 26 A.1.2.53 Water Control Structures Inspection ......................................................... 26 Final Safety Analysis Report Supplement Page A-4 -2O01

.aR...y.

JAmendment 7

Columbia Generating Station License Renewal Application Technical Information coolant water chemistry in accordance with BWRVIP guidelines to ensure the long-term integrity and safe operation of the vessel internal attachment welds.

The BWR Vessel ID Attachment Welds Program credits portions of the BWR Vessel Internals Program and the Inservice Inspection (ISI) Program.

A.1.2.10 BWR Vessel Internals Program The BWR Vessel Internals Program is an existing condition monitoring program that manages cracking due to stress corrosion cracking and irradiation assisted stress corrosion cracking (SCC/IASCC), SCC/IGA, flaw growth, and flow-induced vibration for various components and subcomponents of the reactor vessel internals. The BWR Vessel Internals Program consists of mitigation, inspection, flaw evaluation, and repair in accordance with the guidelines of BWRVIP reports and the requirements of the ASME Boiler and Pressure Vessel Code,Section XI. The BWR Water Chemistry Program monitors and controls reactor coolant water chemistry in accordance with BWRVIP guidelines to ensure the long-term integrity and safe operation of the vessel internal components.

The BWR Vessel Internals Program credits portions of the Inservice Inspection (ISI)

Program.

A.1.2.11 BWR Water Chemistry Program The BWR Water Chemistry Program is an existing program that mitigates degradation of components that are within the scope of license renewal and contain or are exposed to treated water, treated water in the steam phase, reactor coolant, or treated water in a sodium pentaborate solution. The program manages the relevant conditions that could lead to the onset and propagation of a loss of material due to corrosion or erosion, cracking due to SCC, or reduction in heat transfer due to fouling through proper monitoring and control of chemical concentrations consistent with BWRVIP water chemistry guidelines.

The BWR Water Chemistry Program is supplemented by the Chemistry Program Effectiveness Inspection and the Heat Exchangers Inspection, to provide verification of the effectiveness of the program in managing the effects of aging. Additionally, the BWR Water Chemistry Program is supplemented by the BWR Feedwater Nozzle Program, BWR Stress Corrosion Cracking Program, BWR Penetrations Program, BWR Vessel ID Attachment Welds Program, BWR Vessel Internals Program, Inservic Program Inspection (ISI) Program, and Small Bore Class 1 Piping -nsppeetieRn o-provideo verification of the program's effectiveness in managing the effects of aging for reactor pressure vessel, reactor vessel internals, and reactor coolant pressure boundary components.

Final Safety Analysis Report Supplement Page A-1 1 Jaauafy-2O4-9

[Amendment 7 I

Columbia Generating Station License Renewal Application Technical Information Energy Northwest follows the requirements of the BWRVIP ISP and applies the ISP data to Columbia. The NRC has approved the use of the BWRVIP ISP in place of a unique plant program for Columbia.

The provisions of 10 CFR 50 Appendix G require Columbia to operate within the currently licensed pressure-temperature (P-T) limit curves, and to update these curves as necessary. The P-T limit curves, as contained in plant technical specifications, will be updated as necessary through the period of extended operation as part of the Reactor Vessel Surveillance Program. Reactor vessel P-T limits will thus be managed for the period of extended operation.

A.1.2.47 Selective Leaching Inspection The Selective Leaching Inspection detects and characterizes the conditions on internal and external surfaces of subject components exposed to raw water, treated water, fuel oil, soil, and moist air (including condensation) environments. The inspection provides direct evidence through a combination of visual examination and hardness testing, or NRC-approved alternative, as to whether, and to what extent, a loss of material due to selective leaching has occurred.

The Selective Leaching Inspection is a new one-time inspection that will be implemented prior to the period of extended operation. The inspection activities will be conducted within the 10-year period prior to the period of extended operation.

A.1.2.48 Service Air System Inspection The Service Air System Inspection detects and characterizes the material condition of steel piping and valve bodies exposed to an "air (internal)" (i.e., compressed air) environment within the license renewal boundary of the Service Air System. The inspection provides direct evidence as to whether, and to what extent, a loss of material due to corrosion has occurred.

The Service Air System Inspection is a new one-time inspection that will be implemented prior to the period of extended operation. The inspection activities will be conducted within the 10-year period prior to the period of extended operation.

Program A.1.2.49 Small Bore Class 1 Piping Insert A from Page A-24a The Small Borc Class 1 Piping InspcctiOn will dctcct and charactcriZe the conditionS On the inteRna! surfaces of s.mall bore Class 1 piping copoents that -arc ex-posed to reactor coolant. The Small Ber-e Class 1 Piping Insp-e.tion w;..ill provide physie0 evidence as to whether, and to what extcnt, cGraking due to r. to thI-rGmal*

-CC mechaic*Il loading has ocuFrId in smIall b Class 1 pipi*g components. I Gre Small Bore Class 1 Piping nspectio Will a v Y pect that Final Safety Analysis Report Supplement Page A-24 jamwapy-2Q4 IAmendment 7

Columbia Generating Station License Renewal Application Technical Information Insert A to Paqe A-24 The Small Bore Class 1 Piping Program will detect and characterize cracking of small bore Class 1 piping components that are exposed to reactor coolant. This periodic program will provide physical evidence as to whether, and to what extent, cracking due to SCC or to thermal or mechanical loading has occurred in small bore Class 1 piping components. It will also verify, by inspections for cracking, that reduction of fracture toughness due to thermal embrittlement requires no additional aging management for small Class 1 cast austenitic stainless steel valve bodies. The Small Bore Class 1 Piping Program will be a condition monitoring program with no actions to prevent or mitigate aging effects. The program will include visual and volumetric inspection of a representative sample of small bore Class 1 piping, including butt welds and socket welds.

The Small Bore Class 1 Piping Program is a new program that will be implemented prior to the period of extended operation. Inspection activities will start during the fourth 10-year inservice inspection interval and continue through the period of extended operation. The Small Bore Class 1 Piping Program will credit portions of the Inservice Inspection Program. The Small Bore Class 1 Piping Program will verify the effectiveness of the BWR Water Chemistry Program in mitigating cracking of small bore piping and piping components.

Final Safety Analysis Report Supplement Page A-24a Amendment 7

Columbia Generating Station License Renewal Application Technical Information reof fractue toughness due to thermal cmbrittl8eent requlies no addition4al agig anagem~ent for small borFe cast aus6tenitic stainless steel valves.

TChe Small Bore Class 1 Piping Inspection includes vi-su-al -andVolum~etric inspection of a representative sample of small bore Class- 1 pipingcmoet.Tesete provides additional assurance that cr-ackig of rmall boe- Class 1 pipin is not occurring or is inignrificaRt, SUch that an aging ...aagmen. pr is no,,-Rt wa...rranted during th period of extended operation. This one time ORinspetion is appropriate as Columbia ha no~t 8experiened crackin ofIsmall boeF Class 1 piping fromF streSS corrE)oion or thera and m~echanical loading. Should evidence o~f significant aging be reveale~d by the onRe time inspection) or thro)ugh plant operating experience-, periOEdic nspection Will b con.sidered as a plant specifi, an m. aagement program.

The,--Small Bre. C . ClassInspecton berc credits potions of the Inoerrce Inrpection (IS9-) .P..a.. The Small Bore Class 1 Piping InspectionR, is c.ed4ited to verif,' the effectiVeneSs of the BWR ,'atcr Chemi*,stry Program. in mitigatinc,-ra..cing of.-smal*nnre piping and piping componEents-.

The Small Boee Class 1 Pipin,,g ecto is a new one time inspectinr that be impleMeRted pror' t the perid of extended operation. The insprection activitiesq wi4ll be conduc~ted within the porFtion o~f the fourth 10 year IS' inter.'al that occurs priorF to th period of extended operatien..

A.1.2.50 Structures Monitoring Program The Structures Monitoring Program manages age-related degradation of plant structures and structural components within its scope to ensure that each structure or structural component retains the ability to perform its intended function. Aging effects are detected by visual inspection of external surfaces prior to the loss of the structure's or component's intended function. The Structures Monitoring Program encompasses and implements the Water Control Structures Inspection and the Masonry Wall Inspection. This program implements provisions of the Maintenance Rule, 10 CFR 50.65, that relate to structures, masonry walls, and water control structures.

Concrete and masonry walls that perform a fire barrier intended function are also managed by the Fire Protection Program.

The Structures Monitoring Program is an existing program that requires enhancement prior to the period of extended operation.

A.1.2.51 Supplemental Piping/Tank Inspection The Supplemental Piping/Tank Inspection detects and characterizes the material condition of steel, gray cast iron, and stainless steel components exposed to moist air environments. The inspection provides direct evidence as to whether, and to what extent, a loss of material due to corrosion has occurred.

Final Safety Analysis Report Supplement Page A-25 j...a.y

, 20, 1Amendment 7 -

Columbia Generating Station License Renewal Application Technical Information Table A-1 Columbia License Renewal Commitments FSAR Enhancement Commitment Supplement or Item Number Location Implementation (LRA App. A) Schedule 4

eI

49) Small Bore A i no m 11iibore png InSpec I III 11II

i activity.

A.1.2.49 A Within the porFtion ef the fourth 10 Smal:7r Class 1 Piping The Small" or9e Class- 1 Piping InSPectionA Will detecGt and charactcrize IRGPBGt;G44 the- con-ditions On the- inteFrnal surfacos of smal brc Class 1 piping com~ponents, that are eXposed to reac4tor coolant. The Small Bor~e Glass9 1 RPipig In pectionR Will proFvide physical evidence as, to te the peFOE9d ef eateRIed jProgram h-hrQG GWGt03tth eeap i@td faq!nhv epe~atOE)R.

... t ;c"; 0" of.i P a -Va*Cls*,r ,

.G ,G...,;of a repreOsGGentative sample of small boreGsss t p.p..g cG-MGnents. The iRns*pCtio pr;ovide additioeal asur uranc hat+crac~king of small bore Cla-ss 1 piping is not occurrinRg or is

iA;gnif**.Rnt, such that an aging management program is not warranted during the peri o",A"*f extended operationr.

Th~s one-time inspecton, is aplpropriate as, CGlumbia has eot exponencod cra;ckin oa-"f small bore-Class 1 pipin, f stress cerrosIen or thermal and mechanical loading. Shou'd evidence of o;"sigf"i"at ag';ig be revealed by the one-time inspection Er through plat ..... ;. ati.g expe;rence, per;.-d,; ispection Will b8 cnsidered as a elant-specific apine manaaem8Rt roEGram.

i insert A from Page A-61a I [Insert B from Page A-61a Final Safety Analysis Report Supplement Page A-61 IAmendment 7

Columbia Generating Station License Renewal Application Technical Information Insert A to Page A-61 The Small Bore Class 1 Piping Program is a new aging management program.

The Small Bore Class 1 Piping Program will detect and characterize cracking of small bore Class 1 piping components that are exposed to reactor coolant. This periodic program will provide physical evidence as to whether, and to what extent, cracking due to SCC or to thermal or mechanical loading has occurred in small bore Class 1 piping components. It will also verify, by inspections for cracking, that reduction of fracture toughness due to thermal embrittlement requires no additional aging management for small Class 1 cast austenitic stainless steel valve bodies. The Small Bore Class 1 Piping Program will be a condition monitoring program with no actions to prevent or mitigate aging effects. The program will include visual and volumetric inspection of a representative sample of small bore Class 1 piping, including butt welds and socket welds.

Insert B to Page A-61 Implemented prior to the period of extended operation. Inspection activities will start during the fourth 10-year inservice inspection interval, then ongoing.

Final Safety Analysis Report Supplement Page A-61 a Amendment 7

Columbia Generating Station License Renewal Application Technical Information APPENDIX B TABLE OF CONTENTS B.2.48 Service Air System Inspection .......................................................................... 183 4Program-"

B.2.49 Small Bore Class 1 Piping peGt ......................................................... 187 B.2.50 Structures Monitoring Program .......................................................................... 192 B.2.51 Supplemental Piping/Tank Inspection ............................................................... 197 B.2.52 Thermal Aging and Neutron Embrittlement of Cast Austenitic Stainless Steel (C AS S ) P ro g ra m ............................................................................................... 20 1 B.2.53 Water Control Structures Inspection .................................................................. 206 Aging Management Programs Page B-5

[Amendment 7

Columbia Generating Station License Renewal Application Technical Information Table B-1 Correlation of NUREG-1801 and Columbia Aging Management Programs (continued)

Number NUREG-1801 Program _ Corresponding Columbia AMP XI.M10 Boric Acid Corrosion Not Applicable. Columbia is a BWR and does not use boric acid in any systems. The Standby Liquid Control System uses a sodium pentaborate solution (a mixture of boric acid and borax) that is not aggressive to metals.

XI.M1 1A Nickel-Alloy Penetration Not Applicable. This program is applicable to PWR Nozzles Welded to the plants, Columbia is a BWR.

Upper Reactor Vessel Closure Heads of Pressurized Water Reactors XI.M12 Thermal Aging Not credited for aging management. The Inservice Embrittlement of Cast Inspection (ISI) Program (See Section B.2 33 Program Austenitic Stainless Steel the Small Bore Class 1 Piping i"Pe(tie See (CASS) Section B.2.49) is credited for pump casings and valve bodies.

XI.M13 Thermal Aging and Neutron Thermal Aging and Neutron Embrittlement of Cast Irradiation Embrittlement of Austenitic Stainless Steel (CASS) Program Cast Austenitic Stainless See Section B.2.52.

Steel (CASS)

XI.M14 Loose Parts Monitoring Not credited for aging management. The Columbia loose parts detection system has been deactivated and spared in-place, as described in FSAR Section 7.7.1.12.

XI.M15 Neutron Noise Monitoring Not Applicable. This program is applicable to PWR plants, Columbia is a BWR.

XI.M16 PWR Vessel Internals Not Applicable. This program is applicable to PWR plants, Columbia is a BWR.

XI.M1 7 Flow-Accelerated Corrosion Flow-Accelerated Corrosion (FAC) Program See Section B.2.28.

XI.M18 Bolting Integrity Bolting Integrity Program See Section B.2.4.

XI.M19 Steam Generator Tube Not Applicable. Columbia is a BWR design that Integrity does not utilize steam generators.

Aging Management Programs Page B-12 dFy-~4~

Amendment 7

Columbia Generating Station License Renewal Application Technical Information Table B-1 Correlation of NUREG-1801 and Columbia Aging Management Programs (continued)

Number NUREG-1801 Program Corresponding Columbia AMP XI.M31 Reactor Vessel Surveillance Reactor Vessel Surveillance Program See Section B.2.46.

XI.M32 One-Time Inspection Chemistry Program Effectiveness Inspection See Section B.2.12.

Cooling Units Inspection See Section B.2.14.

Diesel-Driven Fire Pumps Inspection See Section B.2.18.

Diesel Starting Air Inspection See Section B.2.16.

Diesel Systems Inspection See Section B.2.17.

Flexible Connection Inspection See Section B.2.27.

Heat Exchangers Inspection See Section B.2.30.

Lubricating Oil Inspection See Section B.2.37.

Monitoring and Collection Systems Inspection See Section B.2.41.

Service Air System Inspection See Section B.2.48.

Supplemental Piping/Tank Inspection See Section B.2.51.

Selective Leaching of Selective Leaching Inspection Materials See Section B.2.47.

Buried Piping and Tanks Buried Piping and Tanks Inspection Program Inspection See Section B.2.5.

One-time Inspection of 9,aI,-,BeF8 Clas. 1 Piping . .SPcctiOn ASME Code Class 1 Small- See Section B.2.49.

Bore Piping External Surfaces Monitoring External Surfaces Monitoring Program See Section B.2.23.

Ilnsert A from IPage B-1 4a Aging Management Programs Page B-14 Jarauy 2010 jAmendment 7

Columbia Generating Station License Renewal Application Technical Information Insert A to Page B-14 Not credited for aging management. The new, plant-specific Small Bore Class 1 Piping Program is credited for aging management.

Aging Management Programs Page B-14a Amendment 7

Columbia Generating Station License Renewal Application Technical Information Table B-2 Consistency of Columbia Aging Management Programs with NUREG-1801 (continued):

Consistent New Consistent with Plant- Enhancement Program Name Existing with NUREG- NUREG- Specific Required 1801 1801 with Exceptions Preventive Maintenance -

RCIC Turbine Existing Yes --

Casing Section B.2.44 Reactor Head Closure Studs Program Existing Yes Section B.2.45 Reactor Vessel Surveillance Existing Yes Program Section B.2.46 Selective Leaching Inspection New Yes Section B.2.47 Service Air System Program Inspection New Yes Section B.2.48 Small Bore Class 1 Piping IiseetOG Nw e Section B.2.49 Structures Monitoring Program Existing Yes Yes Section B.2.50 Supplemental Piping/Tank New Yes Inspection Section B.2.51 Aging Management Programs Page B-24 1-P., ._.-MY 2OW 19 jAmendment 7

Columbia Generating Station License Renewal Application Technical Information B.2.11 BWR Water Chemistry Program Program Description The BWR Water Chemistry Program will mitigate damage related to loss of material due to corrosion or erosion, cracking due to SCC, and reduction of heat transfer due to fouling of plant components that are within the scope of license renewal and contain or are exposed to treated water, treated water in the steam phase, reactor coolant, or treated water in a sodium pentaborate solution. The program manages the relevant conditions (e.g., concentrations of chlorides, oxygen, and sulfates) that could lead to the onset and propagation of a loss of material, cracking, or reduction of heat transfer through proper monitoring and control consistent with the current EPRI water chemistry guidelines. The relevant conditions are specific parameters such as sulfates, halogens, dissolved oxygen, and conductivity that could lead to, or are indicative of, conditions for corrosion or SCC of susceptible materials, as well as erosion and fouling. The BWR Water Chemistry Program is a mitigation program.

The BWR Water Chemistry Program is supplemented by separate one-time inspections of representative areas of treated water systems. One inspection is the Chemistry Program Effectiveness Inspection. This one-time inspection provides further confirmation that loss of material and cracking are effectively mitigated, or to detect and characterize whether, and to what extent, degradation is occurring. The other inspection is the Heat Exchangers Inspection. This one-time inspection provides further confirmation that reduction in heat transfer is effectively mitigated, or to detect and characterize whether, and to what extent, degradation is occurring.

Additionally, the BWR Water Chemistry Program is supplemented by the BWR Feedwater Nozzle Program, BWR Stress Corrosion Cracking Program, BWR IPrograml Penetrations Program, BWR Vessel ID Attachment Welds Program, BWR Vessel Internals Program, Inservice Inspection (ISI) Program, and Small Bore Class 1 Piping

,.,peG.k".,-to provide verification of the program's effectiveness in managing the effects of aging for reactor pressure vessel, reactor vessel internals, and reactor coolant pressure boundary components.

NUREG-1801 Consistency The BWR Water Chemistry Program is an existing Columbia program that is consistent with the 10 elements of an effective aging management program as described in NUREG-1801,Section XI.M2, "Water Chemistry."

Exceptions to NUREG-1801 None.

Aging Management Programs Page B-57 Ja*'-Ry 20!9

[Amendment 7

Columbia Generating Station License Renewal Application Technical Information B.2.49 Small Bore Class 1 Piping hw*xr~ram Program Description Insert A from Page B-1 87a The Small Bore Class 1 Piping In,,spection vill detec-t,-- cha.FaGteFiz*

-* the c-nIditi.R oR the inter*al s*ufac. of -m-all bore Class 1 piping com.po.ets that are exposc reactor coolIant. The one time8 inSPection Will provide physical evOidence ast7hehr and to what extont, cracking duo to SOC orF to therm~al Or mecGhanical loadfingha occrrG1Fed in small bore Class 1 piping compoenRts. it will also verify, by inspection fr cracking, that reduction) of fracture toughness due to thermnal embrittlement requires no additionRal aging management for small Class 1 GASS valve bodies. The Sm:all Bor Class 1 Paipig Inspection will be an evaluation and inspection with no actions to preven eFFitgae g-geffeets.

This one time ins;pection is applicsable to smFall bore ASMVE Code Class 1 piping components less than 4 inches nomin~al pipe size (NPS 4), which includes piping-,

fittings, branch connections;,-and valve bodies. The Smnall BorFe Class 1 Pipin Insp~ection includes visual and volumetricisecto of rpresentative sample o~f ml bo Glass 1 pipin componnts. The inpcinprovides additionLE)al aurnoth-at eihrage related degrada-tfion of small bore ,A.SME Code Class 1 pipin cmoents is not ocurigo that the aging is insignific;,ant,, such that an additional ain mnanagemn progam is not warranted duFrig the period of extended operationR.

Columbia has not exeine cracking of smFall bore Class 1 piping solely due to stres cor Erosoor thermnalI and Mec-hanical loading, and therefore this one time ORSPee~tOEien appFOP~ate, The ORinspetion Will finclude a representativ~e sample of the sm~all bore Class 1 pipn population, and, where practical, Will focGus On the bounding or lead comfpoinients;ms susceptible to aging due to tim in severity of operating con)ditionRs, and loes Ge~e, desi-manwinP. The guidelines of EPRI Report 1000701, "Interim Thermfal Fatigue ManagemRent Guideline (MRP 24)" will be considlered in s~electing the sample size-- and, loaton.Actual inspoctionR locations will be based On physical accessibility, -,-e--pouee leels, NDE= techniques, and locGations identified in NRC IenfrationNoic (INj) 97 46.

Vel'Ut:A exaffllneties Ei~neluding qualified deswE~eti*'e and/eF nondesturuciv techniques;) will be porformed by qualified personnel following procedures thatar consistent with Section AI of the ASMVE Code and 10 CFR 50, ,Appendix;. B.

Unacceptable insEpection findin~gs will be evaluated by the Columbia correctve actionR process USing crteiai accrdance-with the, ASME Code-.4 The evaluation o~f indicativOns will include determinling the extent of conRdfitionR byteepaso ^f the samnple s-.-L.Iz when called for by the Code. Evluation of inspection results mnay lead to the ceto of a plant specific MPOr ma" confiFrm that age related degradation is either no~t ocrrig Or ISinsigniiat Aging Management Programs Page B-187 j~ay21 jAmendmnent

Columbia Generating Station License Renewal Application Technical Information Insert A to Paqe B-187 The Small Bore Class 1 Piping Program will detect and characterize cracking of small bore, less that 4 inches nominal pipe size, Class 1 piping components (piping, fittings, branch connections, and valve bodies) that are exposed to reactor coolant. This periodic program will provide physical evidence as to whether, and to what extent, cracking due to SCC or to thermal or mechanical loading has occurred in small bore Class 1 piping components. It will also verify, by inspections for cracking, that reduction of fracture toughness due to thermal embrittlement requires no additional aging management for small bore Class 1 cast austenitic stainless steel valve bodies. The Small Bore Class 1 Piping Program will be a condition monitoring program with no actions to prevent or mitigate aging effects.

While the ASME Code does not require volumetric examination of Class 1 small bore piping, the Small Bore Class 1 Piping Program includes visual and volumetric inspection of a representative sample of small bore Class 1 piping components; the sample will include butt welds and socket welds, and will focus on the bounding or lead components most susceptible to aging due to time in service, severity of operating conditions, and lowest design margin. Actual inspection locations will be based on physical accessibility, exposure levels, NDE techniques, and locations identified in NRC Information Notice 97-46. Volumetric examinations (including destructive and/or nondestructive techniques) will be performed by qualified personnel following procedures that are consistent with Section Xl of the ASME Code and 10 CFR 50, Appendix B.

In scope components will be grouped into populations based on component type, material and environment. Sample size will be 10% of each population (except socket welds) with a minimum of one location and a maximum of twenty locations; the socket weld sample will include three locations. 100% of each sample will be inspected each 10-year ISI interval, with the breakdown of inspections between outages within the interval per ASME Section Xl, Subsection IWB, Program B.

If a qualified non-destructive volumetric examination technique does not become available for socket welds, destructive examination will be conducted. Opportunistic destructive examination will be performed when socket welds are removed from service for other considerations, such as plant modifications. If socket welds do not become available on opportunistic bases prior to the scheduled inspections within the 10-year interval, then socket welds will be selected for planned destructive examinations.

Unacceptable inspection findings will be evaluated by the Columbia corrective action process. The evaluation of indications will include determining the extent of condition by the expansion of the sample size.

Aging Management Programs Page B-187a Amendment 7

Columbia Generating Station License Renewal Application Technical Information The Sm"all BreP-lass 1 Piping In.pection is a new one time inspection that will be implemented prior to the period of extended operation. The inspectionR acti'.,ities Will be conducted during the portion Of the fourth 10 year IS. innteral -that is prior to the period of extended operation.

The Small Bore Class 1 Piping InSpectionR W.ill credit portionsG of the inscr.,icc Inspcci (ISI) Program. The Small ! re Class Piping In I .... verify the effectiVene.. of the BW,1R V.fatcr Chcmistry Program in mitigating cracking of small boepig and NUREG-1801 Consistency Insert A from Page B-188a T-he Small Bore Class 1 Piping nsrpect-io-n i a new oe time in-specton for Columbia that wfi"l be consistent With the 10 elements o~f an effcctive aging management programn as described in NINUREG 1801, Sectien XI.M35, "One time InsPectGi of AS*,SE Code Class 1 Small Bore Piping."

NeR8.

Aging Management Program Elements The results of an evaluation of each program element are provided below.

" Scope of Program Sp f mlnsert B from Page B-188a The Small Bou Class 1 Piping Ispection is a One time inspection of a sample o ASIVE Code Class I piping and pIp'Ing components less than NPS 4. The inspection willincudemeasures to verify that unacceptable degradation is nOt occrigW i Class 1 small boeF piping and pipin copnents (valv.e bodies), ther~eby, con1firming operation. See MoA orn a-d Trending below for a discussion of sample selectionR Ilnsert C from Page B-1 8a

" Preventive Actions e__

The Small Bore Class 1 Piping Inspection wA.ill be an evaluation and inspectionR with no actions to prevent or mitigate aging effectS-.

" Parameters Monitored or Inspected InetDfo aeB18 The Small Bore Class 1 Piping !nspectionR is a ene time inspectionR that will include volumetoric examfinations (destructive or nondestructive) peifenmed by qualified personnel, using qu-alificd volumctrici examination techniques, an~d fE)E)Wi-Rg procedures consistent With SectionR Xl of the ,ASMVE Code and 10 CFIR 50, Appendix Aging Management Programs Page B-188 Jt~ry~4 jAmendment 7

Columbia Generating Station License Renewal Application Technical Information Insert A to Page B-188 The Small Bore Class 1 Piping Program is a new Columbia program that is plant-specific. There is no corresponding aging management program described in NUREG-1801. Unlike the aging management program described in NUREG-1801,Section XI.M35, "One-time Inspection of ASME Code Class 1 Small-Bore Piping," the Columbia Program is a periodic aging management program rather than a one-time inspection.

Therefore, the program elements are compared to the elements listed in Table A.1-1 of NUREG-1800.

Insert B to Page B-188 The Small Bore Class 1 Piping Program is a periodic inspection of a sample of ASME Code Class 1 piping and piping components (valve bodies) less than NPS 4 for cracking.

Insert C to Page B-188 The Small Bore Class 1 Piping Program will be a condition monitoring program with no actions to prevent or mitigate aging effects.

Insert D to Page B-188 The Small Bore Class 1 Piping Program is a periodic program that will include visual and volumetric examinations (destructive or nondestructive) for cracking. Examinations will be performed by qualified personnel, using qualified examination techniques and following procedures consistent with Section Xl of the ASME Code and 10CFR50, Appendix B.

Aging Management Programs Page B-188a Amendment 7

Columbia Generating Station License Renewal Application Technical Information

Detection of Aging Effects Insert A from Page B-1 89ai This inSpection Will Pe ARM. Volu1-mctric examin~ations on selecated weld locations;.

Columbia has not experienced crackiRg of small bore Class 1 ppinRg dlue to stres corrosion or thermal and mechanc*al loading, and therefore this on*,e time M in.pectio.R is,-apIpFeFate.- Columbia has found cracking due to fatigue and groe.Ath of construction flaws of sm all boe piing See Operating Expgerienca below-for dfiscGussion of site operating expericence to date and lack of stress coroioFo thermal and mechanical loading induced cracks.

Monitoring and Trending Insert B from Page B-189aJ The inspection Will i a reRtatiye sample of the small bo*r Class 1 *p*;in population, and where pF;;rac tical, Will focrus on the boGunding orp lead cmoet moEst susceptible to aging due to time in sevie se.eriy of operating conditions, and lo'xest desigR nmargin. The guidelin-es f E=PRI Repor1 1000701 "Interim Thermal Fatigue Management GUideline (,RP 24)" will be co.i-dered ,i selecti*g the sample size and locations. Actual inspection locations wNill be based on physic accessibility, exposure levels, NDE= techniques, an locGations identified in NRC Information Notice 97 46. Volumetrc- examinations (including qualified destruct4-ve and nondestructive techniques) will be performed by qualified personel fO'l.OWi.g proedWrFe that are cons;itent with Sectio.n Xl *fthe A.SME Code and 10 CFR 50, Unacceptable inspection findiRgs* will be evaluated by the Columbia corretive actioR process. The Small BrIe Class 1 PipOng Ilspecti 3eed size in response to unacceptable inspection findings-. Evyaluation ofiseto results may lead to the creation of a plant-specific aging mnanagement.. prgamo mnay conRfirmF that age related degradation iSeither o curn ri isgnifficant.

" Acceptance Criteria Insert C from Page B-189a]

Unaccepta inspection findings Will be evaluated by the Columbia orrective action procGess usn rtra inccodance with the ASMVE Coe. The evaluation of indications Will incalude dete~lrminig the extent o~f condition by the epanson) of the sample size when called for by the Code. Evaluation of isetorsults may lead to the craeation of a plant specific aging management prora ormay confirm~ that age related degradation) is either no~t occurringo sisgiiat

" Corrective Actions This element is common to Columbia programs and activities that are credited with aging management during the period of extended operation and is discussed in Section B.1.3.

Aging Management Programs Page B-1 89 ,a, y244

[Amendment 7

Columbia Generating Station License Renewal Application Technical Information Insert A to Page B-189 The Small Bore Class 1 Piping Program will perform visual and volumetric inspection of a representative sample of small bore Class 1 piping components, including butt welds and socket welds, and will focus on the bounding or lead components most susceptible to aging due to time in service, severity of operating conditions, and lowest design margin. Actual inspection locations will be based on physical accessibility, exposure levels, NDE techniques, and locations identified in NRC Information Notice 97-46.

In scope components will be grouped into populations based on component type, material and environment. Sample size will be 10% of each population (except socket welds) with a minimum of one location and a maximum of twenty locations; the socket weld sample will include three locations. 100% of each sample will be inspected each 10-year ISI interval, with a breakdown of inspections between outages within the interval per ASME Section XI, Subsection IWB, Program B. Inspections will start during the fourth 10-year interval, which begins prior to the period of extended operation.

Volumetric examinations (including destructive and nondestructive techniques) will be performed by qualified personnel following procedures that are consistent with Section XI of the ASME Code and 10 CFR 50, Appendix B. If a qualified non-destructive volumetric examination technique does not become available for socket welds, destructive examination will be conducted. Opportunistic destructive examination will be performed when socket welds are removed from service for other considerations, such as plant modifications. If socket welds do not become available on opportunistic bases prior to the scheduled inspections within the 10-year interval, then socket welds will be selected for planned destructive examinations.

Insert B to Page B-189 Unacceptable program findings will be evaluated, tracked, and trended by the Columbia corrective action program. Extent of condition will be determined by the expansion of the sample size as called for by the ASME Code,Section XI, Subsection IWB-2400.

Insert C to Page B-189 Acceptance criteria will be in accordance with the ASME Code,Section XI, Subsection IWB-3100. The evaluation of indications will include determining the extent of condition by the expansion of the sample size as called for by the ASME Code, Section Xl, Subsection IWB-2400.

Page B-i 89a Amendment 7 Management Programs Aging Management Programs Page B-1 89a Amendment 7

Columbia Generating Station License Renewal Application Technical Information

Confirmation Process This element is common to Columbia programs and activities that are credited with aging management during the period of extended operation and is discussed in Section B.1.3.

Administrative Controls This element is common to Columbia programs and activities that are credited with aging management during the period of extended operation and is discussed in Section B.1.3.

" Operating Experience Insert A from Page B-190a]

Based On Feview of plant pecific and i.dustr' operatin*gGexperie , theidenti aging effectS require mnanagemenRt forF the period of extene oeationR. The Small BorFe Glass 1 Piping InRspection proevides conRfiFrmationR of mnaterial conditions near the the lncerYice Insp~ection (1SI) ProgrFam, and control) of wVater chemistry, via the BWVR Water: Qhemistr; Proqfgram, provide adequate managemen~

lndUSt~,' operating experience:-

UREG 1801 is based on industry operating exein e)trugh January 2005.

Recent industr,' operating experience has bee reieedFo appli ability. Futur operating exprec is captured through the normnal operating experiec reiw proGess, whýichwl con9111 GARtinue thro)ugh the period of extended operatin Industry operating exprec 9ilbe considered when implementing this oee time inpcto.Pln perating experience forF this activity will be gained as it ims implemented near the period of extended operatioen, and will be factored into the activity. As such, operating eXPerience assures that imlmntto f the Sm~all Bore Class 1 Piping Inspection Will conRfiFrm material condition relative toth effects of aging such that applic~able componRents Will con)tinue to pel~orlmthi intended functionRs consistent With the current liEensing basis for the period of exteded operation.

Columbia o~perating ex(perience eview of Columfbia operatingeprmc identified other piping befinfg FA exa.mined by the same techniques and small bore piping that has expercienced The Small Bore Class 1 Pipig Ins6pection is a new one time inspection activity forF which plant o~perating experience has shown onRly one occurrcence Of stres corrosincacking, and that as; one of several contributorcs to fatigue cracking.

The evaluatins and exami~nations to be pe~fFRmed by this activity-w'illuse, qualified ve~m~etke examinat!e teehniqiies e+ deStF~etive exafflietie Aging Management Programs Page B-190 Ja.*ary 20!0

[Amendment 7 -/

Columbia Generating Station License Renewal Application Technical Information Insert A to Page B-190 The site corrective action program and an ongoing review of industry and plant-specific operating experience, continued through the period of extended operation, will be used to ensure that the Small Bore Class 1 Piping Program identifies cracking before loss of function of the small bore piping.

Industry operating experience:

Industry operating experience will be considered when implementing the Small Bore Class 1 Piping Program. The evaluations and examinations to be performed by this program will use qualified volumetric examination techniques or destructive examination techniques with demonstrated capability and a proven industry record to detect cracking in piping weld and base metal. As such, industry operating experience assures that implementation of the Small Bore Class 1 Piping Program will identify cracking such that applicable components will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.

Columbia operating experience:

A review of Columbia operating experience identified other piping being examined by the same techniques and small bore piping that has experienced cracking due to fatigue. Plant operating experience for this program will be gained as it is implemented near the period of extended operation, and will be factored into the program.

Several cracks due to vibration induced fatigue or construction flaws occurred in small bore piping during.the early years of plant life. Design changes were instituted to reduce vibration and sources of cyclic loading. The occurrence of these small bore leaks has decreased in recent years showing the effectiveness of the actions being taken. No instances of stress corrosion cracking or low cycle fatigue cracking as the sole failure mechanism were identified. A single instance of small bore Class 1 piping failure related to stress corrosion cracking was found in 1993, which also involved other contributing factors that led to fatigue cracking. The weld was removed and configuration was changed to address the vibration and cyclic loading considerations.

No other instances of stress corrosion cracking of small bore Class 1 piping have been identified.

Page B-i 90a Amendment 7 Aging Management Programs Aging Programs Page B-1 90a Amendment 7

Columbia Generating Station License Renewal Application Technical Information techniques with demonstrated capability and" a pro-en indu'str,' record to detect cracking in 'piping w.eld and base m~etal.

Several crackS due to vibration induced fatigue or construction flaws occurred i small borFe piping dluFrig the early years of plant life-. Design changes-- weree instituted to reduc~e vibration and sources Of cyclic loading. The ocurec f theso small bore leaks has decreased in recent years s~hoWing the effee-t~i1.'-enes of the actions being taken. No instances of stress corro~sion cracking o-Fe1W-eydee faiu ~G(R stesl failure mechanismR were ietfd.A single instance of small bore Class 1 piping failure related to stress corrosionR cracking was foun inR 1993, which also involved othcr contributing factors that led to fatigue cracking-.

The weld was removed and configuration was changed to address the vibration and c~yc'ic; loading conSiderations Ne-other intn esf StreGs corrsio The Small Bore Class 1 Piping Inspection Will be developed based on relevant plant.

and indus~tr' operatin epFie ee.. The site corrective action program and a ongong rview of industr,' operating experience will be used to ensr-'e that the oee timne inspection cERoFnfrs mnaterial condition) such that the existing prga (S)i demonGstrated to be effective in m~anaging the identified agingefcso e aging mnanagement pro~gram will b-Ae- de-velopd.

Required Enhancements

lprogram Not applicable, this is a new aetiv4ty.

Conclusion Insert A from Page B-191a The Smnall Bore Class 1 RPipig Inspection Wil Vef ~king due to stres rhac corrosionR and mnechanical loading, and cracking due to reduction 3f fracturc toughnes are not occurrfing or are insignificant, such that an aging management pro~gram isno require during the period of extended operationý. The Small Bore Class 41 Piping Inspection will provide reasonable assurance that the aging effects are no~t occurring such that comFponents subject to aging management review. will- Geeti ie-te Pe~fE)Fm their intended functions GE)nsistent with the current lcnigbasGis for thepeido extended operateion.

Aging Management Programs Page B-191 jiauapýy [Amendment 7

Columbia Generating Station License Renewal Application Technical Information Insert A to Paqe B-191 The Small Bore Class 1 Piping Program will monitor the extent of cracking in small bore piping through the period of extended operation. The Small Bore Class 1 Piping Program will provide reasonable assurance that cracking is identified and repaired such that components subject to aging management review will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.

Programs B-191a Page B-191 a Amendment 7 Management Programs Aging Management Page Amendment 7

Columbia Generating Station License Renewal Application Technical Information Table C-11 BWRVIP-74-A BWR Reactor Pressure Vessel Inspection and Flaw Evaluation Guidelines for License Renewal Applicant Action Item Text Plant-Specific Response (3) 10 CFR 54.22 requires that each LR No technical specification changes are application include any technical specification required for the inspection strategy changes (and the justification for the changes) described in the BWRVIP-74-A report.

or additions necessary to manage the effects of aging during the period of extended operation Technical specification changes due to as part of the LR application. In its Appendix A embrittlement, i.e. Pressure-Temperature to the BWRVIP-74 report, the BWRVIP stated Limits, will be submitted prior to the that the technical specification changes expiration of the currently approved limits, resulting from neutron embrittlement will be as discussed in LRA Section 4.2.

made at the appropriate time prior to the end of the current license. Those LR applicants referencing the BWRVIP-74 report for the RPV components shall ensure that the inspection strategy described in the BWRVIP-74 report does not conflict or result in any changes to their technical specifications. If technical specification changes do result, then the applicant should ensure that those changes are included in its LR application.

(4) The staff is concerned that leakage around The reactor vessel flange leak detection the reactor vessel seal rings could accumulate (VFLD) lines are in the scope of license in the VFLD lines, cause an increase in the renewal. See the scoping and screening concentration of contaminants and cause results in the LRA for the Reactor Coolant cracking in the VELD line. The BWRVIP-74 System Pressure Boundary (piping and report does not identify this component as fittings, flange leak detection lines, Section within the scope of the report. However, since 2.3.1.3 and Table 3.1.2-3). Refer to the VFLD line is attached to the RPV and Section 3.1.2.2.4 of the LRA for further provides a pressure boundary function, LR information, and also see item 3.1.1-19 in applicants should identify an AMP for the VFLD LRA Table 3.1.1.

line.

Cracking of these lines is managed by thh -ProgramI Small Bore Class 1 Piping .... eetie.*.f4his aging management program is described in Appendix B of the LRA.

Response to BWRVIP Applicant Action Items Page C-27 e..Ua.y 2o010 Amendmentt7

GO2-10-135, Response to Request for Additional Information, License Renewal Application

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GO2-10-135, Response to Request for Additional Information, License Renewal Application

Text

Subject:

References:

Dear Sir or Madam:

Attachment:

Enclosure:

Background:

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