ML20195E648
| ML20195E648 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 11/16/1998 |
| From: | Cruse C BALTIMORE GAS & ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9811190057 | |
| Download: ML20195E648 (13) | |
Text
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Cumes II. C;:tUSE 13altimore Gas and Electric Company Vice President Calvert Cliffs Nuclear Power Plant Nuclear Energy 1650 Calven Cliffs Parkway
)
Lusby, Maryland 20657 410 495 4455 November 16,1998 U. S. Nuclear Regulatory Commission Washington, DC 20555 A'ITENTION:
Document Control Desk
SUBJECT:
Calvert Cliffs Nuclear Power Plant Unit Nos.1 & 2; Docket Nos. 50-317 & 50-318 Response to Request for Additional Information for ^c Review of the Calvert Cliffs Nuclear Power Plant, Units 1 & 2, Integrated Plant Assessment Report for the Feedwater System
REFERENCES:
(a)
Letter from Mr. C. H. Cruse (BGE) to NRC Document Control Desk, dated May 23,1997, " Request for Review and Approval of System and Commodity Reports for License Renewal" (b)
Letter from Mr. D. L. Solorio (NRC) to Mr. C. H. Cruse (BGE),
September 1,1998, " Request for Additional Information for the Review of the Calvert Cliffs Nuclear Power Plant, Unit Nos.1 & 2, Integrated Plant Assessment Report for the Feedwater System" (c)
Letter from Mr. D. L. Solorio (NRC) to Mr. C. II. Cruse (BGE),
September 24,1998, " Renumbering of NRC Requests for Additional l Information on Calvert Cliffs Nuclear Power Plant aense Renewal Application Submitted by the Baltimore Gas and Electric Company" Reference (a) forwarded Baltimore Gas and Electric Company (BGE) system and commodity reports for license renewal. Reference (b) fonor ded questions from NRC staff on one of those reports, the integrated Plant Assessment Report on the 1eedwater System. Reference (c) forwarded a numbering system for tracking FGE's response to all of the BGE License Renewal Application requests for additional informatior and the resolution of the responses. Attachment (1) provides our responses to the questions contained m Reference (b). The questions are renumbered in accordance with Reference (c).
9011190057 981116 PDR ADOCK 05000317 P
PDR s.
NRC Distribution Code A036D
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' bocumintControlDesk I'
November 16,1998.
Page 2 Should you have further questions regarding this matter, we will be pleased to discuss them with you.
Very truly yours,
/
- TO WIT:
COUNTY OF CALVERT I, Charles H._ Cruse, being duly sworn, state that I am Vice President, Nuclear Energy Division, Baltimore Gas and Electric Company (BGE), and that I am duly authorized to execute and file this response on behalf of BGE. To the best of my knowledge and belief, the statements contained in this document are true and correct. To the extent that these statements are not based on my personal knowledge, they are based upon information provided by other BGE employees and/or consultants. Such information has been reviewed in accordance with company practice and I believe it to be reliable.
/
/
Subscribed and sworn before me, a Notary Public in and for the State of Maryland and County of h / vfAj /
. thisIMA) day of 71nt/An/A4/1998.
WITNESS my Hand and Notarial Seal:
LASM J k 41I(
Notary Public A I !1FD1 My Commission Expires:
Date CHC/KRE/dtm j
Attachment:
(1) Response to Request for Additional Information; Integrated Plant Assessment Report for the Feedwater System cc:
R. S. Fleishman, Esquire C. I. Grimes, NRC J. E. Silberg, Esquire D. L. Solorio, NRC l
S. S. Bajwa, NRC Resident Inspector, NRC A. W. Dromerick, NRC R. I. McLean, DNR H. J. Miller, NRC J. H. Walter, PSC 1
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1 ATTACHMENT (1) i l
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORT FOR THE FEEDWATER SYSTEM 1
Baltimore Gas and Electric Company Calvert Cliffs Nuclear Power Plant November 16,1998
ATTACHMENT (1)
' RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORT FOR THE FEEDWATER SYSTEM NRC Ouestion No. 5.9.40 He Baltimore Gas and Electric Company (BGE) License Renewal Application (LRA) indicates that thermal stratification is a significant contributor to fatigue usage for the steam generator nozzle and adjacent piping. He application further indicates that the piping adjacent to the Unit 2 steam generator was instrumented with thermocouples to obtain temperature data around the circumference of the pipe.
Provide a sketch of the piping showing the locations of the thermocouples used to measure the temperature data for the steam generator nozzle.
BGE Response The eight pairs of thermocouples were installed under Temporary Alteration 2-95-0061. Figure 1 is a sketch drawn by the technician that installed the thermocouples. The sketch is not to scale.
NRC Ouestion No. 5.9.41 The application indicates that the effect oflocal thermal stratification in the Feedwater System (FWS) j does not extend beyond the first elbow of the vertical pipe run. Provide the basis for this conclusion.
BGE Response 1
The reasons thermal stratification does not extend beyond the first elbow to the vertical run of pipe j
are buoyance forces and fluid mechanics. The trickle flow of cold feedwater completely fills the vertical section of the feedwater piping and then flows along the bottom of the 90* elbow and horizontal piping into the steam generator. Because the vertical piping section is always full of cold water during the low feedwater flow conditions, the hot water in the horizontal piping cannot flow back around the elbow and create a stratified condition in the vertical piping. See Figure 2.7 on page 2-25 of NUREG-0691," Investigation and Evaluation of Cracking Incidents in Piping in PWRs,"
which provides a pictorial representation of thermal stratification at low feedwater flow in horizontal piping adjacent to steam generator nozzles. Figure 2.7 does not show thermal stratification extending beyond the elbow into the vertical run of piping.
Thermal stratification causes both local and global effects. The local effects cause a physical ovalization of the pipe section, inducing a local stress distribution around the circumference of the pipe. The location with the highest stress range and fatigue usage is the safe-end-to-reducer weld located in the horizontal section of feedwater piping leading to each steam generator feedwater nozzle. The global effects cause induced moments in the pipe from the steam generator nozzle to the first pipe restraint. The first pipe restraints were analyzed and shown to have negligible load change due to thermal stratification.
NRC Ouestion No. 5.9.42 He application indicates that a finite element analysis of the steam generator nozzle region was performed to determine the most critical location for fatigue. Provide the following information regarding the finite element analysis:
(a) Indicate the computer code used for the analysis. Describe the method used to verify the computer code.
(b) Provide a description of the model used for the analysis and indicate the assumptions used in the analysis. Include a discussion of stress intensification factors, if any, used in the analysis.
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f ATTACHMENT (1)
' RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORT FOR TIIE FEEDWATER SYSTEM BGE Response (a) The computer code used by BGE's vendor to perform the finite element analysis was ANSYS PC/
Thermal and Linear, Linear Elastic Stress Finite Element Analysis Program, Revision 4.4A. The vendor verified their copy of the ANSYS software under their Quality Assurance Program, which meets the requirements of 10 CFR Part 50, Appendix B. Baltimore Gas and Electric Company has audited and accepted the vendor's Quality Assurance Program.
The vendor took guidance from the ANSYS Verification Manual and chose problems from the manual that represented the ways they use the ANSYS software. They ran the problems and compared the results with those published by ANSYS. They also performed independent hand calculations for each problem to confirm the software results were correct.
(b) A two-dimensional (2-D), axisymmetric finite element model of the feedwater nozzle was
]
constructed. Heat transfer coefficients and boundary fluid temperatures were determined using conventional methods. A unit step temperature transient was input as a boundary condition on the wetted inside surface of the component. Thermal stress Green's Functions were extracted j
from the results for the limiting location in the model (point of maximum stress intensity range).
Pressure stress at the limiting location was also determined using this model. A three-
~ dimensional (3-D) finite element model was also constructed in order to determine stratification i
and moment stresses at the limiting location. Various stratified fluid levels were applied to the model. Heat transfer coefficients and boundary fluid temperatures were determined using conventional methods. The resulting maximum stresses were extracted from the model and were l
conservatively assumed to occur at the limiting location chosen from the 2-D analysis. Stresses i
due to moment loading were also determined using the 3-D model. Multiple circumferential j
locations at the same axial position were selected for monitoring (i.e., same Green's Function, but different stratification and moment stresses). Because two very detailed finite element models were used, no additional stress intensification factors were used.
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NRC Ouestion No. 5.9.43 4
The application indicates that the critical feedwater nozzle welds in Unit I were inspected in 1996, and that no flaws with sizes above the critical flaw size specified in the American Society of Mechanical Engineers (ASME) Code were identified. Characterize the indications, if any, that were identified during i
the inspections. The application also indicates that Unit 2 welds were scheduled for inspection during i
the 1997 refueling outage. Provide the results of the Unit 2 inspections including a characterization of l
any indications identified during the inspections.
4 BGE Response The critical steam generator feedwater nozzle welds that were examined in Unit 1 in 1996, and in Unit 2 in 1997, are the pipe-to-reducer welds and the reducer-to-safe-end welds on all feedwater lines.
All recordable indications were attributed to inside diameter surface geometry.
NRC Ouestion No. 5.9.44 l
Electric Power Research Institute (EPRI) Report TR-107515, " Evaluation of Thermal Fatigue Effects on j
Systems Requiring Aging Management Review for License Renewal for the Calvert Cliffs Nuclear Power Plant," provides the results of the fatigue analyses of the feedwater nozzles. Table 3-16 of the EPRI report indicates that fatigue usage factors, without considering environmental effects, will exceed 2
ATTACHMENT (1)
~ RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORT FOR TIIE FEEDWATER SYSTEM 1.0 prior to 40 years of operation for avo Unit 2 steam generator nozzles. Section 3.1.4 of the EPRI report contains a flaw tolerance evaluation in accordance with criteria in ASME Section XI Non-mandatory Appendix L. The flaw tolerance evaluation, using the environmental crack growth data in proposed ASME Code Case, " Fatigue Crack Growth Rate Curves for Ferritic Steels in PWR Water Environments,"(Revision 1, December 10,1996), indicates that a postulated fatigue flaw in three of the steam generator feedwater nozzles could grow thre, ugh wall in less than one operating cycle. The BGE LRA indicates that corrective actions will be initiated well in advance of reaching a fatigue usage factor of 1.0. Describe the corrective actions that will be initiated when the fatigue usage factor approaches 1.0 at the steam generator feedwater nozzles.
BGE Response The purpose of EPRI report TR-107515, specific to Calvert Cliffs, was to provide evidence that the effects of reactor water environments for ASME Class I components are already compensated for by the portion of the design fatigue curve margin factor of 20 that is ascribed to moderate environmental effects (approximately 4).
The study results do not represent Calvent Cliffs' fatigue analysis of record (AOR) nor the fatigue design basis for any component. The study was intended to be representative of conditions for typical older vintage Combustion Engineering pressurized water reactors, and intended to be compared to the results reported in NUREG/CR-6260, " Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components," for such plants, but was not intended to be a Calvert Cliffs licensing basis calculation. There is no direct applicability of this study to the feedwater nozz!e or piping fatigue AOR or the Fatigue Monitoring Program (FMP). The EPRI study also provided an evaluation of a proposed ASME Nuclear Code Case for environmentally-assisted crack growth, and found the proposed Code Case too overly conservative to be useful in its current form, because the environmental effect is operative at all times. The appropriate ASME Code bodies have been notified of the shortcomings of this particular Code Case.
As disassed in the LRA, the FMP tracks fatigue usage for the critical safe-end-to-reducer weld and before the cumulative usage factor reaches unity, Calvert Cliffs will take corrective action. Calvert Cliffs will determine the appropriate action and may use one or both of the following: 1) Implement ultrasonic testing inspections and crack growth analyses under ASME Section XI Non-mandatory Appendix L requirements; or 2) Replace the reducer, safe end, and horizontal piping.
NRC Ouestion No. 5.9.45 Section 5.9 of the application references a site report (Reference 33 on page 5.9-26 of the application) dated July 1996, for the BGE fatigue evaluation. Other sections of the application reference this report or other apparently similar reports. In December 1997, EPRI issued Report TR-107515. By letter dated February 9,1998, EPRI submitted this report for staffinformation. Describe the extent to which EPRI Report TR-107515 is a current fatigue evaluation and the results of all of the other plant-specific fatigue analyses.
BGE Response Reference 33 of BGE LRA Section 5.9 is the purchase order specification describing the task for the EPRI study published in EPRI report TR-107515. The specification and TR-107515 both refer to the same work effort.
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2, A'ITACHMENT (1)
' RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORT FOR TIIE FEEDWATER SYSTEM The purpose of EPRI report TR-107515, specific to Calvert Cliffs, was to provide evidence that the effects of reactor water environments for ASME Class I components are already compensated for by the portion of the design fatigue curve margin factor of 20 that is ascribed to moderate environmental effects (approximately 4).
The study results do not represent Calven Cliffs' fatigue AOR nor the fatigue design basis for any component. The study was intended to be representative of conditions for typical older vintage Combustion Engineering pressurized water reactors, and intended to be compared to the results i
reported in NUREG/CR-6260 for such plants, but was not intended to be a Calvert Cliffs licensing basis calculation. There is no direct applicability of this study to the feedwater nozzle or piping fatigue AOR or the FMP.
NRC Ouestion No. 5.9.46 Baltimore Gas and Electric Company's program for managing the effects of erosion / corrosion is to monitor the local pipe wall thickness and take corrective action when the wall thickness i.e orojected to fall below a certain minimum value. Your July 30,1998 (Reference 1), response to the stan quest for additional information on the FWS, Question No.13, indicates that this minimum wall thicuess is determined based on internal pressure alone.
(a) Please demonstrate that piping with a pipe wall thinned locally to this minimum wall thickness could withstand all licensing basis loads, including bending.
(b) The minimum wall thickness equation cited in your July 30, 1998, request for additional information response applies only to straight pipes. Please provide the basis for applying this equation to fittings, such as elbows, tees, reducers, and fabricated branch connections.
BGE Response (a) Demonstration that Pressure design governs wall thicknessfor allloading conditions:
The design Code of record for the FWS is B31.1 1967 edition with addenda through 1972. The one exception to this is the piping that is the containment penetration. This piping is B31.71969 edition with addenda though 1971. The penetration piping is classified as Class 2, and per the requirements of B31.7, is designed in accordance with the rule of B31.1 with very few exceptions or additional rules.
The design pressure and temperature of the FWS is 1500 psig at 460 F. The piping material of the system is A-106-C. With these design conditions and materials, the following table of l
minimum wall thicknesses can be easily developed utilizing equation 3 of B31.1. Additionally, a column has been added to reflect the as-designed (nominal) wall thickness.
Size Material Pressure Temperature bo" W
16" A-106-C 1500 psig 460 F 0.663 0.844 0.79 Schedule 80 l
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c, ATTACHMENT (1)
' RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORT FOR THE FEEDWATER SYSTEM A review of the pipe stress calculations for the portions of this system that are in Containment provides the following data regarding the intensified stress and the allowable stress for each ser'; ice level at the peak location (s). Note that the remaining feedwater piping within the scope of license renewal is encapsulated and is, therefore, subject to different monitoring techniques.
Summary of CCNPP Feedwater Piping Primary Stress Levels in Containment (psi)
Weight +
Weight + Pressure +
Weight + Pressure +
Pressure Operating Basis Safe Shutdown Earthquake Earthquake Unit 1 Unit 2 Unit 1 Unit 2 Unit 1 Unit 2 Calculated Stress with/ Tuou 8,538 7,079 13,536 9,503 18,968 12,461 Allowable Stress 17,500 17,500 21,000 21,000 33,160 33,160 Ratio 0.49 0.40 0.64 0.45 0.57 0.38 As shown above, the highest ratio of actual to allowable stress is the pressure design limit. This means that, given the design conditions, geometry, and materials, the global system design is limited by the design pressure. The minimum wall thickness is established based on the design j
pressure. Therefore, if wall reduced globally to the minimum wall based on pressure design, all l
other analyzed loadings will still meet the design basis limits. Localized thinning will have even a lesser impact on the systems ability to meet Code limits because of the reinforcement provided by the adjacent thicker piping. This demonstrates that controlling wall thickness based on pressure considerations will ensure that th.: piping will meet its design function under all required loading conditions. Additionally, maintaining the wall at or above the Code limits will ensure that a large margin exists between the point that the system safety function is challenged and identification of a need for corrective action.
(b) Basis for applying minimum wall equation to items such as elbows, tees, reducers, and fabricated branch connections:
As established under (a) above, the design rules for the FWS are in B31.1 1967 edition with addenda through 1972. The rules for pressure design of components are provided in Section 104.
(Excerpt below) 104.1 Straight Pipe.
104.1.2 Straight pipe Under Internal Pressure.
(a)
Minimum WallThickness 1.
The minimum thickness of pipe wall required for design pressures and for temperatures not exceeding those for various materials listed in the allowable stress tables, including allowances for mechanical strength, shall not be less than that determined by formula (3) as follows:
P.Do 2.(SE. Py) +,
(3y tm =
5
=-
ATTACHMENT (1)
' RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; 4
INTEGRATED PLANT ASSESSMENT REPORT FOR TIIE FEEDWATER SYSTEM As correctly stated in the question, this equation is specifically provided in this context for application to straight pipe. However, further in Section 104 of B31.1, curved piping sections are addressed as follows:
104.2 Curved Segment of Pipe.
104.2.1 Pipe Bends.
(a)
The minimum wall thickness after bending shall not be less than the minimum wall thickness required for straight pipe.
Paragraph 104.2.2 addresses flanged elbows, however, the system in question does not contain American National Standards Institute standard flanged elbows. Instead all large bore elbows have butt weld ends. These are equivalent to a pipe bend with regard to pressure response.
Reducers are addressed by paragraph 104.6, which states in part:
104.6 Reducers.
... Where butt welding reducers are made to a nominal pipe thickness, the reducers shall be considered suitable for use with pipe of the same nominal thickness.
The pipe thickness is established through the use of equation (3) above. The Code as presented here states that the thickness of the reducer is acceptable ifit is the same as the pipe.
For both reducers and bends, the use of the hoop design criteria for straight pipe is expressly applicable from B31.1. Additionally, when one considers that each of these components have circular cross sections, when sectioned along the axis or curve-linear axis, it becomes apparent that :he applied internal pressure will result in similar hoop loading (acknowledging that the pressure will try to straighten the elbow).
For intersections and branch connections, the same is not true. The minimum wall thickness for pressure design is first established as described above, however, additional material must be added to account for the loss of material due to the opening in the main header. The area replacement rules that add this material are described in paragraph 104.3 of B31.1. While additional material must be added, the wall thickness and amount of required material is gos erned by the same minimum wall equation. It is necessary for the engineer to understand the pressure design requirements associated with intersection and follow the rules stated in this paragraph to assure that the minimum wall is correctly establish in the local area of the reinforcement.
Alternatively, this paragraph also states that: "a branch connection may be made by the use of a fitting manufactured in accordance with a standard listed in Table 126.1, and used within the limits of pressure-temperature ratings specified in such standard. A butt welding fitting made in accordance with USAS B16.9 shall be of nominal thickness not less than the nominal thickness required for the adjoining pipe." So here also, the thickness of the straight pipe, which is governed by the design pressure and hoop stress, is the basis for determining the thickness requirements for branch fittings.
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ATTACHMENT (1)
' RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORT FOR TIIE FEEDWATER SYSTEM NRC Ouestion No. 5.9 42 One of the most effective ways of minimizing erosion / corrosion is to control secondary water chemistry, that is, pH and oxygen concentration. Describe whether pH and oxygen concentration are controlled in the FWS and if so, specify the parameter ranges.
BGE Response The Chemistry Program controls pH and dissolved oxygen in the FWS. This is indicated en page 5.9-9 of BGE LRA Section 5.9. During power operation (Mode 1), the following chemistry controls are applicable:
Parameter Units Target Action Level pH (if full-flow condensate polishing) 29.0
<8.8
)
pH (if partial-flow condensate polishing) 29.3
<9.3 pH (if no-flow condensate polishing) 29.7
<9.7 Dissolved Oxygen ppb sl.0
>5.0 Action levels on the FWS chemistry are conservatively set to prevent long-term i
degradation / corrosion of system components. Baltimore Gas and Electric Company takes immediate action to return chemistry parameters to normel values to minimize the time that chemistry parameters are in action level conditions.
NRC Ouestion No. 5.9.48 In order to measure the maximum wall thinning of a given component caused by erosion / corrosion, several measurements at different locations are made and the maximum wall thinning is calculated.
Describe what approach is used for measuring data along a pipe (that is, band, area, moving blanket, or point to point method).
BGE Response The point-to-point method is preferred if previous thickness measurements are available. The band method or moving blanket method is used if there are not previous thickness measurements.
Information concerning the Erosion Corrosion Monitoring Program is provided on pages 5.9-19 and 5.9-20. Detailed information regarding credited aging management programs is readily available for review onsite.
NRC Ouestion No. 5.9.49 Describe the erosion / corrosion degradation of the feedwater check valves which was discovered during their inspection at the Calvert Cliffs plants. How was the inspection performed? Was the wall thinning measured or was the inspection limited only to visual examination?
BGE Response The initial discovery of erosion corrosion in feedwater check valves in 1988 was made through visual inspection subsequent to valve disassembly. The valves were discovered to have material wash-out below the valve seats. No wall thinning was discovered. Periodic ultrasonic testing to measure wall thickness was not instituted until after this event. Please see BGE's previous response to feedwater Question No. 37 in Reference (1), and BGE's response to NRC Question No. 5.9.35 in Reference (2).
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ATTACHMENT (1)
' RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORT FOR TIIE FEEDWATER SYSTEM NRC Ouestion No. 5.9.50 In addition to the predictions by the CHECWORXS computer code, what other selection methods (for example, industry experience and engineerinr, judgment) are used in selecting components for erosion / corrosion inspection (wall thickness measurement)? Describe them briefly.
BGE Response Please see the response to Question No. 31 in Reference (1).
NRC Ouestion No. 5.9.51 To determine the life of the components exposed to erosion / corrosion, it is important to know the rate at which thinning of their walls is occurring. This information can be obtained by using appropriate methods for trending component degradation due to erosion / corrosion. Describe the trending methods used in predicting life of the components. In your trending methods, are you using measured or computer predicted data?
BGE Response Please see the response to Question No.31 in Reference (1). Also see the trending discussion on page 5.9-19 and 5.9-20 of the BGE LRA.
NRC Question No. 5.9.52 What is the frequency of valve inspection in the Preventive Maintenance Program relied on to manage erosion / corrosion?
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BGE Response The Main Feedwater check valve Preventive Maintenance inspection relied on to manage erosion / corrosion is presently conducted at four-year intervals.
NRC Ouestion No. 5.9.53 Describe the materials for replacement components in the FWS due to crosion/ corrosion degradation, such as chromium-molybdenum and carbon steel.
BGE Response The materials used in the FWS, including currently approved replacement materials, are as described on pages 5.9-7 and 5.9-8 of BGE's LRA. Currently approved American Society for Testing and Materials materials for piping include the following: A-106 Grades B and C; and A-335 P11, P12,
?21 and P22.
NRC Ouestion No. 5.9.54 Page 5.9-20 of the application indicates that the Institute of Nuclear Power Operations (INPO) has performed an assessment of the BGE erosion / corrosion program and provided recommendations for enhancements. Please briefly summarize the results of the INPO assessment and outline the INPO recommendations for improvements at the Calvert Cliffs plants.
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ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; LNTEGRATED PLANT ASSESSMENT REPORT FOR Tile FEEDWATER SYSTEM i
1 BGE Response The referenced report summarized the results of a plant evaluation, performed in 1989, that included a status of recommendations related to industry operating experience concerns regardmg eresion corrosion of high energy systems. The specific recommendations were as follows:
Implement a planned inspection program for those steam piping sections and/or fittings that e
are most likely to experience significant erosion during normal and off-normal operating conditions.
Perform an engineering review to identify all piping sections in high-energy (greater than e
200 F) systems that are potentially susceptible to erosion / corrosion wall thinning.
Implement an ongoing inspection program for those piping segments identified as potentially e
susceptible to significant erosion / corrosion. Design the inspection program to detect the onset of pijse erosion, as well as the rate of pipe wall thinning. Conduct extensive spot checks of pipe sections considered most susceptible to erosion / corrosion to identify if unacceptable wall thinning has already occurred.
The plant evaluation concluded that implementation was ongoing and these recommendations were l
considered 'open.'
Within two years of the evaluation, they were re-assessed and considered implemented and ' closed.'
NRC Ouestion No. 5.9.55 Describe incidents of damage or failure of components caused by erosion / corrosion at Calvert Cliffs and associated corrective actions.
BGE Resnonse As noted on page 5.9-20 of the Bf LRA, no component failures of FWS piping within the scope of license renewal occurred since the mecption of the erosion corrosion program.
Baltimore Gas and Electric Company LRA Section 5.9, "Feedwater System," pages 5.9-18 through l
5.9-23, describe FWS Group 3 age-related degradation mechanism / device type combination. Group 3 covers erosion corrosion for piping, check valves, motor-operated valves, and temperature elements.
Historical operating experience, judged to be pertinent, is included in appropriate areas of the BGE LRA to provide insight supporting the aging management demonstration.
l Pertinent operating experience of erosion corrosion degradation of FWS components and the associated corrective actions is described on pages 5.9-20 and 5.9-21 of the LRA. The description on page 5.9-21 (2nd paragraph) states that a replacement of check valves in 1988 was due to valve wall l
thickness being less than minimum wall requirements. In actuality, these two Unit 2 steam generator check valves (2CKVFW-130 and -133) were replaced because they were discovered to have material washout below the valve seats. These valves are within the scope oflicense renewal. Subsequent to that event, three other system check valves (main feed pump discharge check valves), which are not i
within the scope oflicense renewal, were replaced due to minimum wall concerns.
Baltimore Gas and Electric Company's response to Question No.37, forwarded in Reference (1),
identified additional pertinent operating experience for the aforementioned steam generator feedwater header check valves.
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ATTACHMENT (1)
' RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORT FOR TIIE FEEDWATER SYSTEM NRC Ouestion No. 5.9.56 Does the BGE erosion / corrosion program permit weld overlay as a corrective action when degraded components are found?
BGE Response The erosion corrosion program does not specify repair techniques of degraded components. The erosion corrosion program is a discovery program.
HRC Ouestion No. 5.9.57 Describe the extent of inspection of two-inch and less piping as part of the BGE erosion / corrosion program.
BGE Response
'Ihere is no FWS piping two inches and less, within the scope of license renewal, that has erosion corrosion as a plausible age-related degradation mechanism.
Note that Unit 2 Licensee Event Report 98-004, forwarded by Reference (3), describes a moisture separator reheater vent line rupture attributable to flow accelerated corrosion. This two-inch diameter vent line is not in the scope oflicense renewal.
References 1.
Letter from Mr. C. H. Cruse (BGE) to NRC Document Control Desk, dated July 30,1998,
" Responses to Requests for Additional Information for the Review of the Calvert Cliffs Nuclear Power Plant, Units 1 & 2, Integrated Plant Assessment Reports for the Feedwater System and Diesel Fuel Oil System" 2.'
Letter from Mr.
C.
H. Cruse (BGE) to NRC Document Control Desk, dated November 12,1998," Responses to Requests for Additional Information for the Review of the Calved Cliffs Nuclear Power Plant, Units 1 & 2, Integrated Plant Assessment Report for the Feedwater System, and Errata" 3.
Letter from Mr. P. E. Katz (BGE) to NRC Document Control Desk, dated August 24,1998,
" Manual Plant Trip Due to Moisture Separator Reheater Vent Line Rupture" 10
e ATTACHMENT (1)
' ' RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORT FOR THE FEEDWATER SYSTEM
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